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claims | 1. A probe control apparatus, comprising:a probe including,an optical fiber having a tapered end forming an apex,an electronically conductive transparent film formed on a surface of the tapered end, anda first metal film formed on a first surface of said optical fiber other than the surface of the tapered end and electrically connected to said electrically conductive transparent film;a vibrating unit configured to vibrate said probe in a vibrating motion perpendicular to a median axis of said probe;an amplitude detection unit configured to detect an amplitude of said probe;a distance controlling unit configured to control a distance between said probe and a sample;a voltage applying unit configured to apply a voltage between said probe and the sample; andan insulating unit configured to electrically insulate said amplitude detection unit from said probe,wherein said probe control apparatus controls a distance between an apex of said probe and the sample via a shear force gap control method. 2. The probe control apparatus as claimed in claim 1, said voltage applying unit comprising:a pulse voltage applying unit configured to apply a pulse voltage between said probe and the sample in synchronization with a phase of the vibrating motion of said probe. 3. The probe control apparatus as claimed in claim 1, wherein the length of said first metal film is no less than 5 mm and the thickness of said first metal film is from 0.2 μm to 10 μm. 4. The probe control apparatus as claimed in claim 1, wherein the probe further includes:a second metal film formed on a second surface of said optical fiber other than the surface of the tapered end, said second metal film being no less than 10 μm thick; anda transitional metal film formed on a third surface of said optical fiber other than the surface of the tapered end and connecting said first metal film with said second metal film, said transitional metal film having a thickness that continuously increases along a direction from said first metal film to said second metal film. 5. The probe control apparatus as claimed in claim 1, wherein the probe further includes:a material configured to prevent transmission of light from said electrically conductive transparent film to portions of the tapered end other than the apex. |
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summary | ||
046718974 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS EXAMPLE 1 A simulated liquid waste for the concentrated liquid waste occurring in a pressurized water reactor (PWR) nuclear power station was incorporated with an additive in a given amount, and the mixture was dried into powder and solidified with a hydraulic solidifier. The simulated liquid waste had the same composition as the real liquid waste, and an aqueous solution of Na.sub.2 B.sub.4 O.sub.7 was prepared by dissolving H.sub.3 BO.sub.3 with NaOH. The simulated liquid waste contained 10 .mu.Ci of .sup.137 Cs (typical nuclide of nuclear fission products). In an additive tank 9 was placed an aqueous calcium hydroxide solution (0.1 wt %) as the additive, which was maintained at 40.degree. C. by a heater and stirred continuously. Then, a given amount (50 kg/batch) of the simulated liquid waste was introduced into an adjusting and weighing tank 10. The aqueous calcium hydroxide solution was subsequently transferred from the additive tank 9 to the adjusting and weighing tank 10 in such an amount that its calcium content be in equivalent moles to the boric acid present in the simulated liquid waste, and the liquid mixture in the tank was stirred at 40.degree. C. for about one hour. As a result, the sodium borate in the liquid waste reacted with the calcium hydroxide solution to give a hardly water-soluble salt (calcium borate). Subsequently, the simulated liquid waste was introduced into an evaporator 11 and dried into powder. The steam generated by the evaporator 11 was condensed by a condenser 15 and recovered as condensed water, which was stored in a condensed water tank 16 and treated later in a separate system. The exhaust gas passing through the condenser 15 was discharged in the air via a filter 22. The dry powder formed in the evaporator 11 was transferred to a drier 12 provided between the evaporator 11 and a mixer 13, so that the powder is prevented from absorbing water and increasing its water content in the course of its being introduced into the mixer 13. The drier 12 had such a structure that the dry powder could be stored therein for feed to the mixer 13 in a certain rate. Meanwhile, a powdery solidifier (an alkali silicate composition) was introduced into a solidifier tank 17, where it was stored temporarily, and then introduced into a solidifier weighing tank 19 via a rotary feeder 18. The tank 19 was provided with a load cell for controlling the amount of the solidifier introduced. Additional water for solidification was introduced from a water feed system into an additional water weighing tank 20 and weighed. The solidifier comprising the alkali silicate composition and the additional water, after being weighed, were introduced into a solidifier kneading tank 21, where they were kneaded, and then introduced into the mixer 13 containing the dry powder of the simulated radioactive waste. In the mixer 13, the dry powder and the alkali silicate composition in amounts adjusted to 50 wt % each were kneaded and then introduced into a 200-l vessel 14 for solidification. The solidified body obtained in this Example 1 was cut, so that its inside structure was observed. As a result, it was confirmed to be a consistent solid body, with no pores occurring due to the exudation of sodium borate. In the course of the solidification procedure, any exothermic reaction such as the conventional one occurring in the solidification with powdery sodium borate did not occur, either. Since the solidification with powdery sodium borate in prior art had been attended by an exothermic reaction as described above, its content in the solidified product had been limited to at most 30 wt %, and the volume reduction ratio had accordingly been low. In contrast, the present process made it possible to add the solidifier up to at least 50 wt % to thereby raise the volume reduction ratio outstandingly. The solidified product prepared in this Example 1 was further observed for changes in its leaching characteristics and crushing strength with time, and the values obtained thereby were found to be satisfactory. FIG. 2 is a diagram showing changes in relative leaching rate with time, and FIG. 3 showing changes in relative crushing strength with time. The figures shown are relative values assuming the value observed on a solidified body prepared by a process using intact sodium borate to be 1. It was confirmed from these figures that the leaching characteristics were improved on the order of 10.sup.2 and the crushing strength increased 1- to 1.5-fold when the solidification treatment in this Example 1 was conducted after sodium borate was converted into calcium borate. EXAMPLE 2 In the preceding Example 1, the simulated liquid waste incorporated with calcium hydroxide was powdered and the powder was directly solidified. In the present Example 2, however, the powder was solidified after it was further granulated by a granulator, whereby a consistent solid product with good leaching characteristics was likewise obtained. The solidification procedures employed herein are shown in FIG. 4. The concentrated liquid waste occurring in a pressurized water reactor was subjected to the same process of adding calcium hydroxide as in Example 1 and then dried into powder, which was then pelletized by a granulator 23, and about 160 kg of the pellets were packed in the 200-l vessel 14. Subsequently, 160 kg of a solidifier comprising an alkali silicate composition kneaded with water was poured from above into the vessel to effect the solidification. The solidified body prepared in this Example 2 had the same characteristics and effects as the one prepared in Example 1. EXAMPLE 3 The Example 3 used a simulated liquid waste for a concentrated liquid waste consisting chiefly of Na.sub.2 SO.sub.4 occurring in a boiling water reactor, unlike Example 1 and Example 2 for a concentrated liquid waste occurring in a pressurized water reactor. In Example 3, the same procedures as in Example 1 were employed, except that the simulated liquid waste was composed of Na.sub.2 SO.sub.4. It was confirmed that the solidified product prepared in Example 3 had the same characteristics and effects as in Example 1. In this Example 3, the powder was solidified directly. EXAMPLE 4 In Example 4, a powder was solidified after it was pelletized as in Example 2. It was confirmed that the solidified product prepared in the Example 4 had same characteristics and effects as in Example 2. EXAMPLE 5 In the Example 5, as shown in FIG. 5, a concentrated liquid waste occurring in a pressurized water reactor was powdered and granulated, and the granules were micro-encapsulated with a water-insoluble coating and then solidified. A simulated liquid waste used herein had the same composition as in Example 1. The simulated liquid waste was transferred to a storage tank 24, and a given amount (50 kg/batch) of it was transferred from the tank 24 to an evaporator 11, where it was dried into powder. The exhaust gas generated in this case was treated in the same manner as in Example 1. The powder was subsequently shaped into granules, about 0.5 mm in size, by a granulator 25 and then introduced into a reaction tank 27. Separately, a dichloromethane solution of ethylcellulose (9 wt %) and n-hexane as microencapsulation solvents were placed in additive tanks 26 and 29, respectively. In the first step, about 200l of the ethylcellulose solution was introduced into the reaction tank 27 containing the granulated radioactive waste, and the mixture was stirred at 25.degree. C. for 5 minutes to disperse the granules. In the second step, 500l of n-hexane was introduced into the same reaction tank 27, and the mixture was stirred at 25.degree. C. for about one hour. Subsequently, the mixture was cooled rapidly to 4.degree. C. and allowed to stand for 24 hours, after which the supernatant was removed and the capsules formed were separated. In the third step, the capsules were cleaned, and their wall membrane hardened, by 1 m.sup.3 of cold n-hexane, and then transferred into a vacuum drier 28. The organic solvent occurring in this step was stored temporarily in a storage tank 30 and then disposed by burning with a burner 31, while the exhaust gas was passed through a filter 32 and discharged in the air. Subsequently, the capsules were dried completely in the vacuum drier 28 maintained at a temperature of about 60.degree. C., and a given amount (about 160 kg) of the dried capsules were introduced into a mixer 13. A solidifier feed system was arranged in the same manner as in Example 1. About 160 kg of a paste of an alkali silicate composition with water was introduced into the mixer 13 and kneaded with the capsules therein, and the mixture was poured into a 200-l vessel 14 to effect the solidification. The solidified body prepared in this example exhibited the same leaching characteristics and crushing strength as the one prepared in Example 1. EXAMPLE 6 In the Example 6, Na.sub.2 SO.sub.4 solution simulating a concentrated liquid waste occurring in a boiling water reactor was used. It was confirmed that the solidified product prepared in the Example 6 had the same time characteristics and effects as in Example 5. |
description | The present invention relates to an installation for sterilizing objects by means of a radiation source for generating X-ray, gamma or electron radiation, as is used, for example, in industry for the manufacture of medical single-use articles. Such an installation first of all has an irradiation zone, in which the radiation source is arranged. An entry zone is in front of the irradiation zone, and a feed zone for loading the installation with the objects to be treated is in front of said entry zone. An exit zone adjoins the irradiation zone and said exit zone is followed by a subsequent processing zone in which the irradiated objects are output for the subsequent manufacturing process which occurs here. The subsequent processing zone is typically furnished as an isolator or clean room, in which people can be active. Objects to be treated are, for example, containers with a multiplicity of initially empty injection syringes, in which the containers are to be sterilized on the outside in the installation and are filled in the subsequent processing zone. A transportation line which is used to convey the objects passes through the installation. The transportation line is formed by, for example, a conveyor belt. A shield is respectively assigned to the entry zone and the exit zone in order to protect the people located in the vicinity of the installation. Installations for beam sterilization, in which the transportation line is designed in a meandering form in the horizontal plane, are known from experience. The radiation source is positioned in a section lying behind curves, with shields being set-up along the transportation line, which shields attenuate the radiation with each deflection so that radiation which is harmful to people is no longer emitted at the loading and unloading openings of the transportation line, which are open to the outside. WO 2006/111681 A2 discloses an installation for beam sterilization, in which the transportation line is designed in the vertical plane with in each case one step at a horizontally aligned rotary disk. The radiation source lies between the rotary disks and hence is shielded by the latter. The already disclosed installations require a complicated transportation system which moreover generates not insignificant amounts of abrasion as a result of the movement process, which abrasion, as contamination in the installation, must be removed. It becomes more likely that the objects to be treated do not remain in their predetermined position on the transportation line, but rather get stuck or fall off. Finally, curved or stepped transportation lines require more time for the objects to be treated to pass through and need an increased spatial requirement for the entire installation. Given the disadvantages of already known installations, the invention is based on the object of simplifying the conveying of the objects to be treated in the installation and at the same time strictly ensuring radiation safety. A further object consists of achieving a qualified class of cleanliness in the installation, at least from the radiation source up to and including the exit zone. Another object is to design the cladding of the installation to be more mounting and maintenance friendly. Furthermore, it is an object to propose a solution for the sealing of openings on the transportation line, e.g. during the decontamination cycle. Finally, it is an object to efficiently remove the undesired gasses, such as ozone in particular, created at the radiation source from the installation or to at least minimize the creation of such gasses. The installation for sterilizing objects by means of a radiation source first of all has an irradiation zone in which the radiation source is arranged. In front of the irradiation zone there is an entry zone, which in turn is preceded by a feed zone. Adjoining the irradiation zone there is an exit zone which is followed by a subsequent processing zone. A transportation line which is used to convey the objects passes through the installation. In each case, a shield is assigned to the entry zone and the exit zone. The entry zone has a first inlet opening with a passage to the feed zone and a second inlet opening with a passage to the irradiation zone. The exit zone has a first outlet opening with a passage to the irradiation zone and a second outlet opening with a passage to the subsequent processing zone. The transportation line extends flush through the first inlet opening, the second inlet opening, the first outlet opening and the second outlet opening. The shields are movable and in every situation with regard to position of the shields in which one of the two inlet openings or one of the two outlet openings permits the passage of the objects transported through the installation, the other inlet opening and the other outlet opening are covered by the respectively assigned shield. The following features relate to special embodiments of the invention: in one alternative, one shield is composed of two disk-shaped, rotatable elements, a connection extending between the two. The elements have material regions and cut-outs which are offset with respect to each other and do not overlap. The elements are arranged at a distance which allows an object to be held between the former. The shield is connected to a drive. One element has two cut-outs on its circumference, which cut-outs are open in the radial direction and are offset by 180° with respect to one another. Alternatively, the element has one cut-out, whose arc is a multiple of the size of the object and thus permits the latter's passage when the shield is moved. In another alternative, one shield is composed of two plate-shaped, displaceable or pivotable elements, a connection extending between the two. The elements have material regions but no cut-outs and are arranged at a distance which allows an object to be held between the former. The shield in this alternative is also connected to a drive. An air distributor in the form of a filter or a diffuser for generating a unidirectional air flow and a decontamination unit are available in the exit zone. Advantageously, a feed air filter and an exhaust air filter are also arranged in the exit zone. The radiation source acts into a channel through which the transportation line for conveying the objects passes and which firstly extends to the second inlet opening of the entry zone and secondly extends to the first outlet opening of the exit zone. A feed air filter opens into the channel between the first outlet opening and the radiation source. An exhaust air line leads off the channel between the radiation source and the second inlet opening. Alternatively, this exhaust air line can lead off the entry zone. A seal which can be activated is arranged on the second inlet opening in the entry zone with a passage to the irradiation zone, which seal is mounted on a fixed part of the installation and, when activated, seals the second inlet opening against the second element of the shield assigned to the entry zone. Alternatively, the seal can be mounted on the second element of the shield assigned to the entry zone and, when activated, seals the second inlet opening against a fixed part of the installation. The housing of the sterilizing installation is manufactured from sandwich elements which adjoin one another and are sealed off against one another. A sandwich element has a radiation absorbing inner layer, preferably composed of lead, between two outer layers, preferably composed of a stainless steel plate. A seal is provided between adjacently arranged sandwich elements. The treatment space acted upon by the radiation source is filled with an inert gas in order to at least minimize the creation of undesired gasses, such as in particular ozone. The detailed description of an exemplary embodiment of the installation for sterilizing objects by means of a radiation source follows below with reference to the appended drawings. Alternative solutions for the shield used in the installation are presented and, moreover, advantageous design details are disclosed in addition to the essential features. The following provision applies for the entire description that follows: if reference numbers are contained in a figure for the purposes of unambiguity in the drawing but are not referred to in the directly associated text of the description, reference is made to their mentioning in preceding descriptions of figures. In the interest of clarity, the repeated labeling of components in subsequent figures is generally refrained from, as long as the drawing makes it clear without ambiguity that they are “recurring” components. FIGS. 1A, 1B and 10 An irradiation zone 5 forms the center of the installation 1, said zone being preceded by an entry zone 3 and followed by an exit zone 4. In front of the entry zone 3 there is a feed zone 2, in which the transportation line 6 to convey the objects 8 to be treated in the installation 1 starts. Here, the installation 1 is loaded with the objects 8 and the treated objects 8 are output at the end of the exit zone 4 in order to be subject to the continued manufacturing process in the subsequent processing zone 9. The exit zone 4 is adjoined by a subsequent processing zone 9, typically furnished as isolator or clean room, in which people can be active. Such objects 8 are, for example, containers with a multiplicity of initially empty injection syringes, in which the containers are to be sterilized on the outside in the installation and are filled in the subsequent processing zone 9. The housing of the installation 1 is advantageously constructed from a multiplicity of sandwich elements 10. A sandwich element 10 has on its inside a thicker radiation absorbing layer 100—conventionally composed of lead—which lies between two outer layers 101 which, for example, are composed of a stainless steel plate. Seals 102 are inserted between adjoining sandwich elements 10. FIGS. 2A to 3B The transportation line 6 begins in the feed zone 2 and extends flush through the entire installation 1, that is to say the entry zone 3, the irradiation zone 5, and the exit zone 4. Within the irradiation zone 5, the transportation line 6 is surrounded by the channel 103, around which the radiation source 50, preferably in the form of an electron emitter, is arranged. The radiation source 50 acts on the objects 8 passing through the channel 103 on the transportation line 6. A shield 7 is installed in the entry zone 3 and exit zone 4, respectively. A feed air filter 58 connected to the channel 103 is located between the radiation source 50 and the transition to the exit zone 4. An air distributor 45 in the form of a filter or diffuser is provided in the exit zone 4. The entry zone 3 has a first inlet opening 31 with a passage to the feed zone 2 and a second inlet opening 32 with a passage to the irradiation zone 5. The exit zone 4 has a first outlet opening 41 with a passage to the irradiation zone 5 and a second outlet opening 42 with a passage to the subsequent processing zone 9. The transportation line 6 extends flush through the four openings 31,32;41,42. In a first embodiment, the shield 7 has two disk-shaped, rotatable elements 71,72, between which there is a connection 73 which is similar to an axle. A drive 74 acts on the shield 7. The elements 71,72 extend past the four openings 31,32;41,42 to the extent that in certain situations with regard to positions these openings 31,32;41,42 are covered. In the illustrated embodiment, both elements 71,72 are positioned within the entry zone 3 or exit zone 4. Alternatively, the first element 71 could be placed in the feed zone 2 on the first inlet opening 31 and/or the second element 72 could be placed in the irradiation zone 5 on the second inlet opening 32. FIGS. 4, 6A to 7C, 9A and 9B In principle, the shield 7 has two disk-shaped elements 71,72, which are spaced parallel to one another in a congruent form and between which there is an axial connection 73. On its circumference, each element 71,72 has two cut-outs 76 which are open radially and offset by 180° with respect to one another. There are material regions 75 in the remainder of the annulus, while there are large-area apertures toward the center which serve to reduce weight. In accordance with the function of the shield 7 as a protection from the electron flow emitted by the radiation source 50, the elements 71,72 have a radiation-absorbing design with a layer composed of lead. The cut-outs 76 of the first element 71 are offset compared to the cut-outs 76 of the second element 72 to the extent that there is no overlap. The span of the cut-outs 76 must permit the objects 8 to pass through, and the clear distance between the elements 71,72 must be dimensioned in order to be able to hold a section of the transportation line 6 and an object 8 in the latter. A frame 79 serves to support the shield 7 installed in the installation 1. The drive 74 coupled to the shield 7 is preferably an electric motor. This embodiment of the shield 7 is provided for incremental continued rotation. In a position of the shield 7 in which a cut-out 76 comes to rest on the passage of the transportation line 6, an object 8 can be inserted between the elements 71,72, or, on the other hand, can be unloaded. Insertion is effected through the first inlet opening 31 or the first outlet opening 41. Unloading is effected through the second outlet opening 32 or the second outlet opening 42. However, if a material region 75 is in front of the passage of the transportation line 6, that is to say in front of one of the openings 31,32;41,42 in the installation 1, the insertion or unloading of an object 8 is blocked in order to prevent radiation from passing through. The adjustment between the shield 7 installed in the entry zone 3 and the shield 7 positioned in the exit zone 4 will be coordinated depending on the conceptual design of the processing passage of the object 8 to be treated through the installation 1. In the situation with regard to position in accordance with FIG. 9A: the passage between the feed zone 2 and the entry zone 3 via the first entry opening 31 is locked as a result of the material region 75 present; an object 8 is located in front of the entry zone 3; the passage between the entry zone 3 and the irradiation zone 5 via the second entry opening 32 is open as a result of the cut-out 76 present; an object 8 has left the entry zone 3 and passes through the irradiation zone 5; the passage between the irradiation zone 5 and the exit zone 4 via the first outlet opening 41 is open as a result of the cut-out 76 present; the entry of the object 8 into the exit zone 4 is clear; and the passage between the exit zone 4 and the subsequent processing zone 9 via the second outlet opening 42 is locked as a result of the material region 75 present. This ensures that in each situation with regard to position of the advantageously flush transportation line 6 with the correspondingly arranged openings 31,32;41,42, the path for the radiation emitted by the radiation source 5 is never open to the outside in any direction along the transportation line 6. Thus, it is impossible to endanger people located in the vicinity of the installation 1. In the changed situation with regard to position in accordance with FIG. 9B: the passage between the feed zone 2 and the entry zone 3 via the first inlet opening 31 is open as a result of the cut-out 76 present; an object 8 has passed from the feed zone 2 into the entry zone 3; the passage between the entry zone 3 and the irradiation zone 5 via the second inlet opening 32 is locked as a result of the material region 75 present; the object 8 cannot leave the entry zone 3; the passage between the irradiation zone 5 and the exit zone 4 via the first outlet opening 41 is locked as a result of the material region 75 present; and the passage between the exit zone 4 and the subsequent processing zone 9 via the second outlet opening 42 is open as a result of the cut-out 76 present; the object 8 can leave the exit zone 4 toward the subsequent processing zone 9. This case also ensures that the emitted radiation cannot escape to the outside in any direction along the transportation line 6. FIG. 5 In this alternative embodiment with a continuously rotating shield 7, which avoids the disadvantages of the stop-and-go operation, each element 71,72 only has one cut-out 76 whose arc is a multiple of the size of an object 8 and thus permits the latter's passage in the case of continuous movement of the shield 7. The cut-outs 76 of the two elements 71,72 are dimensioned such that they do not overlap. FIG. 8 The present embodiment of the shield 7 has a design which is significantly different and which is composed of two plate-shaped, displaceable or pivotable elements 71,72, between which a connection 73 extends for the coordinated movement of the two elements 71,72. This shield 7 is also connected to a drive 74, but the material region 75 extends over each element 71,72, without cut-outs 76 being present on the latter. FIG. 11 The radiation source 50 acts into a channel 103 through which the transportation line 6 for conveying the objects 8 runs, and which channel 103 firstly extends to the second inlet opening 32 of the entry zone 3 and secondly extends to the first outlet. opening 41 of the exit zone 4. In order to achieve a qualified class of cleanliness from the radiation source 50 up to the entire exit zone 4, the latter is operated with a unidirectional air flow. To this end, an air distributor 45, in the form of a filter or a diffuser, is provided, with a ventilator 47 attached thereto. Furthermore, a conventional decontamination unit 46 is arranged within the exit zone 4. Moreover, the exit zone 4 is advantageously equipped with a feed air filter 48 and an exhaust air filter 49. A feed air filter 58 is positioned in the irradiation zone 5 between the first outlet opening 41 and the radiation source 50, which filter opens into the channel 103. A feed air line 56 extends from the feed air filter 58, and the former leads into the exit zone 4 via a ventilator 57. Feeding via the feed air filter 58 favors the unidirectional character of the flow in the exit zone 4, rinses away ozone possibly still adhering to the treated object 8, and, at the same time, dissipates the ozone created by the electron emitter 50 in the channel 103 against the direction of transportation of the objects 8. To this end, an exhaust air line 55 leaves the entry zone 3 of the channel 103 between the radiation source 50 and the second inlet opening 32. Alternatively, this exhaust air line 55 can also be led out of the entry zone 3. A filter 59 could be provided in the irradiation zone 5. A seal 35 which can be activated is provided for locking the second inlet opening 32 in the entry zone 3, which opening forms a passage to the irradiation zone 5. This seal 35 can be mounted on a fixed installation component and, when activated, seals the second inlet opening 32 against the second element 72 of the shield 7 arranged in the entry zone 3. Alternatively, the seal 35 which can be activated is mounted on the second element 72 of the shield 7 arranged in the entry zone 3 and, when activated, seals the second inlet opening 32 against a fixed installation component. This seal is set when the decontamination phase is in operation in the exit zone 4. In order to avoid e.g. ozone in particular, which is created when the radiation source 50 is switched on as a result of a reaction with the affected air, the treatment space within the channel 103, acted upon by the radiation source 50, is filled with an inert gas 51. |
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claims | 1. A device comprising:a first silicone part derived from a metallic foil stack lamination mold, said first silicone part defining a plurality of surfaces that define a periphery of a layerless volume of said first silicone part, a surface from said plurality of surfaces comprising a plurality of 3-dimensional micro-features, said surface substantially spatially invertedly replicating a stack lamination mold surface formed by a stacked plurality of metallic foil layers comprised by said metallic foil stack lamination mold, wherein at least one micro-feature of said plurality of micro-features is a first electrically conductive micro-feature. 2. The device of claim 1, wherein:at least one of said plurality of micro-features is a protruding undercut. 3. The device of claim 1, wherein:said first silicone part comprises a feature having an aspect ratio greater than 10:1. 4. The device of claim 1, wherein:said plurality of micro-features are arrayed redundantly. 5. The device of claim 1, wherein:said plurality of micro-features are arrayed non-redundantly. 6. The device of claim 1, further comprising:an application specific integrated circuit. 7. The device of claim 1, further comprising:an application specific integrated circuit, wherein said first electrically conductive micro-feature is coupled to said application specific integrated circuit. 8. The device of claim 1, further comprising:an application specific integrated circuit, wherein said first electrically conductive micro-feature conductively connects said application specific integrated circuit with another of said plurality of micro-features. 9. The device of claim 1, wherein:at least one of said micro-features is a detector, said first electrically conductive micro-feature is conductively connected to said detector, and said first electrically conductive micro-feature is adapted to carry an electrical signal from said detector. 10. The device of claim 1, wherein:at least one of said micro-features is a flexible detector, said first electrically conductive micro-feature is conductively connected to said detector, and said first electrically conductive micro-feature is adapted to carry an electrical signal from said detector. 11. The device of claim 1, wherein:at least one of said micro-features is a photoelectrically responsive material, said first electrically conductive micro-feature is conductively connected to said detector, and said first electrically conductive micro-feature is adapted to carry an electrical signal from said detector. 12. The device of claim 1, wherein:said first electrically conductive micro-feature is comprised of a metal. 13. The device of claim 1, wherein:said first electrically conductive micro-feature is comprised of a metal alloy. 14. The device of claim 1, wherein:said first electrically conductive micro-feature is a wire. 15. The device of claim 1, wherein:said first electrically conductive micro-feature is a strip. 16. The device of claim 1, wherein:said first electrically conductive micro-feature is a coil. 17. The device of claim 1, wherein:said first electrically conductive micro-feature is an electrode. 18. The device of claim 1, wherein:said first electrically conductive micro-feature is a fuse. 19. The device of claim 1, wherein:said first electrically conductive micro-feature is an antenna. 20. The device of claim 1, wherein:said first electrically conductive micro-feature is a metallic epoxy. 21. The device of claim 1, wherein:said first electrically conductive micro-feature is a microwell. 22. The device of claim 1, wherein:said first electrically conductive micro-feature is a switch. 23. The device of claim 1, wherein:said first electrically conductive micro-feature is a resistor. 24. The device of claim 1, wherein:said first electrically conductive micro-feature is a capacitor. 25. The device of claim 1, wherein:said first electrically conductive micro-feature is an inductor. 26. The device of claim 1, wherein:said first electrically conductive micro-feature comprises an electromagnetic shield. 27. The device of claim 1, wherein:said first electrically conductive micro-feature is comprised in an accelerometer. 28. The device of claim 1, wherein:said first electrically conductive micro-feature is comprised in a communications network. 29. The device of claim 1, wherein:another of said plurality of micro-features is a second electrically conductive micro-feature. 30. The device of claim 1, wherein:another of said plurality of micro-features is a second electrically conductive micro-feature and is conductively connected to said first electrically conductive micro-feature. 31. The device of claim 1, wherein:another of said plurality of micro-features is a second electrically conductive micro-feature and is separated from said first electrically conductive micro-feature by a dielectric. 32. The device of claim 1, further comprising:a second silicone part attached to said first silicone part, wherein said second silicone part comprises a second electrically conductive micro-feature. 33. The device of claim 1, further comprising:a second silicone part attached to said first silicone part, wherein said second silicone part comprises a second electrically conductive micro-feature, and said first electrically conductive micro-feature is coupled to said second electrically conductive micro-feature. 34. The device of claim 1, wherein:said first silicone part is formed from a material adapted to be bio-compatible with an organism. 35. The device of claim 1, wherein:said first silicone part is formed from a material adapted to be implantable in an organism. 36. The device of claim 1, wherein:said first silicone part is adapted to serve as a bio-sensor. 37. The device of claim 1, wherein:said first silicone part is adapted to serve as a bio-filter. 38. The device of claim 1, wherein:said first silicone part is adapted to serve as a pump. 39. The device of claim 1, wherein:said first silicone part is adapted to serve as a tissue scaffolding. 40. The device of claim 1, wherein:said first silicone part is adapted to serve as a cell sorting membrane. 41. The device of claim 1, wherein:said first silicone part defines a heating channel. 42. The device of claim 1, wherein:said first silicone part defines a cooling channel. 43. The device of claim 1, wherein:said first electrically conductive micro-feature is adapted to serve as an actuator. 44. The device of claim 1, wherein:said first silicone part is flexible. 45. A method comprising:filling with silicone a mold formed from a stacked plurality of lithographically-derived micro-machined metallic layers to form a part having a feature said part comprising a surface that substantially spatially invertedly replicates a mold surface formed by said stacked plurality of lithographically-derived micro-machined metallic layers, said silicone part defining a plurality of surfaces that define a periphery of a layerless volume of said first silicone part, a surface from said plurality of surfaces comprising a plurality of 3-dimensional micro-features, wherein at least one micro-feature of said plurality of micro-features is an electrically conductive micro-feature. 46. The method of claim 45, further comprising:demolding said part from said mold such that said part is not substantially damaged. 47. The method of claim 45, wherein:said stacked plurality of lithographically-derived micro-machined layers defines a protruding undercut. 48. The method of claim 45, further comprising:applying a conductive material to said part to form said electrically conductive micro-feature. 49. The method of claim 45, further comprising:applying a conductive material in multiple stages to said part to form said electrically conductive micro-feature. 50. A device comprising:a first silicone part derived from a metallic foil stack lamination mold, said first silicone part defining a plurality of surfaces that define a periphery of a layerless volume of said first silicone part, a surface from said plurality of surfaces comprising a plurality of 3-dimensional micro-features, said surface substantially spatially invertedly replicating a stack lamination mold surface formed by a stacked plurality of metallic foil layers comprised by said metallic foil stack lamination mold, wherein a subset of said plurality of micro-features are a plurality of electrically conductive micro-features, said plurality of electrically conductive micro-features are composed of a metal or metal alloy;an application specific integrated circuit; anda detector;wherein at least one of said plurality of electrically conductive micro-features is coupled to said application specific integrated circuit and to said detector and is adapted to carry electrical signals between said application specific integrated circuit and said detector. 51. A device comprising:a first silicone part derived from a metallic foil stack lamination mold, said first silicone part defining a plurality of surfaces that define a periphery of a layerless volume of said first silicone part, a surface from said plurality of surfaces comprising a plurality of 3-dimensional micro-features, said surface substantially spatially invertedly replicating a stack lamination mold surface formed by a stacked plurality of metallic foil layers comprised by said metallic foil stack lamination mold, wherein a subset of said plurality of micro-features are a plurality of electrically conductive micro-features, said plurality of electrically conductive micro-features are composed of a metal or metal alloy, a first subset of said plurality of electrically conductive micro-features are microwells, a second subset of said plurality of electrically conductive micro-features are coupled to said first subset and adapted to carry electrical signals from said microwells, said microwells are arranged in an array. 52. A device comprising:a first silicone part derived from a metallic foil stack lamination mold, said first silicone part defining a plurality of surfaces that define a periphery of a layerless volume of said first silicone part, a surface from said plurality of surfaces comprising a plurality of 3-dimensional micro-features, said surface substantially spatially invertedly replicating a stack lamination mold surface formed by a stacked plurality of metallic foil layers comprised by said metallic foil stack lamination mold, wherein a subset of said plurality of micro-features are a plurality of electrically conductive micro-features, at least one of said plurality of micro-features is a pressure sensor, at least one of said plurality of electrically conductive micro-features is coupled to said pressure sensor and adapted to carry electrical signals from said pressure sensor. |
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abstract | A nuclear reactor containment system with passive cooling capabilities. In one embodiment, the system includes an inner containment vessel for housing a nuclear steam supply system and an outer containment enclosure structure. An annular water-filled reservoir may be provided between the containment vessel and containment enclosure structure which provides a heat sink for dissipating thermal energy, in the event of a thermal energy release incident inside the containment vessel, the reactor containment system provides passive water and air cooling systems operable to regulate the heat of the containment vessel and the equipment inside. In one embodiment, cooling water makeup to the system is not required to maintain containment vessel and reactor temperatures within acceptable margins. |
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abstract | A method is provided for decontaminating biological pathogens in a contaminated environment. The method includes: tailoring x-ray radiation to match the absorption characteristics of a contaminated environment; generating x-ray radiation having a diffused radiation angle in accordance with the absorption characteristics of the contaminated environment; and directing the x-ray radiation towards the contaminated environment. |
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053496191 | summary | BACKGROUND OF THE INVENTION 1. Industrial Field of the Invention The present invention relates to a fuel assembly which constitutes a core of a nuclear reactor and, more particularly, to a fuel assembly for a light water reactor whose fuel is plutonium-uranium mixed oxide, and a core of a light water reactor utilizing the same. 2. Description of the Related Art One conventional example of a fuel assembly used for a core of a nuclear reactor is an 8.times.8 type fuel assembly for a boiling water reactor (BWR). This fuel assembly is composed of a bundle of rods comprising a large number of elongated cylindrical fuel rods and water rods. The fuel rod is constituted of a fuel cladding which is filled with a large number of cylindrical UO2 fuel pellets and sealed both at upper and lower ends. The water rod has a cooling water inlet formed in a lower portion thereof and a cooling water outlet formed in an upper portion thereof so that cooling water flows inside of the water rod from the lower portion to the upper portion. FIG. 21 is a horizontal cross-sectional view showing one example of an 8.times.8 type uranium fuel assembly according to the conventional technique. As shown in FIG. 21, the fuel assembly 190 comprises fuel rods 192 of uranium oxide which are arranged in 8.times.8 lattice form, two water rods 196 which are located in a central portion thereof, and a channel box 197 which surrounds the fuel rods 192 and the water rods 196. When the fuel assembly 190 is mounted in a reactor core, a position adjacent to one corner portion of the fuel assembly 190 is used as a space in which a cruciform control rod 191 is inserted. FIG. 22 shows another example of an 8.times.8 type uranium fuel assembly according to the conventional technique. In this fuel assembly 200, one water rod 206 is provided in the center of an 8.times.8 arrangement of fuel rods. The water rod 206, which is a large water rod, is located in an area from which four fuel rods 192 are removed. Except for this water rod, the fuel assembly 200 has substantially the same structure as the above-described fuel assembly 190. In these two conventional examples, it is necessary to suppress the power peaking of the fuel assembly to a predetermined value or less and to prevent power of the fuel assembly per unit length, i.e., a linear heat generation ratio, from exceeding a limit value, in order not to deteriorate the fuel integrity. On the other hand, from the standpoint of effective utilization of uranium resources, there has recently been carried forward a plutonium utilization in LWR program in which plutonium in used uranium fuel taken out of a light water reactor is recycled to the light water reactor. In this program, part or most of uranium fuel rods in a uranium fuel assembly are substituted by fuel rods of mixed oxide (MOX) in which plutonium is enriched, and MOX fuel assemblies thus obtained are mounted, as replacement fuel, on a light water reactor with uranium fuel assemblies. In this case, properties of the MOX fuel assembly should preferably be similar to those of the uranium fuel assembly. Also, the design of uranium fuel has a high-burnup tendency, and accordingly, the design of MOX fuel should preferably have a high-enrichment tendency, i.e., a plutonium load per fuel assembly should preferably be made as large as possible. However, when a load ratio of plutonium of the MOX fuel assembly is increased, there is caused a difference between reactor core properties of the MOX fuel assembly and those of a uranium fuel assembly owing to a difference between nuclear properties of uranium and those of plutonium. That is to say, because plutonium 239 Pu, 241 Pu which is fissile material has a thermal neutron absorption cross section larger than uranium 235 U, and because a neutron resonance effect by plutonium 240 Pu is larger than uranium 235 U, a neutron flux spectrum of MOX fuel is harder than a neutron flux spectrum of uranium fuel, thereby deteriorating neutron moderation. As a result, there are caused a decrease in thermal margin in a transient event due to an increase in an absolute value of void reactivity coefficient, an increase in distortion of the axial power distribution, a decrease in a moderator reactivity coefficient, or a decrease in reactor shutdown margin due to a decrease in a control rod worth. Deteriorations of these properties are in allowable ranges if a fissile plutonium load is about 1/3 or less of the whole fissile material, and consequently, the structural design of the uranium fuel assembly can be used as it is. However, if the fissile plutonium load is higher, a water to fuel ratio of the fuel lattice must be increased to improve the neutron moderation. The following two examples are known as means for improving the neutron moderation of the MOX fuel assembly: (1) Fuel Assembly for Nuclear Reactor (Japanese Patent Unexamined Publication No. 63-172990) In this conventional technique, four fuel rods in the vicinity of one large water rod provided in the center of an MOX fuel assembly are substituted by four water rods having the same diameter as the fuel rods, so that the absolute value of the void reactivity coefficient of the MOX fuel assembly will be decreased. (2) Boiling Water Reactor (Japanese Patent Unexamined Publication No. 63-293493) In this conventional technique, the water to fuel ratio of an MOX fuel assembly is made larger than that of a uranium fuel assembly by increasing the diameter of a water rod located in the center. Thus, various properties of the MOX fuel assembly, such as a void coefficient, an axial power distribution and a reactor shutdown margin, are made substantially equal to those of the uranium fuel assembly. On the other hand, the following five examples are known in relation to the arrangement of water rods in a uranium fuel assembly: (3) Fuel Assembly (Japanese Patent Unexamined Publication No. 60-105990) In this conventional technique, one to three water rods are provided in corner portions of a channel box on the side of a local power range monitor, i.e., on the other side of a control rod, so that errors in the output of the local power range monitor are decreased to thereby improve the load factor and the fuel integrity. (4) Fuel Assembly and Core of Nuclear Reactor (Japanese Patent Unexamined Publication No. 57-23891) In this conventional technique, one to five water rods are provided in corner portions of a channel box on the side of a control rod, so that the thermal duty of fuel rods adjacent to the control rod is decreased so as to prevent generation of cracks. (5) Fuel Assembly (Japanese Patent Unexamined Publication No. 57-583) In this conventional technique, water rods are provided in corner portions of a channel box on the side of a control rod, to thereby prevent a decrease in the thermal margin owing to the control rod history effect. (6) Fuel Assembly (Japanese Patent Unexamined Publication No. 60-201284) In this conventional technique, one part of an upper portion of a fuel rod is formed into a water rod, and such fuel rods are provided in corner portions of a channel box, so as to improve the reactor shutdown margin and flatten the axial power distribution. (7) Nuclear Reactor (Japanese Patent Unexamined Publication No. 60-222791) In this conventional technique, the nuclear reactor comprises fuel assemblies in which the number of water rods provided in each fuel assembly is increased from the center of a reactor core toward the periphery, and there are located in the outermost periphery of the reactor core fuel assemblies in which water rods are provided on the diagonal lines including corner portions of a channel box, to thereby uniform the exposure distribution in the nuclear reactor and to save uranium. The above-described conventional example (1) involves the following problems. That is to say, since the number of water rods is increased and the number of fuel rods is decreased, the average linear heat generation ratio is increased. Therefore, the local power peaking must be further lowered to observe the limit value of the maximum linear heat generation ratio. Consequently, the design of a fuel enrichment distribution becomes more complicated. Further, an increase in the number of water rods and a decrease in the number of MOX fuel rods result in a decrease in the plutonium load per fuel assembly. In the conventional example (2), the void reactivity coefficient of the mixed type fuel assembly is improved. However, the value is still about -9.5 [%K/K/%void]. When it is compared with the value of the uranium fuel assembly which is about -8.3 [%K/K/%void], it can be understood that the improvement effect is insufficient. Moreover, when the known techniques in relation to the arrangement of water rods described in the conventional examples (3) to (7) are applied to the means for improving the neutron moderation, the following problems are induced: In application of the conventional examples (3) to (5), locations of the water rods are asymmetric so that the power peaking will be increased. In application of the conventional example (6), only the part of the upper portion of each fuel rod is formed into a water rod, and consequently, the neutron moderation is not adequately improved. In application of the conventional example (7), the fuel assemblies are special ones which are located in the outermost peripheral range of the reactor core and can not be provided for general use. Further, the disclosed technique is concerned with uranium fuel alone, and MOX fuel is not mentioned. In other words, when this conventional example (7) having more water rods and less fuel rods is compared with a fuel assembly which produces the same power, the fuel rod average heat generation ratio, i.e., the heat generation ratio per unit length of a fuel rod, is larger. In order to avoid this situation, it is suggested to increase the number of water rods in assemblies in the outer peripheral portion of the reactor core which produce low power. Because they are the assemblies which can only be used in the outer peripheral portion of the reactor core, they can not be provided for general use. Furthermore, when a water rod is provided in the second layer from the outermost periphery and located adjacent to a water rod in each corner portion of the fuel assembly, the local power of fuel rods in the outer peripheral portion which are in contact with these two water rods is increased, and it is not favorable from the standpoint of limitation of the linear heat generation ratio. Therefore, location of such fuel assemblies is limited to the outer peripheral portion of the reactor core. The above-described seven problems are concerned with the BWR. On the other hand, in a pressurized water reactor (PWR), the core is not boiling during normal operation so that the void reactivity coefficient as in the case of the BWR is not a problem. However, there is a moderator temperature coefficient serving as an index for-indicating a reactivity change with respect to a water density change. When MOX fuel is used, this moderator temperature coefficient is deteriorated, similarly to the void reactivity coefficient in the case of the BWR. In order to improve this moderator temperature coefficient, it is effective to provide water rods without decreasing the number of MOX fuel rods largely, similarly to the improvement of the void reactivity coefficient in the case of the BWR. SUMMARY OF TEE INVENTION It is an object of the present invention to provide fuel assemblies for the light water reactor and a core of a light water reactor utilizing the same, in which even if MOX fuel assemblies are used in place of uranium fuel assemblies, the void reactivity coefficient or the moderator temperature coefficient can be made substantially equal to that of the uranium fuel assemblies without decreasing the plutonium load largely and without increasing the linear heat generation ratio largely. In order to achieve the above object, this invention provides a fuel assembly for a light water reactor comprising a plurality of fuel rods which contain plutonium as a primary fissile material when exposure is zero, wherein at least one of water rods in which cooling water flows is provided at least in one of each corner portion of an arrangement of the fuel rods and a position adjacent to the corner portion in such a manner that the water rods are located in rotation symmetry, each of the water rods being filled with water over a length at least corresponding to a fuel effective length, and the fuel rods are provided at positions in the second layer from the outermost periphery which are adjacent to those positions at which the water rods are located. In this fuel assembly for the light water reactor, one of the water rods may be located in one of each corner portion of the arrangement of fuel rods and the position adjacent to the corner portion in rotation symmetry, or two of the water rods may be provided in each corner portion of the arrangement of fuel rods and the position adjacent to the corner portion in such a manner that the water rods are located in rotation symmetry, or two of the water rods may be provided in both of the positions adjacent to each corner portion of the arrangement of fuel rods in such a manner that the water rods are located in rotation symmetry. Also, the invention provides a fuel assembly for a light water reactor comprising a plurality of fuel rods which contain plutonium as a primary fissile material when exposure is zero, wherein at least one of solid moderator rods is provided in one of each corner portion of an arrangement of the fuel rods and a position adjacent to the corner portion in such a manner that the solid moderator rods are located in rotation symmetry, each of the solid moderator rods being filled with a solid moderator over a length at least corresponding to a fuel effective length, and the fuel rods are provided at positions in the second layer from the outermost periphery which are adjacent to those positions at which the solid moderator rods are located. In this fuel assembly for the light water reactor, one of the solid moderator rods may be located in one of each corner portion of the arrangement of fuel rods and the position adjacent to the corner portion in rotation symmetry, or two of the solid moderator rods may be provided in each corner portion of the arrangement of fuel rods and in the position adjacent to the corner portion in such a manner that the solid moderator rods are located in rotation symmetry, or two of the solid moderator rods may be provided in both positions adjacent to each corner portion of the ar-rangement of the fuel rods in such a manner that the solid moderator rods are located in rotation symmetry. Further, the invention provides a fuel assembly for a light water reactor comprising a plurality of fuel rods which contain plutonium as a primary fissile material when exposure is zero and a plurality of fuel rods which only contain uranium as a fissile material when exposure is zero, wherein at least one of water rods in which cooling water flows is provided at least in one of each corner portion of an arrangement of the fuel rods and a position adjacent to the corner portion in such a manner that the water rods are located in rotation symmetry, each of the water rods being filled with water over a length at least corresponding to a fuel effective length, and the fuel rods are provided at positions in the second layer from the outermost periphery which are adjacent to those positions at which the water rods are located. In this fuel assembly for the light water reactor, one of the water rods may be located in one of each corner portion of the arrangement of fuel rods and the position adjacent to the corner portion in rotation symmetry, or two of the water rods may be provided in each corner portion of the arrangement of fuel rods and in the position adjacent to the corner portion in such a manner that the water rods are located in rotation symmetry, or two of the water rods may be provided in both positions adjacent to each corner portion of the arrangement of fuel rods in such a manner that the water rods are located in rotation symmetry. Moreover, the invention provides a fuel assembly for a light water reactor comprising a plurality of fuel rods which contain plutonium as a primary fissile material when exposure is zero and a plurality of fuel rods which only contain uranium as a fissile material when exposure is zero, wherein at least one of solid moderator rods is provided at least in one of each corner portion of an arrangement of the fuel rods and a position adjacent to the corner portion in such a manner that the solid moderator rods are located in rotation symmetry, each of the solid moderator rods being filled with a solid moderator over a length at least corresponding to a fuel effective length, and the fuel rods are provided at positions in the second layer from the outermost periphery which are adjacent to those positions at which the solid moderator rods are located. In this fuel assembly for the light water reactor, one of the solid moderator rods may be located in one of each corner portion of the arrangement of fuel rods and the position adjacent to the corner portion in rotation symmetry, or two of the solid moderator rods may be provided in each corner portion of the arrangement of fuel rods and in the position adjacent to the corner portion in such a manner that the solid moderator rods are located in rotation symmetry, or two of the solid moderator rods may be provided in both positions adjacent to each corner portion of the arrangement of the fuel rods in such a manner that the solid moderator rods are located in rotation symmetry. Furthermore, the present invention provides a core of a light water reactor comprising first fuel assemblies each including a plurality of fuel rods which only contain uranium as a fissile material when exposure is zero, second fuel assemblies each including a plurality of fuel rods which contain plutonium as a primary fissile material when exposure is zero, in which second fuel assembly at least one of solid moderator rods is provided at least in one of each corner portion of an arrangement of the fuel rods and a position adjacent to the corner portion in such a manner that the moderator rods are located in rotation symmetry, each of the moderator rods being filled with a filling substance over a length at least corresponding to a fuel effective length, and the fuel rods are provided at positions in the second layer from the outermost periphery which are adjacent to those positions at which the moderator rods are located. In this core of the light water reactor, the second fuel assemblies may have substantially the same shape and dimensions as the first fuel assemblies. Also, in this core of the light water reactor, the second fuel assemblies may be each arranged in such a manner that one of the moderator rods is located in one of each corner portion of the arrangement of fuel rods and the position adjacent to the corner portion in rotation symmetry, or that two of the moderator rods are provided in each corner portion of the arrangement of the fuel rods and in the position adjacent to the corner portion in such a manner that the moderator rods are located in rotation symmetry, or that two of the moderator rods are provided in both positions adjacent to each corner portion of the arrangement of the fuel rods in such a manner that the moderator rods are located in rotation symmetry. In the present invention of the above-described structure, at least one water rod is provided at least in one of each corner portion of the arrangement of the fuel rods and a position adjacent to the corner portion in such a manner that the water rods are located in rotation symmetry, and also, the fuel rods are provided at positions in the second layer from the outermost periphery which are adjacent to those positions at which the water rods are located. Consequently, the effect of improving the void reactivity coefficient per water rod is enhanced. As a result, in respect of the plutonium load and the linear heat generation ratio, there can be provided the MOX fuel assembly which can substitute the uranium fuel assembly. Especially, one water rod is located in each corner portion of the arrangement of fuel rods or in a position in the lattice form which is adjacent to the corner portion, so that the number of additional water rods required for improving the void reactivity coefficient will be suppressed to the minimum. Therefore, the MOX fuel assembly which is substantially equivalent to the uranium fuel assembly can be provided favorably. Further, in order to realize the improvement effect of the void reactivity coefficient, this water rod must have a length substantially equal to that of a region in which nuclear fuel is filled over substantially the entire length of the fuel rod (referred to as a fuel effective portion). |
abstract | A method and apparatus of correcting thermally-induced field deformations of a lithographically exposed substrate, is presented herein. In one embodiment, the method includes exposing a pattern onto a plurality of fields of a substrate in accordance with pre-specified exposure information and measuring attributes of the fields to assess deformation of the fields induced by thermal effects of the exposing process. The method further includes determining corrective information based on the measured attributes, and adjusting the pre-specified exposure information, based on the corrective information, to compensate for the thermally-induced field deformations. Other embodiments include the use of predictive models to predict thermally-induced effects on the fields and thermographic imaging to determine temperature variations across a substrate. |
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summary | ||
claims | 1. An apparatus comprising:an isolation valve assembly including:an isolation valve vessel including a single open end;a mounting flange sealing with the isolation valve vessel to define a sealed volume;a fluid flow line in fluid communication with the mounting flange to flow fluid through the mounting flange; anda valve disposed in the isolation valve vessel inside the sealed volume and operatively connected with the fluid flow line,wherein the fluid flow line both enters and exits the isolation valve vessel through the single open end. 2. The apparatus of claim 1, wherein the isolation valve assembly further includes a forging including the mounting flange and a second flange to which the isolation valve vessel is secured, the forging having a passageway extending between the mounting flange and the second flange through which the fluid flow line passes. 3. The apparatus of claim 1, wherein the valve is a check valve allowing flow in a first direction but not a second opposite direction within the fluid flow line. 4. The apparatus of claim 1, wherein the isolation valve assembly further includes an external isolation valve disposed outside the isolation valve vessel and outside the sealed volume and operatively connected with the fluid flow line. 5. The apparatus of claim 1, further comprising:a nuclear reactor comprising (i) a pressure vessel including a mating flange and (ii) a nuclear reactor core comprising fissile material disposed in the pressure vessel;wherein the mounting flange of the isolation valve assembly is connected with the mating flange of the pressure vessel of the nuclear reactor. 6. The apparatus of claim 5, further wherein the fluid flow line is a makeup line of a reactor coolant inventory and purification system (RCIPS) and the valve is a check valve preventing backflow of reactor coolant from the pressure vessel into the makeup line. 7. An isolation valve assembly for use with a nuclear reactor including a pressure vessel, comprising:an isolation valve vessel having a single open end with a flange;a spool piece having a first flange secured to the pressure vessel, and a second flange secured to the flange of the isolation valve vessel;a fluid flow line passing through the spool piece to conduct fluid flow into or out of the first flange wherein a portion of the fluid flow line is disposed in the isolation valve vessel; andat least one valve disposed in the isolation valve vessel and operatively connected with the fluid flow line,wherein the fluid flow line both enters and exits the isolation valve vessel through the single open end. 8. The isolation valve assembly of claim 7, wherein the at least one valve is a check valve preventing fluid flow out of the pressure vessel. 9. The isolation valve assembly of claim 8, wherein the fluid flow line is a makeup line for supplying reactor coolant to the pressure vessel and the at least one valve is a check valve preventing primary coolant from flowing out of the pressure vessel through the fluid flow line. 10. The isolation valve assembly of claim 7, wherein an end of the fluid flow line is disposed coaxially inside the spool piece. 11. The isolation valve assembly of claim 7, further comprising a redundant valve disposed outside of the isolation valve vessel and operatively connected with the fluid flow line. |
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claims | 1. An electromagnetic wave suppression sheet comprising metallic magnetic particles mixed into a resin and formed into a sheet shape, wherein the electromagnetic wave suppression sheet has a coercive force of 320 A/m or more and a saturation magnetization of 0.35 Wb/m2 or more at a time when an external magnetic field of 1 kOe in an in-plane direction is applied. 2. The electromagnetic wave suppression sheet of claim 1, having a value of (the saturation magnetization at the time when the external magnetic field is applied)/(the coercive force) within a range of 0.0006 to 0.0016 (Wb/m2)/(A/m). 3. The electromagnetic wave suppression sheet of claim 1, wherein the metallic magnetic particle has one of a flat shape and a spherical shape. 4. The electromagnetic wave suppression sheet of claim 1, having an imaginary part of a relative magnetic permeability of 3 or more at 5 GHz. 5. The electromagnetic wave suppression sheet of claim 1, wherein the metallic magnetic particle is made of at least one metallic magnetic material selected from a group consisting of Fe, Co, a Fe—Al—Si-based alloy, a Fe—Si—Cr-based alloy, a Fe—Si-based alloy, a Fe—Ni-based alloy, a Fe—Co-based alloy, a Fe—Co—Ni-based alloy, and a Fe—Cr-based alloy. 6. The electromagnetic wave suppression sheet of claim 1, wherein the resin has particles of a conductive material mixed therein in addition to the metallic magnetic particles. 7. The electromagnetic wave suppression sheet of claim 1, wherein a layer in which the metallic magnetic particles are mixed into the resin and a layer in which a conductive material is mixed into a resin are layered. 8. The electromagnetic wave suppression sheet of claim 1, having a coercive force of 450 A/m or more and a saturation magnetization of 0.5 Wb/m2 or more at a time when an external magnetic field of 1 kOe in an in-plane direction is applied. 9. The electromagnetic wave suppression sheet of claim 1, having a value of (the saturation magnetization at the time when the external magnetic field is applied)/(the coercive force) within a range of 0.0009 to 0.0013 (Wb/m2)/(A/m). 10. The electromagnetic wave suppression sheet of claim 1, having an imaginary part of a relative magnetic permeability of 5 or more at 5 GHz. 11. A device, comprising:an integrated circuit element package; andan electromagnetic wave suppression sheet comprising metallic magnetic particles mixed into a resin and formed into a sheet shape, the electromagnetic wave suppression sheet having a coercive force of 320 A/m or more and a saturation magnetization is of 0.35 Wb/m2 or more at a time when an external magnetic field of 1 kOe in an in-plane direction is applied, the electromagnetic wave suppression sheet being bonded to the package of the integrated circuit element. 12. The device of claim 11, wherein the electromagnetic wave suppression sheet has a coercive force of 450 A/m or more and a saturation magnetization of 0.5 Wb/m2 or more at a time when an external magnetic field of 1 kOe in an in-plane direction is applied. 13. The device of claim 11, wherein the electromagnetic wave suppression sheet has a value of (the saturation magnetization at the time when the external magnetic field is applied)/(the coercive force) within a range of 0.0009 to 0.0013 (Wb/m2)/(A/m). 14. The device of claim 11, wherein the electromagnetic wave suppression sheet has an imaginary part of a relative magnetic permeability of 5 or more at 5 GHz. 15. An apparatus, comprising:an integrated circuit element;a wiring; andan electromagnetic wave suppression sheet comprising metallic magnetic particles mixed into a resin and formed into a sheet shape, the electromagnetic wave suppression sheet having a coercive force of 320 A/m or more and a saturation magnetization of 0.35 Wb/m2 or more at a time when an external magnetic field of 1 kOe in an in-plane direction is applied, the electromagnetic wave suppression sheet being disposed in a vicinity of one of the integrated circuit element and the wiring. 16. The apparatus of claim 15, wherein the electromagnetic wave suppression sheet has a coercive force of 450 A/m or more and a saturation magnetization of 0.5 Wb/m2 or more at a time when an external magnetic field of 1 kOe in an in-plane direction is applied. 17. The apparatus of claim 15, wherein the electromagnetic wave suppression sheet has a value of (the saturation magnetization at the time when the external magnetic field is applied)/(the coercive force) within a range of 0.0009 to 0.0013 (Wb/m2)/(A/m). 18. The apparatus of claim 15, wherein the electromagnetic wave suppression sheet has an imaginary part of a relative magnetic permeability of 5 or more at 5 GHz. |
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claims | 1. A method for controlling a beam extraction irradiation device for heavy ions or protons operating according to the raster scan technique, in which the beam energy, beam focusing and beam intensity are adjusted for every accelerator cycle, wherein the beam extraction is determined for every accelerator cycle. 2. A method according to claim 1, wherein the duration of the beam extraction is adjusted for every accelerator cycle. 3. A method according to claim 1, wherein the particle charge of the extraction beam is adjusted for every accelerator cycle. 4. A method according to claim 1, wherein the beam extraction is interrupted and re-established during an accelerator cycle. 5. A method according to claim 1, wherein the beam focusing is altered during an accelerator cycle. 6. A method according to claim 1, wherein the beam intensity is altered during an accelerator cycle. 7. A method according to claim 1, wherein a field control of the accelerator magnet supply and beam guidance is carried out. 8. A method according to claim 1, wherein the accelerator cycle is terminated on request. 9. A device for controlling a beam extraction irradiation device for heavy ions or protons operating according to the raster scan technique, especially for carrying out the method according to claim 1, wherein an adjusting device is provided for the beam extraction duration of every accelerator cycle. 10. A device according to claim 9, wherein a device for interrupting and re-establishing the extraction beam within an accelerator cycle is provided. 11. A device according to claim 10, wherein extraction and/or injection kickers are provided. 12. A device according to claim 10, wherein a device for KO-extraction of the extraction beam is provided. 13. A device according to claim 1, wherein an adjusting device is provided for modifying focusing and/or intensity of the extraction beam during an accelerator cycle. |
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description | A preferred embodiment of the present invention will be explained by referring to figures. FIG. 1 is a block diagram of an optical device for small angle scattering system relating to an embodiment of the present invention. As shown in FIG. 1, the optical device for small angle scattering system of the present embodiment has an X-ray source 2, a multilayer mirror 1, and a first, second, and third slits 4, 5, 6. The X-ray source 2 uses an X-ray generator of a high output that radiates X-rays from a pointed focus. The multilayer mirror 1 is constituted of a first reflection part 1a and a second reflection part 1b that are orthogonal to each other. The first and the second reflection parts 1a, 1b for use in the present embodiment have a multilayer structure. As shown in FIG. 4, layers 11 of a material having a large atomic number (for instance, nickel Ni, tungsten W or platinum Pt) and layers 12 of a material having a small atomic number (for example, carbon C or silicon Si) are alternately laminated on a surface of a substrate 10. Each layer 11, 12 has a thickness of several to dozens nm, and is formed at a period of 100 to 200 layers. Additionally, in consideration of the effects due to X-ray refraction at each layer 11, 12, each layer 11, 12 is inclined by a predetermined angle with respect to a surface of the substrate 10. Furthermore, each reflection part 1a, 1b is curved in the same elliptic face so as to converge reflected X-rays on one point. The X-ray source 2 is arranged at one focal point A of the above-noted multilayer mirror 1. A distance L2 from the center of the reflection faces of the multilayer mirror 1 to another focal point B (convergent point of reflected X-rays) is set so as to make a convergent angle xcex8c of X-rays at the focal point B almost twice as great as a divergent angle xcex4 of the multilayer mirror 1 (in other words, full width at half maximum of the peak of a rocking curve). This setting may be achieved by adjusting the structure of the multilayer mirror 1, for instance, the elliptical face shapes, materials, and multilayer structures. By such a setting, a distance L1 from the center of the multilayer mirror 1 to the X-ray source 2 is made sufficiently shorter than the distance L2 from the center to another focal point B (L1 less than less than L2). For example, it is assumed that the divergent angle xcex4 of X-rays reflected at the multilayer mirror 1 is 0.05xc2x0, and the solid angle xcex1 of incident X-rays a to the multilayer mirror 1 is 0.27xc2x0. The distance L1 from the center of the multilayer mirror 1 to the X-ray source 2 is assumed to be 250 mm. When the distance L2 from the center of the multilayer mirror 1 to the focal point B (convergent point) is set at 700 mm, the convergent angle xcex8c of the reflected X-rays b becomes nearly twice as great as the divergent angle xcex4 as shown in the following formula. Thus, a preferable arrangement may be obtained. xcex8c=xcex1xc3x97L1/L2=0.27xc3x97250/700≈0.096≈2xcex4 The first and the second slits 4, 5 are provided to prevent the X-rays B reflected at the multilayer mirror 1 from diverging. The third slit 6 shields parastic scattering from the multilayer mirror 1, and is provided near the convergent point B of X-rays. It is preferable to use a quadrantal slit having slit widths that are variable in two axial directions, for the third slit 6. It is preferable to arrange a sample S at the convergent point (focal point B) of the X-rays b reflected at the multilayer mirror 1, and an X-ray detector 3 is installed at the downstream thereof. An imaging plate (IP) is used for the X-ray detector 3 so as to detect small angle scatter X-rays which diverge from the sample S, in a wide range. In the optical device for small angle scattering system of the present embodiment, the X-rays a radiated from the X-ray source 2 enter to the multilayer mirror 1. However, since the distance L1 from the X-ray source 2 to the multilayer mirror 1 is set short as described above, the X-ray intensity a hardly attenuate within this distance and high X-ray intensity may be kept. The X-rays a that entered from one end to the multilayer mirror 1 are alternately reflected between the first reflection part 1a and the second reflection part 1b and output to the other end. Then, the reflecting X-rays b output from the multilayer mirror 1 are converged at the convergent angle xcex8c. As the convergent angle xcex8c is adjusted to nearly twice that of the divergent angle xcex4 herein, almost parallel beam are formed by the X-rays c reflected at the divergent angle xcex4 around the reflected X-rays b as described above (see FIG. 3). Moreover, the convergent angle xcex8c set as above is a small angle as described above. Accordingly, the X-ray detector 3 may be arranged at a location apart from the convergent point B of X-rays (in other words, sample location), and high small angle resolution may be obtained. When the reflected X-rays b are irradiated to the sample S thereby, small angle scattering of X-ray are taken out from the sample S. The small angle scattering of X-ray are detected by the X-ray detector 3 (IP) at the downstream. Additionally, when a distance between the sample S and the X-ray detector 3 is widened, air scattering of X-rays increases and a background rises. Accordingly, S/N ratios in measurement may become worse. In order to solve this problem, it is preferable to cover a gap between the sample S and the X-ray detector 3 with a vacuum pass. Furthermore, for the same reason, it is desirable to cover each gap of the slits 4, 5, 6 therebetween with a vacuum pass. |
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description | This application is a divisional of U.S. patent application Ser. No. 09/993,467 filed Nov. 19, 2001, U.S. Pat. No. 7,112,445, issued Sep. 26, 2006, which is a continuation of International Patent Application PCT/US00/13937 filed May, 19, 2000, which is claims priority on U.S. provisional application 60/135,866 filed May 25, 1999. The specification of each of the above-identified applications is incorporated by reference herein. The present invention relates to the field of identification taggants. More specifically, the present invention relates to the identification tagging of ammunition, such as small arms ammunition. A number of systems have been proposed for use as identification taggants, with an extensive body of work investigating methods for tagging explosives. With respect to ammunition, a system has been proposed and tested wherein the addition of rare-earth elements to ammunition enhanced the delectability of gunshot residue by giving it an unambiguous composition due to incorporation of elements which are easily detected by neutron activation (Bryan et al., 1966). This method was only intended to provide a positive indication of the presence of gunshot residue. It was neither capable of encoding a usefully large number of identification codes, nor was any attempt made to encode any identification information in the taggants. It is an object of this invention to provide a system of and a method for coding taggants which will facilitate economic generation of a very large number of unique identifying codes. The method employs a fragmented coding scheme where a code is comprised of several individual components which are not physically connected to one another. It is further an object of this invention to provide a system of and a method for coding taggants which will minimize the probability of false code readings in chemically reacting or contaminated systems. The method employs a binary or related coding system wherein the value of each bit of the code is indicated by the presence of one component, and the absence of the other component, of a designated pair of chemicals. The method further employs an authentication code system. It is further an object of this invention to provide a system of and a method for tagging ammunition which will minimize concerns about taggant effects on safety and reliability of the tagged ammunition. The method employs a taggant embedded in a thin layer between the primer and propellant in an ammunition round. The method further employs additional layers of material isolating the taggant layer from the primer and the propellant. Known taggant systems and methods fall into three categories. These include: (1) survivable distributed systems and methods; (2) semi-survivable distributed systems and methods; and (3) particulate systems and methods. Distributed Systems Distributed systems encode the taggant information in substances which are distributed through one or more components of the ammunition. These taggants encode information either in the presence or absence of certain chemical substances, or in the relative concentration of certain chemical substances. In distributed systems, the tagging chemicals are directly mixed with other components of the ammunition, and may be exposed to the chemical reactions involved in firing the ammunition. This leads to the further subdivision of the distributed category into the survivable and semi-survivable sub-categories. The survivable systems are those in which the taggant information is encoded in substances which, preferably, will not be altered in any way by chemical reactions. The semi-survivable systems include chemicals which may be affected by the chemical reactions, but for which, preferably, the taggant information has a high probability of surviving the reactions. Of the known systems, only radioactive tracer and isotope ratio systems can be classed as survivable distributed systems. Both of these systems encode information in the isotopic composition of single elements. The chemical reactions involved in firing ammunition will have no significant effect on isotopic compositions. As long as enough atoms can be recovered to determine the isotopic composition of the relevant elements, the taggant information can be read. The semi-survivable systems include chemical tracer and isotopic substitution systems. The chemical tracer system, using rare-earth elements, is considered semi-survivable because the taggant information is encoded in the relative concentration of different elements. Although these ratios are likely to be little affected by the chemical reactions involved in firing ammunition, it cannot be said with certainty that the effect will be negligible. This decreases the degree of reliability of the tagging information obtained by analyzing the residue of expended ammunition tagged with this system. The isotopic substitution system is considered semi-survivable because the chemicals containing the isotopes may be destroyed in the chemical reactions of the ammunition. Although the isotopes themselves cannot be destroyed, the information is encoded in the presence of the isotopes in the substituted chemicals. If the chemicals are destroyed, the taggant information is lost. If the taggant information is encoded in the relative concentration of different substituted chemical compounds, then the taggant information could become corrupted by selective destruction of one of the substituted compounds. In one alternative system information is encoded in the presence or absence of each of a number of chemical elements, isotopes, or compounds in a pre-defined set. This gives improved reliability over the concentration method, but there is still some uncertainty in that some chemical compounds which are initially present in the taggant could be destroyed in firing the ammunition. In the subsequent analysis, it would not be possible to determine whether the absence of a particular compound was the result of its initial absence, or its destruction in the firing. This could lead to incorrect reading of the taggant information. An improved coding scheme has been devised which will provide an indication when tagging chemicals are destroyed. In such a case, the analysis will lead to information which is ambiguous rather than erroneous. The method works by using a binary coding scheme where each bit in the binary code is represented by two chemicals, identified for illustration purposes as chemical A and chemical B. In a representative system the presence of chemical A would indicate a bit value of 0, while the presence of chemical B would indicate a bit value of 1. In analyzing a sample, four outcomes are possible. (1) The presence of only chemical A would indicate a bit value of 0. (2) The presence of only chemical B would indicate a bit value of 1. (3) The absence of both chemicals would indicate that the tagging chemical, and therefore the taggant information, had been destroyed. (4) The presence of both chemicals would indicate that the system had been contaminated, and that therefore the tagging information had been destroyed. Thus, under most circumstances, the analysis will either give the correct result, or indicate that the information had been destroyed. An incorrect result is possible only in a case where the correct tagging chemical had been destroyed, and the system had been contaminated with the incorrect tagging chemical. With only two chemicals, one can tag no more than two separate batches of ammunition. A useful system must be able to provide unique identifying information for far more than two batches, and must be able to encode identifying information corresponding to any type of alphanumeric or other identifier. Most commonly, such an identifier would be a serial number composed of arabic numerals, although other identification systems are possible. The term “serial number” is used hereinafter to encompass all types of symbolic identifiers. By combining multiple pairs of chemicals to build up a binary serial number, an arbitrarily large number of batches can be tagged. For example, to identify one million separate batches would require a binary serial number 20 bits long (220=1,048,576). Tagging these batches using this system would require 40 distinct chemicals, with each of 20 pairs being used to identify the value of one bit in the serial number. If, in analyzing a sample from one of these batches of ammunition, only 19 of the expected 20 chemicals are found, then one bit of the serial number is lost. However, this still narrows the serial number from one million possibilities to only two. While the system is simple with a binary coding scheme i.e., using base-2 numbers, there may be benefits to using other bases. For example, triplets of chemicals could be used to encode a base-3 serial number. In this system, the presence of chemical A, B, or C would indicate a value of 0, 1, or 2 for one trit (base-3 digit) in the serial number. The absence of all three of these chemicals would indicate a loss of information, and the presence of two or more of the chemicals would indicate contamination. Using this system, one million batches of ammunition could be tagged with 39 chemicals in 13 triplets (313=1,594,323). Other bases could also be used, but as the base number gets larger, a point is reached where more rather than fewer tagging chemicals are required. A base-10 system for example, would require 60 chemicals to tag one million batches. The coding system described here could be implemented using ordinary chemical compounds, using compounds in which one or more atoms are substituted with rare isotopes, or using isotopes themselves. While these improvements will make a semi-survivable distributed system more reliable, survivable systems may be preferable. One survivable distributed tagging system of the present invention employs only stable isotopes. In this system, unique taggants, each corresponding to a unique identification code, are created by mixing unique combinations of ratios of multiple stable isotopes of one or more elements. The resulting mixture is added to the substance or product to be tagged. When identification is required, the isotope abundance ratios of the taggant element or elements are measured, and the resultant measurements are compared with the appropriate identification tagging records made at the time the substance was tagged. A code based on an abundance ratio of multiple isotopes of a single element presents two distinct advantages over systems using abundance ratios of elements or compounds. First, the isotopic abundance ratios can be more precisely measured than abundance ratios of elements or compounds. Second, the isotopic abundance ratio will not be modified by non-nuclear physical or chemical processes except those specifically designed for isotope separation, so the taggant code will not be destroyed by chemical reactions or explosions. Elements which could be used for this technique include any element with more than one stable isotope. Of the 83 non-radioactive elements known to exist on earth, 62 have more than one stable isotope, and 40 have more than two stable isotopes. The element tin (Sn) has the largest number (10) of stable isotopes for any single element. The following table lists the symbol of each element under the number of stable isotopes for each of the naturally occurring stable elements. TABLE IElements grouped according to their number of stable isotopes12345678910BeHOSTiCaMoCdXeSnFHeNeCrNiSeRuTeNaLiMgFeZnKrBaAlBSiSrGePdNdPCArCeZrErSmScNKPbWHfGdMnClUPtDyCoVYbAsCuOsYGaHgNbBrRhRbIAgCsInPrSbTbLaHoEuTmLuAuTaBiReThIrTl Among the 40 elements having more than two stable isotopes, there are a total of 222 stable isotopes. These totals include some isotopes which are slightly radioactive, but which have very long half lives and are present in naturally occurring samples of the elements. In most cases, the relative concentrations of the stable isotopes found in any given element anywhere on earth are constant to within one part in fifty thousand. The ratios are easily and precisely measured by various known techniques. Highly enriched samples of most stable isotopes are available commercially. In this system, the abundance ratio of two or more isotopes in each of one or more elements in a substance is artificially controlled to provide for subsequent identification of the substance. For example, for labeling, or tagging, ten commercially prepared batches of ammunition, the element europium (Eu) can be used. It has two stable isotopes with atomic masses of 151 and 153. In natural europium, these two isotopes are present in the concentrations 47.77%, and 52.23% respectively. A code can be created for these batches by preparing a series of isotopic samples containing 151Eu and 153Eu in a patterned series of ten concentration ratios such as 5/95, 15/85, 25/75, 35/65, 45/55, 55/45, 65/35, 75/25, 85/15, and 95/5, with each ratio assigned to one specific batch. These samples can be prepared either with elemental europium, or with europium as an element in a compound such as Eu2O3. A small quantity of one of these samples can be added, by any of a number of means, to each batch of ammunition to be tagged, according to the following table. TABLE IIBatch151Eu/153Eu (Abundance Ratio)0 5/95115/85225/75335/65445/55555/45665/35775/25885/15995/5 Subsequent measurement of the concentration ratio of 151Eu to 153Eu in the ammunition, or in the residue left after it is fired, would yield a ratio identifying the batch in which the ammunition was manufactured. In this example, the ten unique values of the concentration ratio can distinguish each of the ten batches of ammunition. A significant increase in the number of possible unique codes is achieved by using more than one pair of stable isotopes in creating the code. Continuing the above example, the code can be expanded by adding to the ammunition an additional element (e.g. neodymium, Nd) with its own specific concentration ratio of isotopes (e.g. 143Nd and 146Nd). The code can be further expanded by adding a third element with its specific isotope concentration ratio (e.g. dysprosium, 161Dy and 164Dy). The following table illustrates how a system using these three pairs of isotopes can be used to create an identification code (e.g. a three digit serial number). The first column lists the serial number, the remaining columns list the abundance ratios of each of the europium isotopes 151Eu and 153Eu; the neodymium isotopes 143Nd and 146Nd; and the dysprosium isotopes 161Dy and 163Dy, respectively. TABLE IIIIsotope Abundance RatiosSerial Number151Eu/153Eu143Nd/146Nd161Dy/163Dy0005/955/95 5/950015/955/9515/850025/955/9525/75. . .. . .. . .. . .0095/955/9595/9 0105/9515/85 5/950115/9515/85 15/85. . .. . .. . .. . .0995/9595/5 95/5 10015/85 5/95 5/9510115/85 5/9515/85. . .. . .. . .. . .99895/5 95/5 85/1599995/5 95/5 95/5 By reference to this table, measurement of the three abundance ratios 151Eu/153Eu, 143Nd/146Nd, and 161Dy/163Dy in a tagged substance will allow determination of the identification code (e.g. the serial number) of the substance. In this table, not all possible entries are shown. Using the coding scheme of Table III, a total of 103 or 1000 unique serial numbers can be created. Additional pairs of isotopes could be used to provide additional digits, thereby increasing the number of available serial numbers. Following the same pattern, a system using N pairs of isotopes to create serial numbers results in 10N unique serial numbers. The example illustrated in Table III utilized 10% variations in the concentration ratios of each of the isotope pairs. In fact, smaller variations in the isotopic concentration ratios can be used and measured with sufficient accuracy to be useful in the present invention. When two pairs of isotopes are each controlled and measured to within 1% and combined in a single code, there are 1002 or ten thousand (10,000) unique codes available. Three pairs of isotopes at 1% precision would provide for 1003 or one million (1,000,000) unique codes. By extension, N pairs of isotopes, each controlled and measured to within 1% and combined in a single code, would produce 100N unique codes. This system will allow simple and economic generation of a very large number of unique codes, such as would be useful for ammunition tagging. Particulate Systems The particulate category comprises those systems where the taggant information is encoded in small particles which are designed to survive the firing of the ammunition. An example in this category is the color coded plastic beads currently used for tagging explosives in Switzerland. Alternative identifying means also have been proposed for coding the particles, including particle shape, chemical composition, or even microscopic writing. Two principal issues arise when considering application of particulate taggants to ammunition. (1) If the particles are substantially destroyed in the firing of the ammunition, the taggant signal will be degraded or lost. For this reason, the particles are intentionally designed to be robust. This may lead to concerns about their potential effects on firearm mechanisms. (2) The particles are typically manufactured at a remote site, and in large batches, with every particle in a given batch having the same code. Under systems proposed to date, generating one million unique taggant codes would require fabricating one million batches of particles. In the current state of the art, no practical method is available for generating very large numbers of small batches of uniquely identical particles, and for integrating these into an ammunition manufacturing process. A solution to the second problem is to use a fragmented coding system in which each particle encodes only a portion of a serial number. How this system would reduce the required number of distinct batches of particles is best illustrated by example. Suppose it is desired to have a given factory produce a run comprising a series of one million ammunition batches, each with its own serial number. If each taggant particle encodes an entire serial number, this would require one million unique batches of particles. Using a fragmented coding system, the same one million batches could be tagged with 301 batches of taggant particles as follows. The first batch of particles (called the master batch) would contain identifying information about the factory and the run, and could be encoded using any of a number of identifying means as described above. The remaining 300 batches of particles would consist of particles coded with a three element coding system, such as a three-band color code. These batches of particles would be divided equally into three groups; A, B, and C. The one hundred particle batches in group A would consist of particles where the first band is always one color, say blue. The remaining two bands would use a 10 color code to indicate the value of two digits of a digital serial number. The one hundred particle batches in groups B and C would similarly have a first band identifying the group, say yellow and red respectively. The remaining two bands would encode two digits of a digital serial number in the same manner as group A. Each batch of ammunition could then be uniquely identified by introducing particles from the master batch, and from one batch from each of groups A, B, and C. Assume that the 10-color encoding scheme follows the example of the electronics industry and used black, brown, red, orange, yellow, green, blue, violet, gray, and white to represent the digits 0 through 9 respectively. Then ammunition batch number 576,039, for example, would be tagged with the master particles, and with three additional particle batches. The first of these would have blue, green, and violet bands, with the green and violet representing 5 and 7 respectively, and the blue indicating that they encode the first two digits of the serial number. The second batch of particles would have yellow, blue, and black bands, and the third would have red, orange, and white bands. If a sample of residue from the ammunition in this batch is found, the taggant code could be read by finding a particle from each of the four particle batches. The numbers used here were picked for example purposes only. A similar method could be used employing six particle groups, each encoding only one digit of a digital serial number. This would require only 61 batches of particles for one million serial numbers. It is also possible to employ non-digital serial numbers. For example, an 8-color code could be used to encode base-8 serial numbers. Likewise, a 12-color code could be used to encode base-12 serial numbers. Identifying means other than color coding could also be used to encode the serial number components on the particles, or to identify which digits of the serial number are being encoded. The key to reducing the total number of unique batches of particles, and thereby improve manufacturability, is the use of multiple batches of particles to encode a serial number piece by piece. An assembly line would then only need to control the injection of particles from selected batches to build up a large number of serial numbers from a relatively small number of distinct batches of particles. While very useful for ammunition, where identification of large numbers of separate batches would be useful for law enforcement purposes, the method proposed here has more general utility for any field of manufacture where there exists a need to separately identify a large number of discrete units of production. Examples include, but are not limited to, paint, crude oil, fuel oil, hazardous waste, paper, ink, drugs, raw materials used in the manufacture of drugs, chemicals, compact disks, laser disks, computer disks, video tapes, audio tapes, electronic circuits, explosives, currency, clothing, computers, electronic components, and automotive components. Particulate tagging systems can also be combined advantageously with isotopic or chemical tagging systems. One disadvantage of the isotope ratio and chemical tagging systems is that it is not obvious whether or not a taggant is present in a given sample. Without resorting to a sophisticated chemical analysis, a tagged sample will appear identical to an untagged sample. A solution to this difficulty is to combine the isotopic taggant system with another system using particulates that are visible with the unaided eye, or with a simple magnifying glass or microscope. The primary purpose of the particulate taggant would be to indicate the presence of the isotopic or chemical taggant. The particulate taggant may also encode some information, such as the identity of the manufacturer, type of ammunition, date of manufacture, or place of manufacture, but because of its greater versatility, the isotopic or chemical taggant would carry most or all of the identifying information. For any tagging system, there can be a concern about tags which have been counterfeited, altered, or contaminated by other tags. For example, if two rounds of ammunition were produced with powder tagged using the isotope ratio technique, then combining the powder from those two rounds would produce isotope ratios that would match neither of the initial tags. Subsequent reading of the isotope ratio in the powder would not identify either of the initial two batches, but could incorrectly identify a third unrelated batch as the source of the tag. A way to avoid this problem is to use one or more additional pairs or multiples of isotopes to create an authentication code. Each taggant value would have a corresponding authentication code. If a taggant code is accidently created by combining two other codes, or through some other contamination process, it is unlikely that the correct authentication code would also be created. The degree of improbability is determined by the number of unique authentication codes. The following simplified example illustrates the technique. Assume that there are two batches of powder tagged using the isotope ratio system at 10% resolution. The first one is tagged with europium using the isotopes 151Eu and 153Eu in the ratio 25/75. This batch also contains an authentication code in the form of neodymium, with the isotopes 143Nd and 146Nd in the ratio 45/55. The second batch of powder is also tagged with europium, using the isotopes 151Eu and 153Eu in the ratio 45/55. This batch also contains an authentication code in the form of neodymium, with the isotopes 143Nd and 146Nd in the ratio 5/95. If these two batches were mixed in equal amounts, the taggant code of the europium in the combined batch would be read as 35/65, and the authentication code of the neodymium would be read as 25/75. As the taggants were using 10% variations in concentration ratios in forming the code, there is only one chance in 10 that this would be the correct authentication code. By using higher precisions, such as 1% resolution in forming the isotope ratio codes, and additional pairs or multiples of isotopes, the probability of accidently producing a correct authentication code can be made arbitrarily small. Similar authentication coding schemes can be used for particulate and chemical taggants. It may also be advantageous to create an authentication tag using a different system altogether than the identification tag. For example, a fragmented particulate identification taggant could be combined with an isotopic authentication taggant. Other combinations are also possible. Methods of Application Regardless of what type of taggant is used, the taggant must be applied to the ammunition so as to acceptably balance user concerns about possible effects on safety and performance, and the utility of the taggant. The most useful taggant will be one that can be read from the smallest sample of projectile, projectile fragment, or gunshot residue collected from a crime scene. Gunshot residue typically consists of two types of particles. The first is recondensed projectile material which was vaporized by frictional heating of the projectile as it passed through the barrel of the firearm. The second type of particle is composed of the solid residue left behind by the reaction of the primer and propellant charges. Typically, the primer produces the majority of this material. Because most recovered projectiles and projectile fragments will be coated with detectable gunshot residue, a taggant which is uniformly dispersed in the gunshot residue will be of maximum utility. Ideally, it should be present at a concentration high enough to be read from a single residue particle. An obvious way to maximize uniform distribution of the taggant in the residue would be to distribute it uniformly in the propellant charge (typically gunpowder). This method was used in most of the ammunition taggant tests conducted to date. Unfortunately, this method has the drawback that the taggant is in direct contact with the propellant, leading to concerns about sensitizing the propellant for premature ignition. An alternative would be to blend the taggant with the primer reactants. The firing of the ammunition results in mixing of the primer reaction products with the propellant, thereby igniting the propellant. If the taggant is carried in the primer reaction products, it will be blended with the propellant as it is ignited, and will then be distributed throughout the gunshot residue. This method has the advantage that the taggant is not exposed to the propellant before the propellant is ignited. The concern about sensitizing the propellant is removed. However, in this method, the concern is transferred to the primer, which may be even more sensitive to the taggant than is the powder. In an ideal case, the taggant would not be mixed with either the primer or propellant prior to firing the ammunition. This may be accomplished by placing the taggant between the primer and the propellant. When the ammunition is fired, the primer chemicals produce hot reaction products which normally mix with and ignite the propellant. If the taggant is in a layer between the primer and the propellant, it will be fragmented, and/or vaporized by the expansion of the hot primer product vapor. The taggant fragments and/or vapor will be entrained in the expanding gases from the primer, and will be mixed with the propellant as it is ignited. By this method, the taggant will be well dispersed in the gunshot residue. To eliminate any remaining concern about possible sensitization of either the primer or the propellant by the taggant, the taggant can be isolated from both by having it sandwiched between two layers of materials known to be compatible with primer and propellant exposure, respectively. These layers would be of a predetermined thickness sufficient to ensure that the taggant remains isolated from both the primer and the propellant until the ammunition is fired. The isolating layers can be made of any material which is easily shredded, vaporized, burned, or otherwise destroyed by the expanding vapor plume of primer reaction products. Examples of possible barrier materials include paper, wax, and certain plastics. Other materials useful for this application are considered to be equivalents. FIG. 1 is a diagram of a primer showing how this system could be applied. The primer cup 10 contains the primer reactants 12, over which is deposited a protective layer 14, a taggant layer 16, and an additional protective layer 18. The following is a specific embodiment of this system. In manufacturing a round of .38 caliber handgun ammunition, a primer is fabricated using a brass cup containing approximately 15 mg of primer chemicals. Over this is deposited a thin layer of wax, an additional layer containing approximately 15 ng of europium with the isotopes 151Eu and 153Eu in the ratio 25/75, and a final thin layer of wax. The primer is inserted into an empty brass case, to which is added approximately 200 mg of gunpowder propellant, and a projectile. When the round of ammunition has been fully assembled as described, neither the primer nor the propellant is exposed to the europium taggant. When this round of ammunition is fired, the hot expanding vapors from the reaction of the primer chemicals will shred and vaporize the wax layers. The europium will be entrained in the primer vapor and will mix with the propellant as it is ignited. The europium will be oxidized, forming europium oxide, which will condense and mix with the gunshot residue. Since the europium was present initially at one part per million of the primer mass, any residue particle formed of primer material will contain at least 1 ppm of europium. Since the chemical reactions involved will not significantly alter the isotopic abundance ratio, the europium in the gunshot residue particles will have the same isotopic composition as the original taggant. A typical residue particle might have a mass of 3×10−10 g, and will contain at least 3×10−16 g of europium. This is about 1.2 million atoms. Measurement of the isotopic composition of the europium in this particle is possible using various mass spectrometric techniques. The number of atoms present is sufficient to ensure a statistically significant reading of the abundance ratio to better than 1% precision. Reading of this ratio will yield the original tagging isotopic composition, and therefore the serial number of the ammunition batch. An alternative to the wax encapsulated taggant would be to use a pellet insert. The pellet would be fabricated from a material, such as paper, which is easily destroyed by the chemical reaction of the primer or propellant. For example, a small disk of paper would be wetted with a volatile solvent containing a non-volatile taggant. The solvent would be allowed to evaporate, leaving the taggant in the paper. The dry paper disk would then be inserted into the primer cartridge. This is illustrated in FIG. 1, where taggant-containing pellets 20 are shown embedded within the primer reactants. Alternatively, the pellets 22 are attached to the surface of the primer reactants. When the ammunition is fired, the pellet would be destroyed and the taggant would be entrained by the primer vapors, mix with the igniting propellant, and ultimately condense in the gunshot residue. Such paper taggants could also simply be inserted in the cartridge case along with the propellant. This is illustrated in FIG. 2 where the cartridge case 30 contains propellant 32 and a projectile 34. The taggant pellets 36 are distributed throughout the propellant. Alternatively, the taggant pellets 38 can be added after the propellant, and remain between the propellant and the projectile. The paper would be destroyed in firing the ammunition and the taggants would be dispersed. In the pellet system, the taggant would be dispersed throughout the pellet, which acts as a carrier. Alternatively, the taggant may be completely enclosed in a small capsule made of a material easily destroyed in firing the ammunition. This will further ensure that the taggant is completely isolated from the propellant or primer reactants. The taggant capsules could be deployed in the ammunition in the same manner as the pellets described above. To reduce the risk of tampering, the taggant may be deposited such that it is covered by the primer reactants. The taggant may be deposited in the primer case prior to loading the primer reactants. This is illustrated in FIG. 1, where a taggant layer 24 is covered by a protective layer 26, and further covered by the primer reactants 12. If the taggant is easily vaporized, and is covered by a protective layer which is also easily vaporized, the firing of the ammunition would result in the taggant vapor being mixed with the primer vapor as it is expelled into, and ignites, the propellants. The taggant will thus be incorporated in the gunshot residue as it condenses. If it is desired to tag the ammunition without tagging the primer, one could deposit the taggant on the inner wall of the cartridge case, and cover it with a layer of material to isolate it from the propellant. When the ammunition is fired, the covering layer and the taggant will be vaporized, entrained in the burning propellant, and ultimately deposited with the gunshot residue. Were ammunition manufactured on an assembly line, with all the components moving sequentially through the various professing steps into the final packaging for shipment, it would be straight-forward to maintain a clear correspondence between position on the assembly line and the serial number of the ammunition round. This would be very useful for any system incorporating taggants in the primer, since primers are normally manufactured early in the process. Current manufacturing processes, however, typically have the primers being fabricated in batches, which are then installed in cartridge cases in such a way that it would be difficult to keep track of the taggant serial number for any given round of ammunition. A process which would eliminate this issue would be to print a small unique machine-readable label, such as a barcode, on each primer. A record is maintained of the correspondence, between the barcode and the taggant code. As each round of ammunition is boxed for final shipment, the barcode of each primer is read, and a record is maintained of each taggant code in any given box of ammunition. It is understood that the above-described preferred embodiments and examples are simply illustrative of the general principles of the present invention. Other formulations, arrangements, assemblies and materials may be used by those skilled in this art and which embody the principles of the present invention, which is limited only by the scope and spirit of the claims set forth below. |
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abstract | A system for radioisotope production uses fast-neutron-caused fission of depleted or naturally occurring uranium targets in an irradiation chamber. Fast fission can be enhanced by having neutrons encountering the target undergo scattering or reflection to increase each neutron's probability of causing fission (n, f) reactions in U-238. The U-238 can be deployed as layers sandwiched between layers of neutron-reflecting material, or as rods surrounded by neutron-reflecting material. |
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058698418 | claims | 1. A method for imaging a source of radiation comprising: a) using diffracting crystals to focus the radiation; b) analyzing said focused radiation to collect data as to the type and location of the radiation; and c) producing an image using the data. a) supplying a plurality of sources of radiation; b) focussing said radiation onto a detector by means of diffracting crystals; c) analyzing said focused radiation to collect data as to the type and location of the radiation; and d) producing an image using the data. a) supplying a plurality of sources of radiation; b) focussing said radiation onto a detector; c) analyzing said focused radiation to collect data as to the type and location of the radiation; and d) producing an image using the data; wherein the step of focussing said radiation further comprises arranging crystals to diffract said radiation to a predetermined focal point. a) a means for locating the sources of radiation; b) a plurality of diffracting crystals for focussing the radiation emanating from the located sources and directing it to a detector; c) a means for analyzing said directed radiation to collect data as to the type and location of the radiation; and d) a means for converting the data to an image. a) a means for locating the sources of radiation; b.) a means for focussing the radiation emanating from the located sources and directing it to a detector; c.) a means for analyzing said directed radiation to collect data as to the type and location of the radiation; and d.) a means for converting the data to an image; wherein the means for focussing the emitting radiation is a plurality of lenses, each lens comprising a plurality of crystals and wherein the crystals are oriented so as to diffract radiation of a predetermined energy to the same focal point. a) a means for locating the sources of radiation; b.) a means for focussing the radiation emanating from the located sources and directing it to a detector; c.) a means for analyzing said directed radiation to collect data as to the type and location of the radiation; and d.) a means for converting the data to an image; wherein the means for focussing the emanating radiation is a plurality of lenses, each lens comprising a plurality of crystals and wherein the crystals are mounted in concentric rings onto a substrate. 2. The method as recited in claim 1 wherein the diffracting crystals are arranged to diffract the radiation to a predetermined focal point. 3. The method as recited in claim 2 wherein the arrangement of the crystals is determined by a predetermined energy of the radiation. 4. A method for imaging x-ray and gamma radiation comprising: 5. The method as recited in claim 4 wherein the step of supplying said sources of radiation further comprises contacting a body with a radioisotope. 6. The method as recited in claim 4 wherein the step of supplying said sources of radiation further comprises contacting a tumor in vivo with a radioisotope. 7. A method for imaging x-ray and gamma radiation comprising: 8. The method as recited in claim 7, further comprising selecting said crystals to contain random imperfections. 9. The method as recited in claim 7 wherein the step of arranging said crystals further comprises cutting said crystals into thin slabs and bending said cut crystals to assume the shape of circular arcs. 10. The method as recited in claim 4 wherein the step of analyzing said focused radiation further comprises directing said focused radiation to a plurality of detectors with a resolution of from 2 mm to less than 8 mm. 11. The method as recited in claim 4 wherein the step of analyzing said focussed radiation further comprises directing said focussed radiation to a plurality of detectors having a resolution of 2 mm. 12. The method as recited in claim 4 wherein the step of supplying a plurality of said sources of radiation further comprises placing at least one of said sources at precisely known locations. 13. A device for imaging a plurality of sources of x-ray and gamma-ray radiation comprising: 14. The device as recited in claim 13 wherein the means for locating the sources is a plurality of scintillation devices. 15. The device as recited in claim 13 wherein the diffracting crystals form a plurality of lenses. 16. The device as recited in claim 14, where the diffracting crystals are movable relative to the plurality of sources. 17. The device as recited in claim 13 wherein the sources are movable relative to the plurality of diffracting crystals. 18. The device as recited in claim 13 wherein the radiation sources are radioisotopes. 19. A device for imaging a plurality of sources of x-ray and gamma-ray radiation comprising: 20. A device for imaging a plurality of sources of x-ray and gamma-ray radiation comprising: 21. The device as recited in claim 20 wherein the substrate is opaque to gamma-radiation and x-ray radiation. 22. The device as recited in claim 20 wherein the concentric rings onto a substrate are axially juxtaposed medially to the detector and the sources to form an assembly having a longitudinal axis that is perpendicular to the plane formed by the substrate. 23. The device as recited in claim 22 wherein a plurality of assemblies are juxtaposed relative to each other in a plurality of concentric intersecting circular arrays wherein the center of each of the arrays coincides with the location of a source of radiation. |
062787640 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to x-ray optical devices, particularly devices having a tubular shape, and more particularly to high efficiency replicated x-ray optics and a method of fabrication using a super-polished mandrel. 2. Description of Related Art X-ray optical devices are used to change the propagation path of travel of x-rays. These devices can also serve to preferentially select x-rays of a desired wavelength range from a broader band of wavelengths. X-ray optical elements primarily use the mechanism of reflection, in contrast to visible light optics that commonly use refraction. To be efficient, x-ray mirrors must have a surface smoothness on the scale of the x-ray wavelength. Since typical x-ray wavelengths are 1-100 .ANG. for these applications, the surface must be smooth on the atomic scale. To provide such a smooth surface is an exceedingly difficult and time-consuming procedure. In 1952, Wolter proposed the application of a double specular reflection mirror system having a closed surface for focusing of x-rays. This structure was substantially more complex than previous optics and presented serious fabrication difficulties. First attempts to produce Wolter optics were initiated in the 1960's using electrodeposition on negative forms due to the closed surface of these optics. These replication attempts were unsuccessful as very poor figure and surface quality were achieved. In the 1980's, efforts were reinitiated for the development of thin shell structures for space telescopes. These negative form electrodeposition replication efforts have been used in the Czech Republic, Italy, and the United States. Several replication fabricated Wolter structures have been flown in space. These mirrored surfaces achieved the figure and roughness values approaching 15 .ANG. rms that are adequate for those applications, but not for applications requiring greater resolution and using shorter x-ray wavelengths. The replication technique has the potential of lower cost and ease of manufacture. The cost of internally polishing and coating the surface of a tubular optic (typical length 10 cm, average diameter 2 cm) and achieving the smooth internal surface finish required is on the order of $500,000 and requires about one year to fabricate. Each optic device produced would have similar cost and time considerations. By comparison, the use of a negative form mandrel reduces the cost by a factor of 10-100 per mandrel for substrate preparation during development, with further significant cost reductions in the manufacturing stage. In view of the demonstrated effectiveness of the replication approach in the fabrication of moderate resolution Wolter space telescopes, research was directed towards the use of replicated optics for x-ray microscopes used in inertial confinement fusion studies and collimators for x-ray proximity lithography. A primary problem with replicated optics has been achieving smoothness on the replicated part. Past efforts have not been able to achieve a roughness less than 12-15 .ANG. rms. This resulted from the low strength of the layer directly in contact with the mandrel and the lack of control of the adhesion of this layer to the mandrel. Parting of the optic from the mandrel causes plastic deformation of the reflecting layer and degradation of the smoothness of the reflecting surface. The decrease in efficiency and attainable imaging resolution resulting from a surface roughness of 12-15 .ANG. rms is unacceptable. Thus, there is a need for a method to make x-ray optics with a surface roughness less than 12 .ANG. rms. The present invention is based on the recognition that magnetron sputtering deposition can be used, even though previously sputter deposited replicated optics have been of poor quality. The fabrication method of the present invention, based on supporting multilayer structures and a special parting layer, has been developed to produce strong stress-relieved reflecting surfaces with supporting shells that do not deform during the separation process and consequently produce super-smooth surfaces comparable to that of the initial mandrel. SUMMARY OF THE INVENTION It is an object of the present invention to provide replicated x-ray optics having a surface roughness of less than 12 angstroms rms and a method for reproducibly fabricating these x-ray optics with a super-smooth surface. A further object of the invention is to provide x-ray optical devices that have a tubular shape, open at both ends, and an interior surface highly reflective to x-rays within a specified wavelength band. Another object of the invention is to provide x-ray optical devices having shapes that are truncated paraboloidal, ellipsoidal, hyperboloidal, or polynomial shells of revolution. Another object of the invention is to provide a method of fabricating tubular shaped x-ray optics by dc or rf sputter deposition of reflecting layers onto a super-polished reusable mandrel, strengthening the reflecting layers by a sputter deposited multilayer, then further supporting this structure with a low residual stress electrodeposited layer, and separating the layered optical device from the mandrel, resulting in a tubular shell with an interior surface having the shape and surface smoothness of the mandrel. A further object of the invention is to provide increased strength to the reflecting layer resulting from a supporting multilayer, which enhances the ability to part the replica from the mandrel without degradation in surface roughness and performance. Another object of the invention is to provide a parting layer that maintains or enhances the smoothness of the mandrel, provides uniform adhesion, and substantially decreases the adhesion of the reflecting surface material to the mandrel, and reduces the forces required to part the replica structure and thus the potential for increased surface roughness. Yet another object of the invention is to provide a tubular shaped optic wherein the inner reflecting surface can be composed of either a single layer grazing reflection mirror or a resonant multilayer mirror, where the wavelength bandpass of the multilayer mirror can be used to select a specific band of x-ray energies. The invention involves high efficiency replicated x-ray optics and the method of fabrication. The x-ray optical device has a tubular shape that is open at both ends, with the interior surface being highly reflective to x-rays within a wavelength band of interest. A beam of x-rays enters one end, undergoes a single reflection at the interior surface, and exits from the other end with a different direction of travel. The shapes of the optics are truncated paraboloidal, ellipsoidal, hyperboloidal, or polynomial shells of revolution. Optics having a combination of these shapes can also be fabricated from a single mandrel. The tubular optical devices are fabricated using a reusable mandrel with a super-polished surface. The replicated optic is deposited by dc or rf sputter deposition of a reflecting layer or layers onto the mandrel surface, and thereafter the reflecting layers are strengthened by a sputter deposited multilayer, and then this structure is further supported with a low residual stress electrodeposited layer. A special parting layer of sputter deposited amorphous carbon may be deposited on the mandrel surface prior to deposition of the reflecting structure. When the layered device is removed from the mandrel, the tubular shell has an inner surface having the shape and surface smoothness of the master form mandrel. Surfaces having a roughness of less than 10 .ANG. rms, and as low as 3-5 .ANG. rms, have been fabricated. The low stress required to part the replica from the mandrel has made possible the maintenance of the surface figure of the mandrel in the replicated part and has also minimized the potential for damage to the mandrel during parting so that multiple replicas can be manufactured from a single mandrel. The optic elements resulting from the present invention can form single element devices, or combinations of elements can be assembled to form multi-element (compound) devices. The optical elements can be used for applications including x-ray proximity and projection lithography, x-ray crystallography, x-ray microscopy, x-ray radiography, tomography, and x-ray fluorescence analysis. These reflective optics can also be used at longer ultraviolet wavelengths where conventional refractive optics do not exist. Other objects and advantages of the present invention will become apparent from the following description and accompanying drawings. |
046866949 | abstract | An apparatus for analyzing metal alloys includes an electronic unit connected to a hand-held probe unit. The probe unit includes a radiation detector enclosed in a detector housing and a radiation source enclosed in a source housing. The detector housing is generally cylindrical in shape and has an aperture formed in its sidewall. The source housing is formed as a hollow, generally right triangular prism with an open base attached to the detector housing over the aperture and tapering to a tip having an aperture formed therein. The triangular shape of the source housing permits contact measurements in hard to get at places. A shutter drive mechanism is utilized to move a shutter means between a first position blocking radiation and a second position passing radiation from the source to the aperture in the tip of the source housing. Radiation from the source generates X-rays from a sample of material to be analyzed which X-rays pass through the aperture in the tip of the source housing and the aperture in the side wall of the detector housing to the radiation detector. |
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abstract | This disclosure pertains to a holder for a radioactive source capsule with pivoting first and second parts, with redundant mechanisms for retention of the capsule during transportation and handling. These retaining mechanisms include a set screw in the first pivoting part to engage the capsule and to prevent radial movement of the capsule, a locking shelf in the second pivoting part to fix the axial orientation of the pivoting bottom end of the capsule holder, and a capture tooth within the capsule to prevent release of the capsule while in the shipping and/or handling tube of the capsule holder. |
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description | The present application is a Division of application Ser. No. 14/261,991 filed Apr. 25, 2014, which claims benefit under 35 U.S.C. § 119(e) of U.S. Provisional Patent Application No. 61/826,598 filed May 23, 2013 entitled “Hybrid Indirect-Drive/Direct-Drive Target for Inertial Confinement Fusion,” the disclosure of which is hereby incorporated by reference in its entirety for all purposes. The United States Government has rights in this invention pursuant to Contract No. DE-AC52-07NA27344 between the United States Department of Energy and Lawrence Livermore National Security, LLC for the operation of Lawrence Livermore National Laboratory. The present invention relates to relates to inertial confinement fusion, inertial fusion energy, and more particularly to a hybrid indirect-drive/direct-drive target for inertial confinement fusion. In inertial confinement fusion (ICF), a driver—i.e., a laser, heavy-ion beam or a pulse power system—delivers an intense energy pulse to a target containing around a milligram of deuterium-tritium (DT) fusion fuel in the form of a hollow shell. The fuel shell is rapidly compressed to high densities and temperatures sufficient for thermonuclear fusion to commence. The goal of present ICF research is to obtain ignition and fusion energy gain from a DT target. The gain of an ICF target is defined as the ratio of the fusion energy produced to the driver energy incident on the target and is a key parameter in determining economic viability of future inertial fusion energy power plants. The two primary methods of driving ICF targets are “indirect-drive” and “direct-drive.” The National Ignition Facility (NIF) is presently seeking to demonstrate laser-driven ICF ignition and fusion energy gain in the laboratory for the first time by means of indirect-drive. In the latter, the laser energy is first converted to x-rays in a hohlraum surrounding the fuel capsule and the x-rays then perform the ablatively-driven compression of the capsule. Direct-drive is an alternative method of imploding ICF targets where the laser beams impinge directly on the capsule surface and directly cause ablation compression. In both cases, ignition is initiated by the PdV work of the high-velocity converging shell stagnating on a central hotspot. Applicant can define this ignition method as “fast compression ignition”. Features and advantages of the present invention will become apparent from the following description. Applicants are providing this description, which includes drawings and examples of specific embodiments, to give a broad representation of the invention. Various changes and modifications within the spirit and scope of the invention will become apparent to those skilled in the art from this description and by practice of the invention. The scope of the invention is not intended to be limited to the particular forms disclosed and the invention covers all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the claims. The present invention provides a hybrid indirect-drive/direct drive apparatus for inertial confinement fusion utilizing laser beams from a first direction and laser beams from a second direction, comprising: a central fusion fuel component; a first portion of a shell surrounding said central fusion fuel component, said first portion of a shell having a first thickness; a second portion of a shell surrounding said fusion fuel component, said second portion of a shell having a second thickness that is greater than said thickness of said first portion of a shell; and a hohlraum containing at least a portion of said fusion fuel component and at least a portion of said first portion of a shell; wherein said hohlraum is in a position relative to said first laser beam and to receive said first laser beam and produce X-rays that are directed to said first portion of a shell and said fusion fuel component; and wherein said fusion fuel component and said second portion of a shell are in a position relative to said second laser beam such that said second portion of a shell and said fusion fuel component receive said second laser beam. The present invention includes the hybrid indirect-drive/direct drive apparatus for inertial confinement fusion further comprising a fill tube extending through said first portion of a shell or said second portion of a shell to said fusion fuel component. The present invention provides a hybrid indirect-drive/direct drive method for inertial confinement fusion utilizing laser beams from a first direction and laser beams from a second direction, comprising the steps of: providing a unit of fusion fuel, assembling a first portion of a shell having a first thickness partially surrounding the fusion fuel unit, assembling a second portion of a shell having a second thickness greater than the first thickness of the first portion of a shell partially surrounding the fusion fuel unit to complete the shell, assembling a hohlraum containing at least a portion of the fusion fuel unit and at least a portion of the first portion of a shell in a position relative to the first laser beam, shock igniting the first portion of a shell and the fusion fuel using the first laser beam to produce X-rays that are directed to the first portion of a shell and the fusion fuel; and shock igniting the second portion of a shell and the fusion fuel using the second laser beam. The present invention includes a hybrid indirect-drive/direct drive method for inertial confinement fusion further comprising the step of using a fill tube extending through the shell to inject fusion fuel into the unit of fusion fuel. The present invention provides a hybrid, high-gain target for inertial confinement fusion that combines the symmetry advantages of indirect-drive fuel assembly with the efficiency of radial-direct-drive shock ignition in a capsule with thick fuel shells. A slow, thick spherical shell segment of fusion fuel is assembled on a high-density metal guide cone (e.g. gold) by indirect radiation drive in a one-sided hohlraum. It is then shock ignited on the opposite side by radial-direct-drive on a corresponding spherical fuel segment inside the cone. The two fuel segments communicate hydrodynamic energy and momentum at late time via a hole at the cone tip. Such a target is well suited for the laser beam geometry of the National Ignition Facility because the direct-drive side is pure radial; thus it would not require a future polar-direct-drive qualification campaign or new phaseplates in the final optics and will minimize laser cross beam transfer. Its natural two-sided laser illumination geometry and high-gain prospects also make it attractive for future inertial fusion energy power plants. The present invention has use as a high gain target for inertial fusion energy power plants. The present invention has use as an inertial confinement fusion platform for the National Ignition Facility to obtain thermonuclear ignition and fusion energy gain. The invention is susceptible to modifications and alternative forms. Specific embodiments are shown by way of example. It is to be understood that the invention is not limited to the particular forms disclosed. The invention covers all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the claims. Referring to the drawings, to the following detailed description, and to incorporated materials, detailed information about the invention is provided including the description of specific embodiments. The detailed description serves to explain the principles of the invention. The invention is susceptible to modifications and alternative forms. The invention is not limited to the particular forms disclosed. The invention covers all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the claims. The present invention provides a hybrid, high-gain target for inertial confinement fusion that combines the symmetry advantages of indirect-drive fuel assembly with the efficiency of radial-direct-drive shock ignition in a capsule with thick fuel shells. A slow, thick spherical shell segment of fusion fuel is assembled on a high-density metal guide cone (e.g. gold) by indirect radiation drive in a one-sided hohlraum. It is then shock ignited on the opposite side by radial-direct-drive on a corresponding spherical fuel segment inside the cone. The two fuel segments communicate hydrodynamic energy and momentum at late time via a hole at the cone tip. Such a target is well suited for the laser beam geometry of the National Ignition Facility because the direct-drive side is pure radial; thus it would not require a future polar-direct-drive qualification campaign or new phaseplates in the final optics and will minimize laser cross beam transfer. Its natural two-sided laser illumination geometry and high-gain prospects also make it attractive for future inertial fusion energy power plants. Definition of “direct drive:” In the “direct drive” method energy is delivered to the outer layer of the target using high-energy beams. The heated outer layer explodes outward, producing a reaction force against the remainder of the target, accelerating it inwards, compressing the target. A Prior Art Direct Drive target 100b with laser beams 104b converging on a pellet 120b is illustrated in FIG. 1B. Definition of “Indirect drive:” In the “indirect drive” method the lasers heat the inner walls of a gold cavity called a hohlraum containing the pellet, creating a superhot plasma which radiates a uniform “bath” of soft X-rays. The X-rays rapidly heat the outer surface of the fuel pellet, causing a high-speed ablation, or “blowoff,” of the surface material and imploding the fuel capsule in the same way as if it had been hit with the lasers directly. Symmetrically compressing the capsule with radiation forms a central “hot spot” where fusion processes set in—the plasma ignites and the compressed fuel burns before it can disassemble. A Prior Art Indirect Drive target 100c with laser beams 104b converging on a hohlraum 106c and a pellet 102c is illustrated in FIG. 1B. The National Ignition Facility (NIF) is presently seeking to demonstrate laser-driven ICF ignition and fusion energy gain in the laboratory for the first time by means of indirect-drive. In the latter, the laser energy is first converted to x-rays in a hohlraum surrounding the fuel capsule and the x-rays then perform the ablatively-driven compression of the capsule. Direct-drive is an alternative method of imploding ICF targets where the laser beams impinge directly on the capsule surface and directly cause ablation compression. In both cases, ignition is initiated by the PdV work of the high-velocity converging shell stagnating on a central hotspot. Applicant can define this ignition method as “fast compression ignition”. The attractive features of indirect-drive include the radiation smoothing of low-mode laser drive asymmetries by the hohlraum and strong ablative stabilization of Rayleigh-Taylor instabilities due to the deeply penetrating x-rays. It is, however, inefficient due to the low conversion efficiency of laser energy to x-rays in the hohlraum. Not only does this result in only modest fusion energy gain but necessitates thin, high aspect ratio, high velocity fuel shells in order to achieve ignition; such thin shells are susceptible to breakup and mix that may impair the attainment of the ignition temperature. By contrast, because the laser impinges directly on the fuel capsule, direct-drive is a more efficient at converting laser energy into hydrodynamic motion of the shell and higher fusion energy gains can result. However, it lacks the smoothing features of indirect radiation drive and thus the imploding shell can be more susceptible to asymmetry and stability issues. Moreover, the laser beams on NIF are configured for indirect-drive—that is, they are arranged in four hemispherical-opposed cones from 23.5 deg to 50 deg in order to thread through holes at the top and bottom ends of the hohlraum—whereas, in principle, direct-drive requires symmetric drive beams over 4€ solid angle. Tests of direct-drive on NIF are possible in “polar-direct-drive” where the beams are retained in the present up-down, indirect-drive port configurations but where sufficient drive uniformity may be achievable by a combination of beam repointing and partial defocusing. The latter fix may also incur cross-beam power transfer where incoming beams from one direction scatter power off beams refracted from other directions. Finally, while fully spherical beam illumination geometry is ideally required for direct-drive target performance, it is not an optimum geometry for an inertial fusion energy power plant because of the large number of penetrations required through the target chamber vessel for the beam ports. A two-sided-drive geometry is much preferred. In the case of direct-drive (but not indirect-drive), in addition to the conventional ignition method of fast compression of a high velocity shell on the hotspot, a developing ignition concept called “shock ignition” is under study where compressed fusion fuel is separately ignited by a strong late time shock. Here, the fuel assembly and ignition phases are decoupled as follows: The cryogenic shell is initially imploded on a low adiabat using a laser main drive of modest peak power and lower total energy. While the resulting low implosion velocity yields only a low temperature central region, the low adiabat of the fuel leads to high values of the assembled areal and mass densities. The compressed fuel is then separately ignited from a central hotspot heated by a strong, spherically-convergent shock driven by a high intensity spike at the end of the laser pulse. The launching of the ignition shock is timed to reach the center just as the main fuel is stagnating and starting to rebound. Because the implosion velocity is significantly less than that required for conventional (fast-compression) hotspot ignition, considerably more fuel mass can be assembled for the same kinetic energy in the shell. This larger burning fuel mass then provides higher fusion gains/yields for the same laser drive energy or, equivalently, retaining acceptable gains at lower drive energies. Shock ignition can be considered a more efficient way to perform direct-drive and higher target gains result. However, being direct-drive, it still ideally requires the uniformity of full symmetric beam illumination and thus on NIF will have to depend on polar-direct-drive illumination with possibly attendant cross beam energy transfer issues. It should be noted that that it is not feasible to achieve shock ignition in pure indirect-drive. While the drive laser would be capable of providing the required fast rise of the shock pulse, the resulting radiation drive temperature rises only slowly due to the thermal inertia of the hohlraum. Secondly, because the shock is not launched until late time where the capsule has converged to around one third of its original radius, the now large case-to-capsule ratio results in low radiation coupling efficiency. The result is that only a weak decaying shock would be launched through the shell with no contribution to the central temperature at stagnation. An alternative and more speculative approach to inertial confinement fusion presently under study is “fast ignition” that, like shock ignition, seeks to decouple fuel assembly from the ignition process, and may circumvent some of the above issues that encumber both conventional indirect and direct-drive. However, fast ignition requires two physically distinct, time-synchronized laser systems—a main “slow” laser driver (˜20-40 ns) to compress the fuel and a separate fast, ten-petawatt-class laser (˜10's ps), to create a high energy (˜MeV) electron beam in the target to ignite the fuel. In particular, given the very demanding timing and spatial focusing requirements, present studies suggest that fast ignition may not be viable without some breakthrough in the efficiency of the energy channeling from the fast igniter laser beam to the ignition hotspot. Because of these issues, there are no plans to attempt fast ignition on NIF in the foreseeable future. A variant of fast ignition known as “impact fast ignition” attempts to circumvent the difficulty of channeling the energy of the high power, short pulse laser into the high energy electron beam by instead causing the fast laser to drive a thin, high velocity (2×10{circumflex over ( )}8 cm/s) flyer plate that stagnates against the compressed fuel. This approach has the critical issues of inflight breakup of the flyer plate during to instabilities and mix of the flyer plate material with the high density fuel during stagnation that can prevent the hotpot from igniting. This approach shares with the hybrid target the concept of employing a guide cone for spherical fuel segments. The major difference is that impact ignition is a fast ignition variant in that the fuel is assembled isochorically (constant density) with no low density hotspot and is then fast ignited by the impacting flyer plate. By contrast, the hybrid target employs conventional isobaric assembly such that the high density cold compressed fuel and low temperature hotspot are in pressure equilibrium during stagnation and ignition of the hotspot is produced by shock ignition. Other design and operational differences are highlighted in FIG. 3 below The present invention provides a new concept for a hybrid, high-gain ignition target for inertial confinement fusion that combines the symmetry features of indirect-drive fuel assembly with the efficiency of radial-direct-drive shock ignition in a capsule with thick fuel layers. The present invention provides a hybrid, high-gain target for inertial confinement fusion that combines the symmetry advantages of indirect-drive fuel assembly with the efficiency of radial-direct-drive shock ignition in a capsule with thick fuel shells. A slow, thick spherical shell segment of fusion fuel is assembled on a high-density metal guide cone (e.g. gold) by indirect radiation drive in a one-sided hohlraum. It is then shock ignited on the opposite side by radial-direct-drive on a corresponding spherical fuel segment inside the cone. The two fuel segments communicate hydrodynamic energy and momentum at late time via a hole at the cone tip. Such a target is well suited for the laser beam geometry of the National Ignition Facility because the direct-drive side is pure radial; thus it would not require a future polar-direct-drive qualification campaign or new phaseplates in the final optics and will minimize laser cross beam transfer. Its natural two-sided laser illumination geometry and high-gain prospects also make it attractive for future inertial fusion energy power plants. Conventional indirect-drive offers low mode drive symmetry and strong ablative stabilization during the capsule drive. It is, hover, inefficient, requires thin, high velocity fuel shells and results in low fusion energy gains. By contrast, shock ignition in direct-drive offers higher drive efficiency in slow, thick shells but implementation ideally requires fully symmetric laser beam illumination which is not optimum for inertial fusion energy plant applications; testing of shock ignition on NIF will necessitate polar-direct-drive illumination and potentially new phaseplates in the laser final optics and may incur the penalty of cross beam energy transfer. In this hybrid concept, Applicant exploits the advantages of both approaches in a hybrid target configuration. A design example is shown in FIG. 1A. A slow, thick ˜250 deg spherical shell segment of DT fuel is assembled on a high density metal guide cone (e.g., gold) by indirect-drive in a one-sided hohlraum. The gold surface is coated with a low-atomic number anti-mix layer, e.g. C. It is then shock ignited on the opposite side via direct-drive on a ˜110 deg spherical fuel segment inside the cone. Given that the target has a natural two-sided symmetry and the direct-drive side is pure radial drive over a ˜110 deg fuel segment, it doesn't require a polar-direct-drive qualification campaign on NIF or new phaseplates and should eliminate cross beam for transfer between the direct-drive beams. Its two-sided illumination geometry and high-gain prospects also make it attractive for future inertial fusion energy plants. The two fuel segments both comprise thick layers of solid cryogenic DT. On the indirect-drive side, the fuel shell segment is backed by an ablator shell segment comprising plastic (CH) or other low-atomic-number material such as Be, diamond, SiC, B4C, etc., containing a small fraction (˜2-4% atomic) of higher atomic number dopant (e.g., Si) to absorb the high-frequency M-band radiation from the hohlraum. The radial composition of this shell is similar to those under consideration for the present NIF indirect-drive target for the National Ignition Campaign (hereinafter the “NIC” target), except the DT fuel layer is much thicker and the ablator is thinner (see design specifications below). The direct-drive side segment is simply a very thick layer of solid cryogenic DT that acts as both ablator and fuel; approximately half the DT ablates outwards during the drive and the other half is compressed inwards. A thin (˜10-15 ìm) plastic CH seal coat is required on the outside of this all-DT segment but this burns off early in the laser drive. If it should prove difficult to produce smooth solid DT “ice” layers on each side of the cone using the conventional “beta-layering” technique established for the NIC target, Applicant has the option of employing liquid DT wicked into low density (˜25 mg/cc) CH foam shells. This latter variant adds the possible complication of impurity mix in the DT fuel, but because the DT fuel is in liquid form it would not exhibit the surface structure (“roughness”) of solid frozen DT that can form a seed source for instability growth. The two converging fuel segments communicate energy and momentum at late time via a ˜50-100 μm diameter hole at the cone tip. The hotspot on the direct-drive side is shock ignited by a spherically-converging shock driven the high intensity spike at the end of the direct-drive laser pulse with the hotspot tamped by the assembled areal density of the stagnated fuel on the indirect-drive side. The ignition energy from the direct-drive shock ignition side is transmitted to the indirect-drive side and a thermonuclear burn wave propagates into the cold compressed fuel of the latter. Given that the majority (˜80%) of the DT fuel resides on the indirect-drive side, a corresponding majority of the total thermonuclear yield accrues from that side. FIG. 1A illustrates the general layout of the indirect-dive/direct drive target for inertial confinement fusion of the present invention. The following numbered components are illustrated in FIG. 1A: 100 HYBRID INDIRECT-DRIVE/DIRECT-DRIVE TARGET ASSEMBLY 102 GOLD HOHLRAUM 104 GOLD HOHLRAUM AI BACKUP 106 GUIDE CONE 108 INDIRECT-DRIVE FUEL SEGMENT 110 DIRECT-DRIVE FUEL SEGMENT 112 CENTER LINE 114 INDIRECT-DRIVE LASER BEAM ARRAY 116 DIRECT-DRIVE LASER BEAM ARRAY 118 GAS FILLED VOLUME D-T GAS 120 INNER LAYER SOLID DEUTERIUM-TRITIUM (D-T) 122 OUTER LAYER CH PLASTIC+2% Si The reference number 100 indicates the target assembly that is symmetrical about the centerline 112. The main components that comprise the target assembly 100 are the gold hohlraum 102 and it's aluminum backing 104, a guide cone 106 an indirect-drive fuel segment 108 and a direct-drive fuel segment 110. The two polar laser beams arrays the indirect-beam array 114 and the direct-drive beam array 116. Some of these items will be described in greater detail in FIG. 2. In FIG. 1A Applicant shows the design specifications for a high gain experimental test platform for this hybrid target for fielding on NIF. The DT fuel layer thicknesses on the indirect-drive is 210 μm and in this design example uses a Si-doped (2%) CH plastic ablator of thickness 120 μm; alternative ablator materials could include Be, diamond, SiC, B4C, etc. The direct-drive side comprises a 350 μm-thick segment of DT that acts as both fuel and ablator within a thin 15 μm seal coat. The two fuel layers are factors of ×3 and 5× thicker than the fuel layer for the present NIC baseline design, an important consideration in determining stability and corresponding potential for shell breakup and mix. Applicant performed two-dimensional (2D) design optimizations with the LASNEX radiation-hydrodynamics implosion code on the prospective NIP hybrid platform shown in FIG. 1A. Applicant first optimizes the implosion dynamics of the indirect-drive segment to maximize the areal density of the stagnated fuel layer by tuning the laser drive pulse shape in the hohlraum. This stagnated fuel shell then acts as a tamp mass for the ignition hotspot initiated on the other side. The indirect-drive assembly side requires a one-sided drive energy of 0.71 MJ at a peak power of 157 TW, resulting in a hohlraum peak radiation temperature of 248 eV. (This should be compared with the present NIC target that requires a two-sided drive of ˜1.8 MJ at 500 TW and ˜300 eV). Applicant then optimizes the foot and main drive of the laser pulse shape on the direct-drive ignition side to maximize the areal density of that stagnating shell segment, followed by tuning of the launch time of the ignition shock pulse to maximize the temperature in the ignition hotspot. The radial direct-drive shock ignition side requires a one-sided drive of ˜0.53 MJ at 231 TW (This should be compared with prospective polar-direct-drive shock ignition targets for NIF that require two-sided (polar) drives of ˜0.7-1 MJ at ˜350-450 TW). Finally, because the radial-direct-drive is more efficient that the indirect-drive compression, synchronizing the two segments to optimize the 2D burn dynamics—that is, maximize the fusion yield and energy gain—requires that the start time of the former drive pulse must be delayed by ˜8.2 ns after the start of the indirect-drive side. Under the above burn optimization, this target achieves a 2D fusion yield of 39 MJ and thus a gain of ˜32. Applicant also observed a unique burn history for this platform in that the rate of production of fusion energy as a function of time exhibits a double maxima as the smaller, faster direct-drive side ignites and the burn wave then propagates into the main fuel mass on the other side. With conventional single shell targets—indirect or direct-drive, fast compression ignition or shock ignition—only a single narrow maximum is observed in simulations. The primary critical issues for this target that may ultimately determine ignition and fusion yield performance include (1) Stability of the two fuel segments during inflight convergence and late time stagnation (Such stability considerations attend all classes of inertial confinement fusion targets). (2) The achievement of adequate implosion symmetry in a one-sided hohlraum during the assembly phase of the indirect-drive side (Although Applicant notes that, unlike conventional indirect-drive, Applicant is not seeking high velocity nor ignition here, rather just the attainment of a reasonable stagnation areal density to tamp the hotspot formed on the other side) (3) Minimizing high-atomic-number mix into the hotspot from the gold guide cone that separates the two fuel segments. A low-z (e.g., C) anti-mix surface coating is applied to the cone to ameliorate this. (Recent simulations of an analogous guide cone for a fast ignition target indicates that the drag on the sliding DT plasma at the Au/DT interface tends to leave the Au mix behind). Future simulations and proof of principle experiments will determine the importance of these issues. FIG. 2 illustrates a fill tube used with the indirect-dive/direct drive target for inertial confinement fusion of the present invention. The following numbered components are illustrated in FIG. 2: 200 Half section of target assembly 100 102 Gold hohlraum 20 um thick 104 Hohlraum aluminum backing 108 Indirect-drive fuel segment 202 Outer layer CH plastic+2% Si 204 Inner layer solid deuterium-tritium (D-T) 50:50 atomic fraction 210 um thick 206 Gas filled volume D-T gas 0.0003 glcm3 208 Outer layer CH plastic (no Si) 15 um thick 210 Inner layer solid deuterium-tritium (D-T) 50:50 atomic fraction 350 μm thick 106 Gold support cone 212 Fill tube FIG. 2 is a half section 200 of the target assembly 100. The indirect-drive fuel segment 108 is made up of two layers, an outer layer 202 of CH plastic+2% Si and an inner layer 204 of solid Deuterium-tritium (D-T) 50:50 atomic fraction 204 um thick. The direct-drive fuel segment 110 has two layers, the outer layer 208 is CH plastic (no Si) and the inner layer 210 is solid Deuterium-tritium (D-T) 50:50 atomic fraction and is 350 um thick. The hollow volume 206 of the two fuel segments 108 and 110 is filled with D-T gas 0.003 glcm3. As shown the two fuel segments 108 and 110 are mounted on the cone section 106 of the hohlraum. FIG. 3 illustrates an indirect-drive/direct drive target assembly with alternate fuel layers. The following numbered components are illustrated in FIG. 3: 300 Hybrid indirect-drive/direct drive target assembly with alternate fuel layers 102 Gold hohlraum 104 Gold hohlraum aluminum backup 106 Guide cone 108 Indirect-drive fuel segment 110 direct-drive fuel segment 112 Center line 202 Outer layer CH plastic+2% Si 304 Inner layer low density (0.025 glcm3) CH foam (aerogel) filled with liquid D-T 206 Inner gas volume 208 Outer layer CH plastic (no Si) 310 Inner layer low density (0.025 glcm3) foam (aerogel) filled with liquid D-T Although the description above contains many details and specifics, these should not be construed as limiting the scope of the invention but as merely providing illustrations of some of the presently preferred embodiments of this invention. Other implementations, enhancements and variations can be made based on what is described and illustrated in this patent document. The features of the embodiments described herein may be combined in all possible combinations of methods, apparatus, modules, systems, and computer program products. Certain features that are described in this patent document in the context of separate embodiments can also be implemented in combination in a single embodiment. Conversely, various features that are described in the context of a single embodiment can also be implemented in multiple embodiments separately or in any suitable subcombination. Moreover, although features may be described above as acting in certain combinations and even initially claimed as such, one or more features from a claimed combination can in some cases be excised from the combination, and the claimed combination may be directed to a subcombination or variation of a subcombination. Similarly, while operations are depicted in the drawings in a particular order, this should not be understood as requiring that such operations be performed in the particular order shown or in sequential order, or that all illustrated operations be performed, to achieve desirable results. Moreover, the separation of various system components in the embodiments described above should not be understood as requiring such separation in all embodiments. Therefore, it will be appreciated that the scope of the present invention fully encompasses other embodiments which may become obvious to those skilled in the art. In the claims, reference to an element in the singular is not intended to mean “one and only one” unless explicitly so stated, but rather “one or more.” All structural and functional equivalents to the elements of the above-described preferred embodiment that are known to those of ordinary skill in the art are expressly incorporated herein by reference and are intended to be encompassed by the present claims. Moreover, it is not necessary for a device to address each and every problem sought to be solved by the present invention, for it to be encompassed by the present claims. Furthermore, no element or component in the present disclosure is intended to be dedicated to the public regardless of whether the element or component is explicitly recited in the claims. No claim element herein is to be construed under the provisions of 35 U.S.C. 112, sixth paragraph, unless the element is expressly recited using the phrase “means for.” While the invention may be susceptible to various modifications and alternative forms, specific embodiments have been shown by way of example in the drawings and have been described in detail herein. However, it should be understood that the invention is not intended to be limited to the particular forms disclosed. Rather, the invention is to cover all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the following appended claims. |
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claims | 1. A pattern drawing method for drawing a desired pattern on a base material by irradiating an electronic beam onto the base material, said method comprising:scanning a first region of the base material with the electronic beam in a first dose amount;scanning a boundary region continuous with the first region with the electronic beam alternately in the first dose amount and in a second dose amount; andscanning a second region continuous with the boundary region with the electronic beam in the second dose amount. 2. The pattern drawing method of claim 1, wherein at least one surface of the base material comprises a curved surface having groove sections formed thereon with a predetermined pitch;wherein each groove section includes a side wall and an inclined section extending from a top of the side wall to a bottom of a side wall of an adjacent groove section; andwherein the inclined section comprises at least one of each of the first region, the boundary region and the second region. 3. The pattern drawing method of claim 1, wherein a difference between the fist dose amount and the second dose amount is a minimum adjustable dose amount based on a minimum clock of a D/A converter for driving an electronic beam gun for irradiating the electronic beam. 4. The pattern drawing method of claim 1, wherein in the boundary region, areas scanned with the electronic beam in the first dose amount are alternated in a sub scanning direction with areas scanned with the electronic beam in the second dose amount. 5. The pattern drawing method of claim 1, wherein in the boundary region, each of a number of areas scanned with the electronic beam in the first dose amount and a number of areas scanned with the electronic beam in the second dose amount is adjusted in accordance with a desired inclination angle. 6. The pattern drawing method of claim 1, wherein in the boundary region, a duty ratio of a first area scanned with the electronic beam in the first dose amount and a second area scanned with the electronic beam in the second dose amount within one pitch of the first area is adjusted in accordance with a desired inclination angle. 7. The pattern drawing method of claim 1, wherein in the boundary region, both (i) each of a number of first areas scanned with the electronic beam in the first dose amount and a number of second areas scanned with the electronic beam in the second dose amount, and (ii) a duty ratio of a first area scanned with the electronic beam in the first dose amount and a second area scanned with the electronic beam in the second dose amount within one pitch of the first area, are adjusted in accordance with a desired inclination angle. 8. The pattern drawing method of claim 1, wherein in the boundary region, areas scanned with the electronic beam in the first dose amount are alternated in a main scanning direction with areas scanned with the electronic bean in the second dose amount. 9. The pattern drawing method of claim 8, wherein in the boundary region, positions of the areas scanned with the electronic beam in the second dose amount in the main scanning line are staggered with respect to positions of areas scanned with the electronic beam In the second dose amount in adjacent main scanning lines. 10. The pattern drawing method of claim 8, wherein in the boundary region, the areas scanned with the electronic beam in the first dose amount are alternated on a main scanning line with the areas scanned with the electronic beam in the second dose amount. 11. The pattern drawing method of claim 10, wherein in the boundary region, positions of the areas scanned with the electronic beam in the second dose amount in the main scanning line almost align with positions of areas scanned with the electronic beam in the second dose amount in adjacent main scanning lines. 12. The pattern drawing method of claim 10, wherein in the boundary region, positions of the areas scanned with the electronic beam in the second dose amount in the main scanning line are different from positions of areas scanned with the electronic beam in the second dose amount in adjacent main scanning lines. 13. The pattern drawing method of claim 1, wherein in the boundary region, widths of areas scanned with the electronic beam in the first dose amount and widths of areas scanned with the electronic beam in the second dose amount are smaller than a diameter of the electronic beam. 14. A pattern drawing apparatus, comprising:a drawing section to draw a desired pattern on a base material by irradiating an electronic beam to the base material;a storing section to store dose amount information including a first dose amount and a second dose amount;a control section to control the drawing section based on the dose amount distribution information to scan a first region of the base material with the electronic beam in the first dose amount, to scan a boundary region continuous with the first region with the electronic beam alternately in the first dose amount and the second dose amount, and to scan a second region continuous with the boundary region with the electronic beam in the second dose amount. 15. The pattern drawing apparatus of claim 14, wherein the storing section stores information corresponding to a shape of the base material; andwherein the pattern drawing apparatus further comprises a calculating section to calculate the dose amount distribution information for the boundary region based on the information corresponding to the shape of the base material. 16. The pattern drawing apparatus of claim 14, wherein the drawing section comprises:an electronic beam emitting section to emit an electronic beam;an electronic lens to adjust a focus point of the electronic beam;a stand to support the base material; anda measuring section to measure a drawing position on the base material;wherein the control section controls the focus point of the electronic beam in accordance with the drawing position by adjusting an electric current of the electronic lens. 17. The pattern drawing apparatus of claim 14, wherein a difference between the fist dose amount and the second dose amount is a minimum adjustable dose amount based on a minimum clock of a D/A converter for driving an electronic beam gun for irradiating the electronic beam. |
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description | The present application claims priority from Japanese application JP 2006-027861 filed on Feb. 6, 2006, the content of which is hereby incorporated by reference into this application. The present invention relates to a method and an apparatus for inspecting an electrical defect of a microstructure circuit formed on a semiconductor wafer. As a method for detecting a defect of a circuit pattern formed on a wafer by comparing images in a manufacturing process of a semiconductor device, for example, a pattern comparing inspection method is disclosed in JP-A No. 258703/1993. The method uses an SEM in which, a point focused electron beam is scanned. The SEM type inspection apparatus is higher in resolution than optical inspection systems and it has a feature for enabling such an electrical defect as a connecting failure to be detected. However, because the SEM type inspection apparatus scans an electron beam on a specimen surface two-dimensionally to obtain an image, the scanning time is long. This disadvantage will become a substantial obstacle for reducing the inspection time. As an electron beam inspection method that has successfully reduced the inspection time, for example, the JP-A No. 249393/1995 discloses a projection type inspection apparatus, which radiates a rectangular electron beam onto semiconductor wafer and forms an image of buck scattering and secondary electrons with use of electron lenses. The projection type inspection apparatus can radiate an electron beam with a larger current than that of the SEM type at a time, thereby obtaining a plurality of images collectively. The projection type is thus expected to form images more quickly than those of the SEM type, that is, the scanning electron type. On the other hand, a secondary electron mapping type inspection apparatus cannot obtain a sufficient resolution due to the aberration of the objective lens, thereby it is difficult to obtain a required defect detection sensitivity. The JP-A No. 108864/1999 points out such disadvantages of the apparatus. The JP-A No. 108864/1999 discloses a mirror electron imaging type wafer inspection apparatus that uses electrons pulled back before colliding with a specimen due to a negative electric field formed just above the wafer (hereunder, to be referred to as mirror electrons or mirror reflecting electrons) as imaging electrons. Here, the mirror electron imaging type wafer inspection apparatus will be described. The mirror electron imaging type wafer inspection apparatus obtains an image to be used for inspection with use of a mirror electron microscope. An inspection image is obtained by radiating an electron beam onto a specimen and forms an image of the reflecting electron beam. At this time, a negative potential is applied onto the surface of the specimen in advance so that the radiated electron beam is reflected on a specific equipotential surface in the vicinity of the specimen surface without reaching the specimen surface. The electrons reflected on an equipotential surface in the vicinity of the specimen surface such way are referred to as “mirror electrons”. Because the equipotential surface of the specimen surface is influenced by the information of an unevenness and a potential change of the specimen surface itself, the image to be obtained is also influenced by the information of the unevenness and the potential change of the specimen surface when the mirror electrons are imaged. Consequently, shape and electrical defects of the specimen surface can be detected by comparing such a mirror electron image with a reference image, respectively. As described above, the mirror electron imaging type wafer inspection apparatus differs completely from any of the conventional SEM type inspection apparatuses. Consequently, to put the mirror electron imaging type wafer inspection apparatus for practical use, it is needed to think out a method for setting inspection conditions optimized for the apparatus. Under such circumstances, it is an object of the present invention to realize an inspection condition setting method optimized for the object mirror electron imaging type wafer inspection apparatus and make it easier to operate the apparatus. Upon thinking out a method for setting such inspection conditions optimized for the mirror electron imaging type wafer inspection apparatus, the present inventor has examined the following circumstances. In a semiconductor device manufacturing line, a mirror electron imaging type wafer inspection apparatus is often used for defect inspection in all or some specific portions of every wafer flowing on the manufacturing line. Thus, it should be carefully determined between which semiconductor processes an inspection process is to be inserted allocated for the inspection process by giving consideration to the productivity of the semiconductor manufacturing line. In other words, if accurate inspection process is designed to improve the yield, the productivity is lowered in proportion to an increase of the inspection time. On the other hand, if the inspection time is reduced, both the inspection accuracy and the productivity are lowered. As such, for various manufacturing lines, each of which has different defect generating rate, defect generating process, and productivity, there is an optimum inspection time for each manufacturing line to achieve the best productivity. The optimum inspection time for each manufacturing line is varied among those manufacturing lines. This is why the conditions for the inspection speed should be set flexibly for the mirror electron imaging type wafer inspection apparatus so as to make inspections most efficiently by giving consideration to the circumstances specific to each of various semiconductor device manufacturing lines. The inspection speed of the mirror electron imaging type wafer inspection apparatus means an area of a wafer that can be inspected per unit time. FIG. 2 shows the arrangement of pixels for composing an inspection object image. In FIG. 2, each cell means a pixel 201. The call is usually a square of which the length of this side is represented by D. An inspection image of the mirror electron imaging type wafer inspection apparatus is obtained with use of a time delay integration data acquisition method (TDI method). In the TDI method, integration is made by sending image signals in the vertical direction of the image synchronously with the movement of the wafer (as shown with a white arrow in FIG. 2). The cycle in which one signals of pixel are sent in the vertical direction is defined as P. And the length in a direction (horizontal direction in FIG. 2) normal to the movement of the wafer in the image region is defined as L. The image data of length L×width D area (gray region shown in FIG. 2) is sent in a cycle P to an image processing apparatus. Consequently, the inspection speed S can be described by using D, P, and L in the following expression:S=D×L×P. To operate the apparatus at an optimized speed, therefore, the user is requested to satisfy the relationship among D, L, and P shown above and adjust the D, L, and P values so as not to degrade the inspection sensitivity. Particularly, in the case of the mirror electron imaging type wafer inspection apparatus, the pixel size optimized for inspection is changed depending on the magnification of the imaging optical system for mirror reflecting electrons. This change depends on the characteristics such as the material, structure, etc. of the specimen. And such characteristics are never generated in any of SEM and secondary electron projection type electron optical systems; the characteristics are specific to the mirror optical systems. The user of the apparatus, therefore, is requested to adjust the D, L, and P values by giving consideration to the magnification of the optical system. Conventionally, the apparatus manager and the apparatus developer have set such D, P, and L values by trial and error, by giving consideration to the characteristics of the mirror electron imaging type wafer inspection apparatus and the inspection object. Furthermore, the user's interest is usually just the inspection speed. The user would thus feel very troublesome when requested to set such conditions and will feel that the apparatus is inconvenient when operating the apparatus. In order to solve the above described conventional problems, therefore, the present invention has enabled such S, D, L, and P values to be displayed on an operation screen so that the user can examine such conditions as inspection speed, inspection sensitivity, etc. intuitively. Furthermore, the present invention has provided a process newly for converting user determined conditions to conditions for operating an electron optical system, a time delay integration type imaging device, and a wafer moving stage respectively so that the user can make inspections in accordance with the circumstances of various semiconductor manufacturing lines without understanding the details of the inspection apparatus. According to the present invention, therefore, it is possible to set such conditions as optimized pixel size, irradiated area size, etc. to easily realize an inspection speed capable of preventing an semiconductor device manufacturing line from delay so that the user can inspect defects of each semiconductor pattern efficiently under optimized conditions. Because such inspection conditions can be set easily such way, the total inspection time from condition setting to end of inspection can be reduced. And because the apparatus can be operated easily, the apparatus will also have advantages in sales policy. Hereunder, a description will be made in detail for a configuration of a mirror electron imaging type wafer inspection apparatus in a preferred embodiment of the present invention with reference to the accompanying drawings. FIG. 1 shows an example of a hardware configuration of the mirror electron imaging type wafer inspection apparatus in the preferred embodiment of the present invention. In FIG. 1, vacuum pumps, their controller, pipes for evacuating systems, etc. are omitted. At first, main elements of the electron optical system of the apparatus will be described. An electron beam 100a emitted from an electron gun 101 is focused by a condenser lens 102 and deflected by an ExB deflector 103 to form a cross-over 100b, then radiated onto a specimen wafer 104 as an approximate parallel flux. In FIG. 1, although only one condenser lens 102 is used, a plurality of lenses may be combined into a lens system to optimize the optical conditions. The electron gun 101 is usually a Zr/0/W type Schottky electron source. Such voltages and currents as an extracting voltage applied to the electron gun 101, an accelerating voltage to extracted electrons, a heating current of an electron source filament, etc. required for operating the apparatus are supplied and controlled by an electron gun controller 105. The ExB deflector 103 is disposed in the vicinity of an imaging plane 100d of an imaging electron beam 100c. At this time, an aberration occurs in the radiated electron beam 100a due to the ExB deflector 103. If this aberration must be corrected, another ExB deflector 106 for correcting the aberration is disposed between an radiation system condenser lens 102 and the ExB deflector 103. The electron beam 100a deflected by the ExB deflector 103 so as to go along an axis perpendicular to the wafer 104 is formed by an objective lens 107 as a planar electron beam entered in a direction perpendicular to the surface of the specimen wafer 104. On the focal plane of the objective lens 107 is formed a finer cross-over by the radiation system condenser lens 102. Thus the electron beam can be radiated onto the specimen wafer 104 just in parallel. A region of the specimen wafer 104, in which the electron beam 100a is radiated, is an area as large as, for example, 2500 μm2, 10000 μm2, or the like. The specimen wafer 104 mounted on a wafer stage 108 receives a negative voltage almost equal to or slightly higher larger absolute value than the accelerating voltage of the electron beam. This negative potential works on the radiated electron beam 100a so that it slows down just before reaching the wafer 104 and pulled back upward to become as reflecting mirror electrons, thereby it does not collide the wafer 104 in most cases. The voltage applied to the wafer 104 is supplied and controlled by a wafer voltage controller 109. In order to make the radiated electrons reflected in the vicinity of the wafer 104, a difference from the accelerating voltage of the radiated electron beam 100a is required to be adjusted accurately and the wafer voltage controller 109 and the electron gun controller 105 are controlled so that they are interlocked with each other. Mirror electrons flying from the wafer side includes information related to an electrical defect of an object circuit pattern formed on the wafer 104. Thus its image is formed with use of an electron imaging optical system to be fetched in the apparatus as an image for determining whether there is a defect in the pattern or not. The mirror electrons are focused by the objective lens 107. And the ExB deflector 103 is controlled so as not to deflect an electron beam advancing from below, so that the mirror electrons go up perpendicularly as are, then magnified and projected by an intermediate lens 110 and a projection lens 111 at an image detection part 112. In FIG. 1, only one projection lens 111 is used, but a plurality of lenses may be composed into a projection system to provide a higher magnification and correct distortions of images. The image detection part 112 converts an image to an electrical signal and sends a distribution of the local charging potential of the surface of the wafer 104, that is, a defect image to an image processing part 112. The electron optical system is controlled by an electron optical system controller 113. Next, the image detection part 112 will be described. A fluorescent plate 112a, an optical image detector 112b, and an optical image transmission system 112c are used for optical coupling to convert a mirror electron image to an optical image and detect the image. In this embodiment, an optical fiber bundle is used as the optical image transmission system 112c. The optical fiber bundle consists of the same number of thin optical fibers as the number of pixels and it can transmit optical images efficiently. In case where a fluorescent image is used with a sufficient light, the optical transmittance may be set lower. In such cases, instead of the optical fiber bundle, an optical lens is used and an optical image on the fluorescent plate 112a is formed by the optical lens on a light detecting surface of the optical image detector 112b. Furthermore, an amplifier is inserted in the optical image transmission system to transmit an optical image with a sufficient light. The optical image detector 112b converts an optical image formed on the light receiving plane to an electrical image signal and outputs the signal. As the optical image detector 112b, an TDI sensor is used. The TDI sensor uses a time delay integration (TDI) type CCD. The image processing part 116 is composed of an image memory 116a and a defect determination part 116b. The image memory 116a inputs electron optical condition, image data, and stage position data from the electron optical system controller 113, the image detection part 112, and the stage controller 115 respectively and stores the image data by relating it to the coordinate system used on the specimen wafer. The defect determination part 116b uses image data related to the coordinates on the wafer and compares it with a preset value or with an adjacent pattern image or an image of the same pattern position in an adjacent die, or the like with use of various defect determination methods so as to determine a defect. The defect coordinates and an intensity of its corresponding pixel signal are transferred to and stored in the inspection apparatus controller 117. The user sets any one of those defect determination methods or the inspection apparatus controller 117 selects a method corresponding to the object wafer type in advance. The inspection apparatus controller 117 inputs/outputs conditions for operating each part of the apparatus. The inspection apparatus controller 117 inputs beforehand various preset conditions such as electron beam accelerating voltage, current conditions for electron optical devices, wafer stage moving speed, timing for acquisition an image signal from an image detection element. The inspection apparatus controller 117 controls the controller of each element as an interface with the user. The inspection apparatus controller 117 may be composed of a plurality of computing devices connected to each another through a communication line and having a specific function. The apparatus further includes user interface device 118 with a monitor. In the mirror electron imaging type wafer inspection apparatus, the electron beam hardly collides with the object wafer. Thus the specimen wafer may not be charged sufficiently in some cases. To detect an electrical defect, however, the wafer must be charged sufficiently to cause a difference from that of normal parts. The present invention has solved this problem by providing pre-charging devices 119a and 119b. Those devices are controlled by a pre-charging controller 120. The pre-charging controller 120 controls a charging potential generated on the wafer by the pre-charging devices 119a and 119b with the wafer voltage controller 109 and the electron gun controller 105 not to disturb the status of the electron beam that is reflected in the vicinity of the wafer surface. FIG. 3 is a first embodiment of a screen on which the user operates the inspection apparatus. This screen 301 is an “inspection condition setting screen” on which the user select an inspection speed and an inspection sensitivity or part of the screen. The screen 301 belongs to the user interface device with a monitor 118. A graph 302 displayed on the screen 301 has a horizontal axis that indicates a pixel size D and a vertical axis that indicates an inspection speed S. In the graph 302, the inspection speed S means an area on a wafer surface to be inspected per unit time and it is represented by an inspection area (cm2/h) per hour. The user may select this unit to make it easier to understand. For example, the unit may be the number of wafers to be processed per hour or an inverse number to define a processing time of one wafer and a time required for a unit area inspection (e.g., h/cm2). In the graph 302, a range from 0 cm2/h to 600 cm2/h is shown. The pixel size D is represented by a value corresponding to an actual size on the object wafer. It is within a range from 0 nm to 250 nm. The graph 302 also has a plurality of characteristic straight lines 303. These straight lines are used for different cycle P values of the TDI sensor respectively. The graph 302 uses values of 100 to 700 kHz selected as P values. The width L of an inspection image is displayed on an inspection image width display part 304. In this example, it is set as 60 μm. A plurality of conditions are displayed as a pull-down menu for this value when the user clicks the selection arrow 305. The user can select and change any of the conditions. If the user selects another value, a newly calculated straight line is displayed as shown in FIG. 4. FIG. 4 shows an example in which 120 μm is selected as an L value. The L value has its upper limit, which depends on the aberration of the objective lens. If the value goes over 200 μm, the distortion and the resolution degradation in marginal area of the field of view advance significantly. Consequently, the upper limit of the field of view usable as an inspection image is about 200 μm×200 μm. The user can search an inspection speed and a pixel size by moving a white arrow pointer with a mouse on the graph 302. The values of the inspection speed, pixel size, TDI cycle calculated from the position of the pointer on the graph 302 are displayed in a display field 307 at the bottom of the screen. The user can select conditions with reference to such concrete values. In the graph 302, not only the values on straight lines, but also values between straight lines are calculated from the pointer position and displayed. When conditions are determined on the graph 302, the user clicks the mouse button (not shown) or press a specific key on the keyboard (not shown) to fix the conditions. Those conditions are sent to the controller of the inspection apparatus when the user clicks the [ENTER] button 308 on the screen, then those conditions are converted to detailed operation conditions of the apparatus. Each of those conditions has its upper limit, which depends on the specifications of the apparatus. For example, if P=700 kHz is the upper limit in the specifications of the TDI camera and a condition is set in a region over 700 kHz, the condition is ignored. The user can determine values of a set of the pixel size, inspection image width, and TDI sensor operation cycle with reference to the drawing. Using the condition setting method in this embodiment makes it possible for the user to set an inspection speed of the mirror electron imaging type wafer inspection apparatus without trial and error. In the first embodiment, the user determines conditions for operating the inspection apparatus with reference to mainly the values of inspection speed and pixel size and the relationship between the defect detection sensitivity and the pixel size is not clear. In this second embodiment, therefore, the horizontal axis of the graph displayed on the inspection speed S setting screen is used for defect detection sensitivity, thereby the user comes to know the relationship between the defect detection sensitivity and the pixel size intuitively. Instead of the horizontal axis, the vertical axis may also be used for the defect detection sensitivity. FIG. 5 shows a schematic diagram of the inspection speed S setting screen displayed on the user interface device with a monitor 118 of the mirror electron imaging type wafer inspection apparatus. The user operation screen shown in FIG. 5 is displayed on the monitor of the mirror electron imaging type wafer inspection apparatus or mirror electron imaging type specimen inspection apparatus. Unlike that shown in FIG. 3, the horizontal axis of the graph 501 is used for the defect detection sensitivity. Explanations of the same items such as the pointer, the characteristic curve, etc. as those shown in FIG. 3 will be omitted to simplify the description. The relationship between the pixel size and the defect detection sensitivity is based on the magnification function of defect images specific to the mirror electron imaging type wafer inspection apparatus. An inspection object image of the mirror electron imaging type wafer inspection apparatus is obtained by imaging a distortion of an equipotential surface caused by existence of a defect. FIG. 6 is a diagram for describing principles of such mirror electron imaging. FIG. 6A shows a view of an equipotential surface 603 and a view of a trajectory 604 of radiated electrons which are reflected from the equipotential surface 603 when a protruded defect 601 such as a foreign particle and a recessed defect 602 such as a scratch are detected on the object wafer surface. FIGS. 6B and 6C are diagrams for showing a distortion of the equipotential surface 606 and a trajectory 607 of radiated electrons when connecting failure 605a and 605b exist as defects in vias 605 used for connecting to the lower layer wiring embedded in an oxide film respectively. In FIG. 6B, the electrically open via 605a is negatively charged. In FIG. 6C, the open via 605b is positively charged. Due to any of the unevenness of the shape of the wafer surface and the electrical difference on the wafer surface, the area of the distortion of the equipotential surface is larger than the actual size of the defect. In addition, when the equipotential surface is positioned higher, the area of the distortion is larger while the distortion level is low. Consequently, a larger image is obtained when compared with the actual defect size by adjusting the mirror electron imaging optical system. This means that defects can be detected even when defect images are obtained with pixels larger than the actual defect size. FIG. 7 is an example of an inspection image obtained with use of a mirror electron imaging method. FIG. 7A shows a schematic diagram of a circuit pattern. This pattern consists of 200 nm×200 nm square vias 702 embedded in an oxide film 701 and composed like a matrix patterns arranged in 5 rows×5 columns at pitches of 800 nm. Each normal via is continued to a wiring 703 in the lower layer. FIG. 7B is an inspection image, that is, a mirror electron image obtained with use of a mirror electron imaging method. The size of the via patterns in this mirror electron image, except for the center one, is as large as 600 nm, which is about 3 times the actual one. This magnification is made due to a distortion of the equipotential surface caused by a difference of voltage between the via voltage and the voltage of its peripheral insulation film. In the mirror electron image shown in FIG. 7B, the center via 704 has a disconnect defect and its voltage differs from that of other normal ones by about +1.5V. The size of the mirror electron image of this defect via 704 is about 1200 nm, which is about double that of a normal via and magnified up to 6 times that of the actual via pattern. It can thus be concluded from those data that the size of a mirror electron image is magnified to from 3 times to 6 times the actual size due to the via voltage. According to this result, in this embodiment, it is expected to be able to detect defects of patterns up to ⅓ of the pixel size and the horizontal axis of the graph 501 for the inspection speed is displayed for the detection sensitivity, which is ⅓ of the pixel. The value indicated by the horizontal axis of the graph 501 shown in FIG. 5 is converted from the value indicated by the horizontal axis of the graph 501 shown in FIG. 3 on the basis of the above detection sensitivity information. The calculation for converting a value of the horizontal axis or vertical axis such way is executed by a computing device built in the inspection apparatus controller 117 or user interface device with a monitor 118. In the same way, the relationship between the detection sensitivity and the pixel is stored in the memory means built in the inspection apparatus controller 117 or in the user interface device with a monitor 118. As the memory means, for example, any of a memory, a hard disk, etc. may be used. Next, a description will be made for how to transmit such conditions as user specified inspection speed, etc. to the electron optical system and the wafer stage. FIG. 8 shows a schematic diagram of a hardware configuration of a mirror electron imaging type wafer inspection apparatus in this embodiment. In FIG. 8, the same reference numbers will be used for the components of the same functions and operations as those shown in FIG. 1. In the mirror electron imaging type wafer inspection apparatus shown in FIG. 8, the inspection apparatus controller 117 is provided with a conversion part 801. Conditions for inspection operations inputted by the user through the user interface device 118 are sent to the inspection apparatus controller 117. In this embodiment, the inspection apparatus controller 117 is provided with a condition conversion part 801. Conditions inputted through the user interface device with a monitor 118 are values of the pixel size D, inspection speed S, TDI camera image acquisition cycle P, and size of the field of view L. The moving speed Vs of the wafer stage 108 is calculated on the basis of those values. The Vs is determined by the following relational expression according to the TDI camera image acquisition cycle P and the pixel size D.Vs=P×D This Vs value is sent to the stage controller 115. The stage controller 115 controls a stage driving mechanism by monitoring the stage position information received from the position detector 114 so as to keep the speed Vs while the stage is moving. The TDI camera image acquisition cycle P is sent to the condition conversion part 801 as is. The condition conversion part 801 controls the TDI camera image acquisition cycle so that the image acquisition is synchronized with the stage movement. The pixel size D, as well as a preset pixel size Dp on the TDI camera light detecting surface are used to calculate a magnification Dp/D of the imaging electron optical system. The magnification Dp/D of the imaging electron optical system is sent to the electron optical system controller 113 and used to control the voltage and the electromagnet current applied to the objective lens 107, the intermediate lens 110, and the projection lens 111 respectively. The voltage and current conditions of each electron optical element of the imaging optical system with respect to the magnification of the imaging optical system are stored as a numerical table beforehand in the electron optical system controller 113 or condition conversion part 801. The voltage and current conditions are determined with reference to those values in the table. If a magnification value that is not stored in the table is referred to, current and voltage values are determined by interpolating the values for the nearest magnification value. The condition conversion part 801 or electron optical system controller 113 stores a numerical table that records conditions of both the condenser lens 102 and the objective lens 107 with respect to the size of the field of view L. And upon a user's determination for the size of the field of view L, the voltage and current values of the radiated electron optical system are referred to from the numerical table for controlling. With such a configuration, the user selected inspection conditions are reflected correctly in the inspection apparatus. As described above, therefore, in this embodiment, the user can set such inspection conditions as inspection speed and defect detection sensitivity as parameters. Consequently, the user can operate the apparatus more easily. In the second embodiment, a description is made for a user operation screen on which the defect magnification is set as 3 times. This third embodiment enables the user to change the defect magnification. FIG. 9 shows a user operation screen in this third embodiment. The user operation screen shown in FIG. 9 is displayed on the monitor of the mirror electron imaging type wafer inspection apparatus or mirror electron imaging type specimen inspection apparatus. In this third embodiment, the description for the same components as those shown in FIG. 5 will be omitted. In FIG. 9, there is only a difference from that shown in FIG. 5; a defect magnification selection field 901 is provided. The user can select a magnification from a plurality of defect magnification values by clicking the arrow in the defect magnification selection field 901. The value of the horizontal axis of the graph 902 is corrected by the selected defect magnification, thereby the displayed characteristic curve is also corrected. The user can change the defect magnification according to the objective lens focal condition, the height of the equipotential surface for reflecting mirror electrons, etc. in the mirror electron imaging. A mirror electron image is always formed due to a distortion of the equipotential surface even when the defect type is an uneven surface or a voltage variation caused by an electrical defect. Consequently, the user can estimate a defect image magnification in advance by using the height of the equipotential surface for causing mirror reflection of electrons and focal conditions of the objective lens as parameters. For such an estimate, the user is just requested to obtain a mirror electron image with respect to a different focal point of the objective lens and a different voltage potential value of the wafer and measure a magnification according to the actual defect size by using an Si wafer of which size is already known and having a protruded or recessed shape defect that is already processed. If the defect type is not such a shape defect, but it is a potential variation, a voltage that causes the equipotential surface to be distorted as much as a distortion generated by a protruded or recessed shape may be calculated by computer simulation and adjusted precisely with the relationship between the electron optical condition and the magnification in the shape defect. Because such calculation of a level of a distortion of an equipotential surface with respect to a voltage is simple calculation of an electrical field, it is so easy. Instead of using a standard specimen as described above, it is also possible to analyze a trajectory of electrons by computer simulation and change the condition of the objective lens, thereby calculating an image to be obtained, then obtaining a relationship with a magnification. In this embodiment, the inspection apparatus is provided with an inspection condition evaluation device 1001 for holding a table that stores a condition of the objective lens, a negative voltage value to be applied to each wafer to change its equipotential surface used for mirror reflection, and a defect magnification obtained as described above. FIG. 10 shows a schematic diagram of a system provided newly with the inspection condition evaluation device 1001. When an inspection is made with a defect magnification using this inspection apparatus, the equipotential surface for reflecting mirror electrons must be kept constant. Thus it is important to keep the wafer surface potential constant. This is why pre-charging devices 119a and 119b are used. Those charging devices are controlled by a pre-charging controller 120. For example, assume now that a wafer is passed under the pre-charging device, then just under the objective lens and moved just under the pre-charging device 119b. In such a case, the wafer surface potential is set to a prescribed potential by the pre-charging device 119a. This potential makes it possible to obtain a desired defect magnification. The potential is given from the numerical table of the inspection condition evaluation device 1001 and controlled by the pre-charging controller 120. As the pre-charging device, for example, such an electron beam radiation device as a flood gun may be used. After the wafer passes just under the objective lens, the disturbance of the equipotential surface potential, caused by slight charging of the wafer when in observation of mirror electrons, must be eliminated so to as return the potential to a required level with use of the pre-charging device 119b again. According to this third embodiment, therefore, it is possible to optimize the such conditions as the inspection time including the user specified defect magnification, thereby the object semiconductor manufacturing line can be managed efficiently. Although a description has been made for the preferred embodiments of the present invention, the present invention also includes a combination of the first to the third embodiment described above. |
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claims | 1. A wafer defect inspection method, comprising the steps of:radiating an electron beam on a certain area of a wafer;reflecting radiated electrons by applying a negative voltage to said wafer just before said radiated electrons reach a surface of said wafer;forming an image of the reflected electrons;moving said wafer at a certain speed with respect to said radiated electron beam;obtaining a digital image with a time delay integration data acquisition synchronously with a moving speed of said wafer; andextracting a defect of said wafer using an obtained image and displaying a position and an image of said defect,wherein said method further includes a step of displaying on an operation screen a relationship among the values of S, D, L, and P when L is defined as a length in a direction perpendicular to a moving direction of said wafer of which image is being obtained in a range of an inspection object image in the region on which said electron beam is radiated, S is defined as an area of said wafer to be inspected per unit time, D is defined as a size on said wafer, corresponding to a unit pixel of said obtained digital image, and P is defined as an image signal acquisition frequency in said time delay integration formula, as well as a step of adjusting an electron optical system and a wafer moving speed on the basis of user determined S, D, L, and P values,wherein a graph for S=D×L×P is displayed with respect to a plurality of L values and P values respectively by assuming the horizontal axis as D and the vertical axis as S, andwherein said horizontal axis of said graph is displayed as a defect sensitivity obtained by multiplying said D value by a constant less than 1. 2. The method according to claim 1, wherein the L value is 200 μm or under. 3. The method according to claim 1, wherein the D value is 250 nm or under. 4. The method according to claim 1, wherein a condition for operating an electron lens is determined to realize user determined values of D and L with reference to a numerical table that holds said condition for operating said electron lens corresponding to each of said D and L values beforehand. 5. The method according to claim 1, wherein said constant is ⅓ or under. 6. The method according to claim 1, wherein an Si wafer marked with an uneven pattern is used when said user selects said constant, thereby measuring beforehand a potential of said Si wafer and a focal point of an objective lens, as well as a ratio between the size of said marked pattern and the size of a mirror electron image so that said measurement results are set in said numerical table, then a wafer surface potential with respect to said constant is related to an objective lens focal condition according to said results in said numerical table. 7. A defect inspection apparatus, comprising:means for radiating an electron beam on a certain area of a wafer;means for reflecting radiated electrons by applying a negative voltage to said wafer just before said electrons reach a surface of said wafer;means for forming an image of the reflected electrons;means for moving said wafer at a certain speed with respect to said radiated electron beam;means for obtaining a digital image with a time delay integration formula synchronously with a moving speed of said wafer; andmeans for extracting a defect of said wafer using an obtained image and displaying a position and an image of said defect,wherein said method further includes means for displaying a relationship among values of S, D, L, and P when L is defined as a length in a direction perpendicular to a moving direction of said wafer of which image is being obtained in a range of an inspection image in the region on which said electron beam is radiated, S is defined as an area of said wafer to be inspected per unit time, D is defined as a size on said wafer, corresponding to a unit pixel of said obtained digital image, and P is defined as an image signal acquisition frequency in said time delay integration formula, as well as means for adjusting an electron optical system and a wafer moving speed on the basis of user determined values of S, D, L, and P,wherein said apparatus further includes means for displaying an graph for S=D×L×P with respect to a plurality of values of L and P respectively by assuming the horizontal axis as D and the vertical axis as S, andwherein said apparatus further includes means for displaying the horizontal axis of said graph as a defect sensitivity obtained by multiplying a D value by a constant less than 1. 8. The apparatus according to claim 7, wherein the L value is selected within 200 microns or under. 9. The method according to claim 7, wherein the D value is displayed as 250 nm or under. 10. The apparatus according to claim 7, wherein said apparatus further includes means for holding a numerical table in which a condition for operating an electron lens corresponding to each of D and L values is set beforehand and means for determining a condition for operating said electron lens to realize user determined values of D and L with reference to said numerical table. |
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description | 1. Field of the Invention The invention relates to semiconductor circuit design, and more particularly to modeling temperatures of a self-heating semiconductor device. 2. Background Description Accurate measurement of self-heating of SOI and SiGe based MOSFET devices is important because DC currents of such devices are typically depressed significantly due to self-heating. This is in contrast to CMOS circuits where transients are too rapid for significant self-heating to occur. Thus, simulation (compact) models must be adjusted to correctly account for self-heating in order to correctly predict circuit performance. In SOI technologies, this effect ranges from about 3 percent to 12 percent, while in SGOI (SiGe on SOI) these effects are expected to exceed 30 percent. Such large temperature effects in SGOI devices are due in part to the active region of the device being almost entirely surrounded by layers of material having poor thermal conductivity properties. For example, the active region of the SGOI device is SiGe, and the SiGe is arranged on top of an oxide layer. The SiGe layer has a limited length and width on top of the oxide layer, and thus forms what is referred to as a “island” on the oxide layer. During subsequent fabrication steps, the SiGe island is surrounded on its sides by an oxide, and then further covered over its top by an oxide. Thus, the SiGe island is relatively small and almost entirely surrounded by an oxide. Due to the surrounding oxide, the SiGe island has extremely limited thermal pathways by which to dissipate any heat generated in the SiGe island. The small dimensions of the SiGe island also increase the device's susceptibility to thermal effects. In particular, because the SiGe island is relatively small, it has a comparatively low thermal mass. With the low thermal mass, the SiGe island quickly responds to any heating by a device thereon. As such, the SiGe island itself fails to act as its own heat sink for the device and the device quickly heats the island up to the devices own temperature. Thus, any device fabricated on the SiGe island will be particularly influenced by its own self-heating effects. Known methods of measuring semiconductor device performance versus temperature include placing a diode proximate to the device for which the temperature will be measured, and using the diode's change in electrical performance as a function of temperature to measure the temperature at that point. However, such a method is difficult to implement because it is difficult to build such a diode close to a device to be measured to provide an accurate gauge of the active region of the device. Another method of measuring the temperature effect on the electrical characteristics of a semiconductor device includes running the device at a particular power level to heat itself, and using the device's own change in electrical characteristics as a function of temperature to determine the temperature of the device. While simple to fabricate such a temperature measurement configuration, the data produced by such a configuration is less than reliable because of various hysteresis-like effects. For example, the device's sensitivity to temperature changes, may be based on, for example, among other things, on the prior electrical history of the device. Such a sensitivity to electrical history makes determining the actual temperature of the device to be less than reliable. The invention is designed to solve one or more of the above-mentioned problems. In a first aspect of the invention, a method of measuring performance of a device includes thermally coupling a first heating device to a first sensing device, and generating heat at the first heating device. The method also includes measuring a change in at least one electrical characteristic of the first sensing device caused by the heat generated at the first heating device, and calculating a temperature of the first heating device using the measured change in the at least one electrical characteristic. In another aspect of the invention, a method of measuring performance of a device, includes thermally coupling a heating transistor to a measurement transistor at one or more predetermined distance, and calibrating the measurement transistor by measuring a particular electrical characteristic of an active region of the measurement transistor with the measurement transitory held at a known temperature. The method also includes generating heat at the heating transistor, and incrementally measuring a change in the at least one electrical characteristic of the measurement transistor caused by the heat generated at the heating transistor. The method additionally includes calculating a temperature of the heating transistor using the measured change in the at least one electrical characteristic. In another aspect of the invention, an apparatus for measuring semiconductor device temperature, includes a silicon island, and at least one pair of transistors, each pair of the at least one pair of transistors comprises a transistor configured to generate heat and a transistor configured to sense temperature, wherein each transistor of each pair of transistors is arranged a prescribed distance from its corresponding transistor. In another aspect of the invention, an apparatus for measuring semiconductor device temperature, includes at least one silicon island, and at least one heating field effect transistor configurable to generate heat arranged within the silicon island. The apparatus also includes at least one sensing field effect transistor arranged within the at least one silicon island corresponding to each heating field effect transistor of the at least one heating field effect transistor, wherein each sensing field effect transistor is arranged a prescribed distance from its corresponding heating field effect transistor and each sensing field effect transistor is configurable to sense a temperature. The apparatus additionally includes means to calculate a temperature of the each heating field effect transistor using a measured change in at least one electrical characteristic of the each sensing field effect transistor caused by the heat generated at the each heating field effect transistor. In another aspect of the invention, an apparatus for determining the temperature of an active region of a semiconductor device includes three silicon sections, and three pairs of active regions, wherein each pair of active regions is arranged on a respective silicon section, wherein each pair of active regions is configurable to produce and sense heat. The apparatus also includes three thermal conductors, wherein each thermal conductor is arranged between each active region of each respective pair of active regions. In another aspect of the invention, a computer program product comprising a computer usable medium having readable program code embodied in the medium, the computer program product includes at least one component to measure a change in at least one electrical characteristic of a first sensing device caused by heat generated at a first heating device, and calculate a temperature of the first heating device using the measured change in the at least one electrical characteristic. The invention is directed to determining the temperature of an active region of a device fabricated using SGOI technology as a function of electrical power through the device. Such information may be used to characterize the effect of temperature on the performance characteristics of the device. This information may be utilized by design engineers to predict device performance during the circuit layout process, as a function of temperature, and to accommodate such temperature effects in the circuit design. To reliably measure the temperature of a semiconductor device, and thereby determine the effects on performance of the device, a method includes fabricating two devices in thermal communication with one another, where the first device is run at a predetermined power level, and the second device is a prescribed distance from the first device. The second device, otherwise known as the measuring device, is capable of providing temperature information at its position as a function of the first device, e.g., power level of the heating device. In operation, a particular electrical characteristic of the second device is monitored to determine the temperature of the second device. This characteristic may be, for example, the sub-threshold voltage slope (also referred to as “Sub Vt”) as shown in FIG. 1. In this embodiment, multiple devices are used at varying distances from the heating device to make various temperature measurements at varying distances. The measurements are used to calculate the temperature of the active region of the heating device by extrapolating the distanced-based temperature measurements back to the origin, i.e., the position of heating device. In one embodiment, the fabricated device may include multiple SiGe islands which are fabricated where each SiGe island has its own measurement and heating device pair. For example, two FETs may be fabricated on a SiGe island a prescribed distance apart. After fabrication, a particular electrical characteristic of the second FET is measured at multiple ambient temperatures. The power levels and distances as well as other measurable phenomena and features are shown for illustration purposes in FIGS. 1 and 2 as well as the exemplary data and results shown in Tables 1-5. The method is illustrated in more detail in FIGS. 9 and 10 as discussed below. By carrying out the measurement process for each measurement/heating device pair, the temperature and power information can be used to calculate the temperature of the active device. Although the measurement device typically has multiple electrical characteristics which will vary as a function of its temperature, an embodiment of the invention uses a sub-threshold voltage slope as a function of temperature to measure temperature. Referring to FIG. 1, a graph showing sub-threshold voltage slope versus power at 25° C. ambient temperature is shown. The y-axis represents channel current of the measurement device, and the x-axis represents various voltages as the source input voltage is varied, where the two gates and the two drains of the devices are common to one another making Vgs (voltage between the gate and source) equal to Vds (voltage between the drain and source). Accordingly, Vgs=Vds as noted on the x-axis of FIG. 1. The common diffusion input and the measurement source may be connected with Kelvin connectors. The information of FIG. 1 is an example of an electrical characteristic of the sensing device from which its temperature is inferred. This information may be used in the method of the invention as described herein. The data for each power point is a voltage ramp of the measurement source from, for example, Vgs=Vds=1.0V to Vgs=Vds=0.6 in 0.010V increments. The sub-threshold voltage slope is fit below Vt and above the point where the behavior deviates from log-linear. For typical devices, Vgs=Vds=0.1v to 0.2v was used. For reference, the single point Vt is 11.25 μA for a device. It should be noted that similar plots are typically constructed and the slope calculated for a device at 50° C., 75° C., and 100° C. with no power in the heating device. From these plots, the relationship between sub-threshold voltage slope and temperature of the measurement device may be determined. The temperature of the measurement device can be determined by comparing the sub-threshold voltage slope at a given heating device power to the relationship features sub-threshold slope and temperature previously determined with no power to the heating device. Referring to FIG. 2, a change in temperature versus power of a self-heating semiconductor device is shown. The graph of FIG. 2 is an example of the graphs generated for each measurement and heating device pair at various distances. Thus, the measurement with heating devices for FIG. 2 is illustrative examples, and should not be considered a limiting feature. Other examples are contemplated and can be easily determined by those of skill in the art in view of the present invention. This illustrative information of FIG. 2 is the final result after extrapolation and all measurements and calculating steps resulting from the flow of the invention as described in more detail with reference to FIGS. 9 and 10. In the graph, the y-axis represents the change in temperature in degrees centigrade, and the x-axis represents the power applied to the device in micro Watts (μW) at 25° C. ambient temperature. The line on the graph was derived for heating bias conditions of Vgs=Vds=0V to 1V in 0.1V increments. These slopes were converted to temperature changes and plotted versus the applied power. FIGS. 3 through 7 illustrate steps in fabricating an embodiment of the self-heating monitor. Referring to FIG. 3, an SOI wafer is shown. The SOI wafer includes a silicon substrate 12 overlaid by a thin, buried oxide layer 14. On top of the buried oxide layer 14 is arranged a layer of silicon 16. The buried oxide layer 14 and silicon layer 16 may be formed on a silicon substrate 12 by any of the methods well-known in the art, such as, for example, a high-energy oxygen implant and then activating the oxygen to form the buried oxide layer 14. Referring to FIG. 4, well-known standard photo lithographic imaging and etching techniques are preformed to remove portions of the silicon layer 14 to form a silicon island 18 on top of the buried oxide layer 14. Next, shallow trench isolation oxides 20 are formed surrounding the edges of the silicon island 18. The oxide is deposited into the regions surrounding the edge of the silicon island 18 by any of the oxide deposition techniques well-known in the art. The silicon island 18 is then implanted with a well doping ion (P-type for N-channel devices, and n-type for P-channel devices) using any of the doping techniques well known in the art. In the example shown in FIG. 4, the channel regions of the two devices being fabricated are doped with P-type dopant. Referring to FIG. 5, the surface of the silicon island 18 is cleaned and a gate oxide 22 is formed on the surface of the island 18. The gate oxide 22 can be formed by any of the methods for depositing, imaging and etching gate oxides well known in the art. A gate polysilicon layer 24 is formed on top of the gate oxide 22, and patterned and etched to form the device gate structure using any of the methods well known in the art to fabricate gate polysilicon layers. After the gate polysilicon 24 is formed, spacers 26 are formed on top of the gate oxides 22 and abutting the edges of the gate polysilicon 24. The spacers 26 can be formed from materials and using methods well known in the art and include, for example, nitride or oxide spacers. After the spacers 26 are formed, the region of the silicon island 18 not covered by the gate structures is implanted 19 with the appropriate dopant relative to the channel dopant. In this example, the channels are doped with p-type ions, and thus the regions in the silicon island 18 to either side of the gate structures is doped with n-type ions, as well as doping the gate polysilcon 24. Doping techniques which may be used for this step are well known in the art. In this example, the n-doped regions form the source and drain regions 28, and the p-doped region forms the channel 30 of the measurement and heating device pair The dopant types, concentrations and energy levels would be those appropriate for whichever type of device is being fabricated, and are well-known in the art. Referring to FIG. 6, the silicide layer 32 is formed over the source and drain regions, 28, and the gate-polysilicon 24. After the silicide layer 32 is formed, a planar oxide layer 34 is deposited over the shallow trench oxide regions 20, silicide 32 and gate structures. The planar oxide 34 can be deposited using any of the methods well-known in the art. Referring to FIG. 7, vias 35 are patterned and etched in the planar oxide 34 from a top surface of the planar oxide 34 down to the silicon island 18. The vias 35 are filled with conductive material, such as metal to form contacts 36 to the source and drain regions 28 of the silicon island 18. After the contacts 36 are formed, a first metal wiring layer 38 is deposited on top of the planar oxide 34. The first metal wiring layer is then etched to form metal contacts 38 on a top surface of the planar oxide 34, in electrical contact with the contacts 36. Referring to FIG. 8, a top view of an embodiment of the device 50 is shown, illustrating various metal contacts to the devices on silicon island 52. The silicon island 52 has a heating device drain metal contact 60, and a measurement device drain contact 62. Also included on the silicon island 52 is a common source contact 54 which leads to the source of both the underlying measurement device and the heating devices. A measurement gate contact 56 and a heating gate 58 are also included on the silicon island 52. It should be noted that although a particular example of forming the heating and measuring devices is discussed above, any fabrication process which produces a pair of semiconductor devices such as FETs in thermal communication with one another may be suitable for the temperature measurement. As such, the temperature measurement method can work with any pair of devices where one of the devices produces heat, and the other device responds to that heat in some measurable way, such as one of its electrical characteristics changing in accordance with its temperature. It should also be noted that while the examples discussed above use two devices of similar design for heating and measuring, the measurement device and the heating device may be of completely different designs as long as the heating device is capable of heating and the measurement device is capable of measuring temperature and the two are in thermal contact. Additionally, two devices may be in thermal contact with one another without actually being in physical contact with one another. Thus, two devices may be in thermal contact while actually touching one another, and two devices may be in thermal contact where the thermal contact is through an intermediate thermal conductor such as a length of silicon in physical contact with one another. It should be noted that the silicon island may have a perimeter which describes a square as well as other geometric shapes such as rectangles, circles, oblongs, triangle, etc. Additionally, although the measurement device in FIGS. 3-7 show two FETs, the measurement device can be made with more FETs. In general, the measurement method relies upon two semiconductor devices where one device functions as a heat source, and the other devices functions as a temperature measurement. For example, the temperature measurement method may rely on two FETs fabricated on a single silicon island, which is substantially surrounded on all sides by material having low thermal conductivity as discussed below. One of the FETs is referred to as the heating device, and generates heat when it is powered up. The second device is referred to as the measurement device, and measures the temperature in its active region by sensing a change in a prescribed electrical characteristic of its active region, which is then correlated to temperature. In implementation, the distance between the heating device and the measurement device is varied and thus the temperature at the heating device may be extrapolated from multiple measurements. To determine the correspondence between power applied to the heating device and its temperature, a series of measurements for each heating/measurement pair is performed where varying amounts of power are applied to the heating device and the temperature at the measurement device is determined. This process may be repeated for the multiple pairs of heating/measurement devices where each pair has a different distance between the heating and measurement device. By measuring the temperature at varying distances from the heating device, the data can be extrapolated and the temperature of the active region of the heating device under various amounts of power can be determined. Typical distances between the heating device and the measurement device range from ¼ of a micron to one or two microns. It should be realized that due to the thermal characteristics of the silicon island, multiple measurements using different pairs of heating and measurement devices in different geometries may be required to accurately determine the relationship between power applied to the heating device and its temperature. For example, multiple pairs of heating and measurement devices may be made where each measurement device corresponds to a particular heating device and with a different separation distance. Before actual measurements can be made, the measurement device of each heating/measurement pair should be calibrated. The calibration of the measurement device is performed by measuring a particular electrical characteristic of the active region of the measurement device with the measurement device held at a known ambient temperature. For example, the measurement device can be held at 25° C. and the sub-threshold voltage slope is measured in a range from 0-0.4 volts driving voltage of the heating device. The sub-threshold voltage slope may be measured by holding the gate voltage at 0 and sweeping the drain through the desired voltage range. Next, 0.1V is applied to the drain of the heating device and another sweep of the drain of the measurement device is done. This process maybe repeated in 0.1V steps of the drain of the heating device until the entire desired range is swept through. Other voltage steps are also contemplated. The collected data produces a sub-threshold voltage slope at each different increment of power corresponding to the voltage applied to the drain of the heating device as illustrated by the example of FIG. 1. Additionally, the current of the heating device is measured so that the power being dissipated by the device is known. Such a process is repeated for multiple temperatures and a sub-threshold voltage slope versus temperature relationship is derived. For example, the process is repeated for 15° C., 75° C. and 100° C. temperature points. Accordingly, four sub-threshold voltage slopes as a function of ambient temperature are determined with no self-heating at the heating device to calibrate the measurement device. For the typical geometry of heating/measurement device, there may be thermal effects which would not be present during operation of the heating device in an actual application as distinguished from having its temperature measured. For example, the presence of a measurement device may require metal contacts which otherwise would not normally be there. Such metal contacts may act as heat sinks which conduct heat away from the heating device, thereby effectively reducing the temperature below that at which it would normally operate. This thermal effect can be understood and accounted for by fabricating the heating/measurement device pairs so that such thermal effects are the same from pair to pair. Accordingly, it is advantageous to make the thermal resistance between the devices for each pair of devices as similar as possible. Referring to FIG. 9, a flow chart of an embodiment of the measurement method is shown. FIGS. 9 and 10 may equally represent a high-level block diagram of components of the invention implementing the steps thereof. Several of the steps of FIGS. 9 and 10 may be implemented on computer program code in combination with the appropriate hardware. This computer program code may be stored on storage media such as a diskette, hard disk, CD-ROM, DVD-ROM or tape, as well as a memory storage device or collection of memory storage devices such as read-only memory (ROM) or random access memory (RAM). Additionally, the computer program code can be transferred to a workstation over the Internet or some other type of network. In step S10, a sensing device of a device pair having a sensing device and a heating device is calibrated. The calibration may include of determining the variation of a particular electrical characteristic of the sensing device as a function of temperature. In step S20, the calibration process S10 is repeated for a different distance between the sensing and heating device. This typically involves a different device pair arranged at a distance different from the previously calibrated device pair. In S30 of FIG. 9, a temperature measurement is processed for a device pair at a particular distance. The temperature measurement typically involves running the heating device at various power levels. In S40, the temperature measurement is performed at a different distance for a device pair, and typically involves a different device pair arranged at a distance different from the previously measured device pair. S50 includes calculating a temperature versus power level relationship for the heating device using the data collected at different power levels and different distances between the sensing and heating devices. Referring to FIG. 10, an example of an embodiment of the measurement method is shown. In S100, a sensing device of a first device pair having a sensing and a heating device is calibrated for a chosen separation distance. The calibration process includes monitoring a particular electrical characteristic of the sensing device while the device pair is held at a selected ambient temperature with no power being applied to the heating device. For example the sub-threshold voltage slope of the sensing device may be monitored while the device pair is held at 25° C. In S120, the calibration process is repeated while the device pair is held at a second ambient temperature, such as, for example, 50° C. In S140, the calibration process is repeated at a third temperature, such as, for example 100° C. A relationship between the temperature of the device and the chosen electrical characteristic is then calculated by fitting the data to a curve, such as by a least squares method. Alternatively, the data may be fitted to a curve using any of the curve fitting software packages which are well known in the art In S160, the calibration process is checked to determine whether the process had been performed for each selected distance. If device pairs are to be calibrated at other distances, the calibration process is repeated at S180 for all the chosen temperatures at a different distance between a sensing and heating device. This typically involves a new device pair having of a similar sensing and heating device as in the previous device pair, but arranged with a different separation distance. After the calibration steps, S100, S120, S140, S160 and S180 are complete, a measurement is made in S200 at a first power level for the heating device and a first distance separating the device pair. The measurement is repeated for a second and third power level in steps S220 and S240, respectively. When all the predetermined power levels have been measured, the process is repeated in S260 and S280 if other distances are to be measured. Once all measurement data has been collected in steps S200, S220, S240, S260 and S280, a relationship between temperature and power of the heating device is calculated using, for example, any of the commercially available curve fitting software packages capable of fitting data to a curve. The relationship between the temperature and power of the heating device is typically an exponential relationship, although other relationships are possible and can be accommodated by the method. Table 1 shows examples of dimensions of various devices which were fabricated for measurements of temperature using the above-described method. As can be seen from Table 1, silicon islands of various lengths and widths were used as well as gate spacings between the gates of the heating and measurement device. The devices also had various numbers of contacts, ranging from 2 to 4 to 6. Additionally, the comments section includes information about the device geometry such as gate spacing and length. TABLE 1GateRxLdWdSpacingLengthDevice(μm)(μm)(μm)(μm)ContactsCommentsM1S10.875.12250.26250.89252/3XVerynarrow RxM1S20.8751.750.981.614/4XWidespacebetweengateswithout AlbridgeM1S30.8751.750.50751.13754/4XMediumspacebetweengates withAl bridgeM1S40.8751.750.26250.89254/3XBasedeviceM2S10.8750.8750.26250.89252/3XNarrow RxM2S30.8751.750.26250.89256/3XExtracontacts &coolingfinsM2S40.8751.750.262535.93514/3XVery longRxM3S10.8751.750.26252.64254/3XMediumlong RxThe various device geometries are labeled, for example, M1S1. Thus, various device geometries were tested. Table 2 shows the heating effects for three different gate spacings of 0.26 micron, 0.51 micron and 0.98 micron. The column labeled Waf T5TY and at Waf Q878 each represent different SOI wafers from which the respective measurement devices were manufactured. For example, the device type M1S4 was manufactured on two different wafers, the first labeled T5TY, and the second labeled Q878. TABLE 2Waf T5TYWaf Q878PC spacingDeg C/mW/uDeg C/mW/uCommentsBase device,33.2536.75M1S4-PC − PC = 0.264CA/3XPC − PC = 0.5118.819.1M1S3-4CA/4X, A1bridgePC − PC = 0.985.57.9M1S2-4CA/4X, NoA1 bridge Referring to Table 3, the results of adding contacts is shown to the device M1S4 and the device M2S3. TABLE 3Waf T5TYWaf Q878Added CADeg C/mW/uDeg C/mW/uCommentsBase device,33.2536.75M1S44CA/3X6CA/3X &26.2531.5M2S3“cooling fins”Referring to Table 4, devices of different widths are shown and the results thereto. TABLE 4DeviceWaf T5TYWaf Q878widths/Ca/uDeg C/mW/uDeg C/mW/uCommentsBase device,33.2536.75M1S44CA/3X,W = 1.752CA/3X,28.530.3M2S1W = 0.8252CA/3X,3.10.8M1S1W = 0.1225Referring to Table 5, the results of various RX paths of the device are shown for the devices M1S4, M3S1 and M2S4. TABLE 5Waf T5TYWaf Q878RX past deviceDeg C/mW/uDeg C/mW/uCommentsBase device,33.2536.75M1S4RX = 0.89RX = 2.6433.2526.3M3S1RX = 35.930.9N/AM2S4 In the embodiments described above, the heat reducing effects of the contacts between the gates was considered negligible. However, to more accurately determine the self heating, the embodiments of FIGS. 11-13 take into account the thermal resistance of the contacts in order to more accurately assess the self heating. As should be understood, the contacts create a thermal resistance. That is, the contacts act as heat sinks which take heat away from the silicon island 52 which, in turn, reduces the temperature at the measurement gate contact 56. Accordingly, without taking into account the number of contacts, the reading of the temperature may vary by a certain offset, equal with a proportionality to the number of contacts. But, to compensate for this offset, the technique of FIGS. 11-13 use a measurement differential taken with at least two devices having a different number of contacts and extrapolating the results to zero contacts. In embodiments, this is accomplished by adding or subtracting the number of contacts between two or more different devices, as discussed below. The measuring technique associated with FIGS. 11-13 is applicable to both wide and narrow type devices, known in the art. In the measuring technique of this aspect of the invention, the number of contacts will vary on the diffusion between the measurement gate contacts 56 and heating gate contacts 58 of different measured devices to establish the rate of temperature change per contact. Once the rate of temperature change is established for the devices, it is then possible to extrapolate that change to zero contacts to determine the actual device temperature without the effect of the offsetting contacts between the gates. By way of example, FIGS. 11 and 12 represent a top view of a wide device 50. Similar to the previous embodiments, in FIGS. 11 and 12, the silicon island 52 has a heating device drain metal contact 60 and a measurement device drain contact 62. Also provided on the silicon island 52 is a common source drain contact 54 which leads to the source drain of both the underlying measurement device and the heating device. A measurement gate source contact 56 and a heating gate source 58 are further provided on the silicon island 52. FIG. 11 shows 12 contacts on the common drain 61; whereas, FIG. 12 shows six contacts. The different number of contacts between FIGS. 11 and 12 provides for a measured temperature differential in accordance with this aspect of the invention. It should be understood that any number of contacts may be used with the extrapolative method of this embodiment by either adding or subtracting contacts. By illustrative example, in the embodiment of a wide device such as shown in FIGS. 11 and 12, it is preferable to subtract contacts. Thus, FIG. 11 includes 12 contacts and FIG. 12 shows six contacts. However, in narrow devices such as shown in FIG. 13, it is preferable to add contacts. Thus, in the narrow device example, it may be necessary to add an additional contact 61A to reduce the thermal resistance which, in turn, results in a measured heat differential (e.g., difference of the thermal load) between a first narrow device with one contact and a second narrow device with two contacts. Once these temperature differentials are measured, the results are then extrapolated to zero contacts, in any well known extrapolative technique. In one aspect of the invention, for a wide device the procedure should be applied to the smallest gate spacing, since the measurement temperatures (signal) is highest and the effect will be largest. The number of contacts is then reduced to allow the slope of the measured temperature to be determined versus the number of contacts. To help calibrate the narrow device the total number of contacts is kept the same as a case where all the contacts are an equal distance from the gate edge to help further quantify the results for one case. Additional cases with fewer contacts could also be added to further determine the proper scaling. While the invention has been described in terms of embodiments, those skilled in the art will recognize that the invention can be practiced with modification within the spirit and scope of the appended claims. For example, the invention can be readily applicable to multiple measurement devices associated with the single heating device. |
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description | Referring to FIGS. 1 and 2, a Positron Emission Tomography (PET) system 10 is shown including a gantry 12, a rotatable transmission ring 14 including a bore 15. In use, a patient 16 is positioned within bore 15 and PET system 10 is utilized to image portion or organs of patient 16 as is known in the art. Patient 16 is positioned on a table 17 which is translatable to move into and out of bore 15. System 10 also includes a storage device 18 for storing one or more radioactive source pins. In the exemplary embodiment, three source pins 20, 22, and 24 are stored in storage device 18. One radioactive source pin 20, 22, or 24 is removed from storage device 18 and installed in transmission ring 14 to calibrate PET system 10. In one embodiment, source pin 20, 22, or 24 is also removed from storage device 18 and installed in transmission ring 14 to provide attenuation measurements during patient scanning. FIG. 3 is a partial rear view of PET system 10 including storage device 18. Storage device 18 includes primary shielding 30, a rotatable shielding cylinder 32, storage cavities 34, 36, and 38, and a rotating mechanism 40. Primary shielding 30 provides sufficient attenuation of radioactive source pins 20, 22, and 24 to protect the environment near the PET system 10, including personnel. Rotatable shielding cylinder 32 is within primary shielding 30 and is selectively rotated or indexed. Storage cavities 34, 36, and 38 have cylindrical cross-sectional profiles that are substantially concentric with respect to respective axes 42, 44, and 46. Each storage cavity 34, 36, or 38 is sized to contain a portion of one source pin 20, 22, or 24. In the exemplary embodiment, rotatable shielding cylinder 32 is indexed by rotating mechanism 40 to four positions, including a storage position (not illustrated), and an access position 50 for each storage cavity 34, 36, and 38. When rotatable shielding cylinder 32 is indexed to the storage position, rotatable shielding cylinder 32 is positioned such that storage cavities 34, 36, and 38 are substantially centered within primary shielding 30. FIG. 3 illustrates storage cavity 34 in access position 50 such that axis 42 is aligned substantially perpendicular to transmission ring 14 and co-axially with one of a plurality of receiver openings 58 in transmission ring 14. In one embodiment, receiver openings 58 include magnetic material to secure source pins 20, 22, or 24. Transmission ring 14 is also indexed to ensure receiver openings 58 are aligned to access position 50. Control of rotating mechanism 40, transmission ring 14 rotation, and operation of PET system 10 are controlled as is known in the art. FIG. 4 is an exploded perspective view of one storage cavity 34 of storage device 18 and receiver opening 58 together forming a radioactive source pin transport system 60. System 60 includes a housing 62 and an electromagnet 64 positioned within housing 62. An electromagnet core 66 is positioned within electromagnet 64. A magnetic cover 68 maintains a ring magnet 70 against electromagnet core 66. Source pin 20 includes a radiation portion 72 and a non-radiation portion 74. A proximity sensor 80 is positioned to detect a presence of source pin 20 within housing 62. In one embodiment, proximity sensor 80 is a normally open Negative-Positive-Negative (NPN) inductive sensor. Also, in an exemplary embodiment, proximity sensor 80 and source pin 20 are axially aligned such that sensor 80 axially senses a presence of source pin 20 within housing 62. Pin transport system 60 also includes a transmission ring magnetic pin holder 82 that is positioned on transmission ring 14. At least one magnet 84 is positioned within holder 82 and maintained in place with a holder cover 86. In one embodiment, magnet 84 includes two ring shaped permanent magnets each having a force of about 5.34 Newtons (N) providing a combined force of about 10.67 N. Additionally, ring magnet 70 also has a force of 5.34 N and is similarly sized to magnet 84, and because magnets 70 and 84 are thus interchangeable, construction of system 60 is simplified over designs using magnets of different strengths and/or sizes. FIG. 5 is a cut away view of source pin 20 (shown in FIG. 2) positioned at least partially within electromagnet core 66 positioned within electromagnet 64 (shown in FIG. 4). FIG. 6 is a partially cut away view of sensor 80 (shown in FIG. 4) positioned to sense a presence of source pin 20 (shown in FIG. 2) within housing 62 (shown in FIG. 4) in accordance with one embodiment. In use of system 60, radio-active source pin 20 is released from storage device 18 when there is a net force along the axis centerline of source-pin 20 that is pointing towards magnetic pin holder 82 on transmission ring 14. This state is reached when electromagnet 64 is de-energized and the only pull force towards storage device 18 is that of permanent magnet 70 positioned inside housing 62, in this situation a pull force of transmission ring magnetic pin holder 82 on transmission ring 14 of about 2.4 Pound-force (lbf) (10.67 N) is approximately twice of that of permanent magnet 70 (about 1.2 lbf, 5.34 N) inside housing 62. Consequently, a net force exists of about 1.2 lbf (5.34 N) towards magnetic pin holder 82, and hence source-pin 20 is accelerated over a small distance to end up positioned flush with transmission ring magnetic pin holder 82. Additionally, system 60 allows for an easy removal of source pin 20 from transmission ring 14. During this removal process, a reverse logic is utilized. Conversely, in this removal process, electromagnet 64 is energized, which produces a nominal pull force of approximately 5.3 lbf (23.6 N). The orientation of permanent magnet 70 inside housing 62 is such that the cumulative effect of the total pull force is the vectorial sum of permanent magnet 70 and a electromagnet force of attraction from electromagnet 64, thus resulting in a net pull force of approximately 4.1 lbf (18.24 N). This force accelerates source-pin 20 towards housing 62 over a small distance between transmission ring 14 and storage device 18 and maintains source-pin 20 in a storage position within housing 62. Housing 62 is rotated away from transmission ring magnetic pin holder 82, and electromagnet 64 is de-energized, and source-pin 20 is maintained within housing 62 solely via permanent magnet 70 in housing 62. In one embodiment, transmission ring 14 is aligned with source pin 20 within storage device 18 wherein storage device has at least two magnetic forces including a permanent magnet force of at least about 5.34 Newtons (N) and an electromagnet force of at least about 23.6 N holding the source pin in place. Source pin 20 is moved by de-energizing the electromagnet force and moving the source pin from the storage device to the transmission ring using a magnetic force of at least about 10.67 N. These herein described forces have empirically shown to be highly effective for accurately and quickly moving source pin 20 back and forth between transmission ring 14 and storage device 18. Additionally, in one embodiment, system 10 includes a processor (not shown) programmed to perform the functions herein described. It is contemplated that the benefits of the invention accrue to embodiments employing a programmable circuit other than those known in the art as processors, therefore, as used herein, the term processor is not limited to just those integrated circuits referred to in the art as processors, but broadly refers to computers, processors, microcontrollers, microcomputers, programmable logic controllers (PLCs), application specific integrated circuits (ASICs), field programmable gate array (FPGA), and other programmable circuits. Additionally, although the herein described methods are described in a medical setting, it is contemplated that the benefits of the invention accrue to non-medical imaging systems such as those systems typically employed in an industrial setting or a transportation setting, such as, for example, but not limited to, a baggage scanning system for an airport or other transportation center. The benefits also accrue to micro PET systems which are sized to study lab animals as opposed to humans. Also provided herein is a Fail Safe Mode. The fail safe mode is to continuously energize electromagnet 64 during source-pin transit and only de-energize electromagnet 64 during the above described pin-release process. In this mode, permanent magnet 70 inside housing 62 acts as a fail safe feature, such that if electromagnet 64 lost power, then housing 62 is still capable of retaining source-pin 20 via the pull force of magnet 70. The herein described methods and apparatus facilitate an increase in component and system reliability, since the radio-active source pin exposure and storage process is of importance relative to system operation and up-time. This is at least partially due to the reason that typical software operating on PET systems is configured such that the system will stop functioning and log a system error if this fault occurs. The herein described methods and apparatus facilitate a secure and reliable means of grabbing and releasing the source-pin. The methods and apparatus herein described also facilitate a cost savings based on production costs. System 60 utilizes no moving parts, and uses an electrical signal as a means of latching radioactive source-pin 20, and verses known mechanical transport systems that utilize moving components which wear due to cyclical motion, system 60 provides a long lasting and cost effective method to transport source pins between a transmission ring and a storage device. PET system embodiments of the present invention are cost-effective and highly reliable. A storage device includes a rotatable shielding cylinder that rotates a selected storage cavity to an access position that is aligned with a receiver opening in a transmission ring. A source loader linearly transports a source pin and installs the source pin in the transmission ring. Similarly, the source pin is removed from the transmission ring and returned to the storage cavity. The rotatable shielding cylinder then rotates to a storage position. As a result, embodiments of the present invention facilitate quick and reliable handling of radioactive source pins. While the invention has been described in terms of various specific embodiments, those skilled in the art will recognize that the invention can be practiced with modification within the spirit and scope of the claims. |
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048636711 | claims | 1. A plasma confinement system comprising a toroidal vacuum chamber, a toroidal coil which generates a magnetic field in a toroidal direction within said vacuum chamber, current transformer coils which are wound in the toroidal direction, equilibrium magnetic field coils which are wound in the toroidal direction in order to control a plasma, alternating current coils which are wound mainly in the toroidal direction and through which alternating currents are caused to flow for enabling forming and rotating of a deformed non-circular magnetic surface and for causing rotation of the plasma in a poloidal direction, and power source means for causing currents to flow through the various coils. 2. A plasma confinement system as defined in claim 1 wherein said alternating current coils are wound axi-symmetrically with respect to a symmetry axis of rotation of said toroidal vacuum chamber. 3. A plasma confinement system as defined in claim 1 wherein said alternating current coils are wound non-axisymmetrically with respect to a symmetry axis of rotation of said toroidal vacuum chamber. 4. a plasma confinement system as defined in claim 1, wherein said alternating current coils are arranged in a plural number in the poloidal direction and are respectively supplied by said power source means with the alternating currents of different phases. 5. A plasma confinement system as defined in claim 1, wherein a rotating angular velocity of w of the deformed magnetic surface satisfies the following formula: ##EQU5## where v.sub.th denotes a thermal velocity of ions and r denotes a radius of the plasma. 6. A plasma confinement system as defined in claim 1 wherein said alternating current coils numbering 2M are arranged at substantially equal poloidal angular intervals, and are fed with the currents so as to shift current phases of the respectively adjacent coils in an amount of: ##EQU6## where N denotes an integer which is smaller than M and which is at least 2. 7. A plasma confinement system as defined in claim 6, wherein said alternating current coils in a plurality of sets are fed with the currents by and power source means including an alternating signal generator, a plurality of phase shifters and a plurality of power amplifiers. |
claims | 1. A controller-executed method for detecting changes in a performance metric in an automated information technology (IT) management system comprising:defining, using a hardware processor, segments as sets of contiguous time samples wherein time samples within a particular segment are mutually more similar in terms of performance metric behavior than time samples in previous and subsequent segments;determining, using the hardware processor, for each of the segments, a weighted cost based on a compactness parameter and a sum over a plurality of time samples derived from a difference between a metric value for each of the plurality of time samples and an average value of the performance metric for the plurality of time samples; andfinding, using the hardware processor, a segmentation that minimizes a sum of the weighted costs for the segments. 2. The method according to claim 1 wherein determining the weighted cost for each of the segments is based on the compactness parameter and a sum over the plurality of time samples of a square of the difference between the metric value for each of the plurality of time samples and the average value of the performance metric for the plurality of time samples. 3. The method according to claim 1 wherein the compactness parameter is based on an anomaly definition. 4. The method according to claim 1 further comprising:detecting changes in service conditions using the segmentation; anddetecting changes in cyclic performance behavior using the segmentation. 5. An automated information technology (IT) management system comprising:a computer comprising:a hardware processor to execute:logic to define segments as sets of contiguous time samples wherein time samples within a particular segment are mutually more similar in terms of performance metric behavior than time samples in previous and subsequent segments;logic to determine, for each of the segments, a weighted cost based on a compactness parameter and a sum over a plurality of time samples derived from a difference between a metric value for each of the plurality of time samples and an average value of the performance metric for the plurality of time samples; andlogic to find a segmentation that minimizes a sum of the costs for the segments. 6. The system according to claim 5 further comprising:a response tool executable in the computer to respond to detection of changes in the performance metric. 7. The system according to claim 5:wherein the computer further comprises:logic to determine the weighted cost wi,j for each of the segments according to an equation: w i , j = ∑ t = i j ( X t - μ i , j ) 2 + λ wherein Xt is the metric value, μi,j is the average value of the performance metric, and λ is the compactness parameter. 8. The system according to claim 5:wherein the logic to find the segmentation is based on a compactness parameter λ, and wherein the performance analyzer further comprises:logic to define an anomaly determined by a length m of an anomalous segment and deviation of the anomalous segment from normal behavior;logic to determine the compactness parameter λ according to the defined anomaly; andlogic to scale the compactness parameter λ by an estimate of a data variance according to an equation: λ = m ( μ n - μ m ) 2 2 K ,wherein a segment is a mixture of two distributions comprising a first distribution of mean μn with n time samples and a second distribution of mean μm with m time samples, and K is the estimate of data variance. 9. The system according to claim 5:wherein the computer further comprises:logic to determine the weighted cost wi,j for each of the segments according to equation: w i , j = 1 K ∑ t = i j ( X t - μ i , j ) 2 + λ wherein Xt is the metric value, μi,j is the average value of the performance metric, λ is the compactness parameter, and K is estimated data variance. 10. An article of manufacture comprising:a non-transitory computer-readable storage medium storing computer readable program codes executable by a computer for detecting changes in a performance metric, the computer readable program codes comprising:code causing the computer to define segments as sets of contiguous time samples wherein time samples within a particular segment are mutually more similar in terms of performance metric behavior than time samples in previous and subsequent segments;code causing the computer to determine, for each of the segments, a weighted cost based on a compactness parameter and a sum over a plurality of time samples derived from a difference between a metric value for each of the plurality of time samples and an average value of the performance metric for the plurality of time samples; andcode causing the computer to find a segmentation that minimizes a sum of the weighted costs for the segments. |
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053176154 | summary | FIELD OF THE INVENTION AND RELATED ART This invention relates to an exposure view-angle limiting means suitably usable, for example, in a step-and-repeat type exposure apparatus wherein a mask and a wafer are disposed close to or in contact with each other and wherein exposures are made to divided zones on the wafer sequentially. The manufacture of integrated circuits or semiconductor devices includes repetition of a process of optically transferring a pattern, formed on a mask, onto a wafer coated with a photosensitive material. Particularly, in consideration of the decreasing linewidth of the pattern, exposure apparatuses of the type using X-rays as an exposure light source are used. A practical example of such X-ray exposure apparatuses is a step-and-repeat type exposure apparatus wherein the surface of one wafer is divided into plural zones so as to reduce the size of the exposure region which is an area to be exposed at once, and wherein exposures are made to the zones of the wafer while sequentially moving the wafer relative to a mask. Such a step-and-repeat method is effective in a high-precision X-ray exposure apparatus. In this method, it is necessary to use some means for limiting the exposure region so as to prevent a neighboring portion of a wafer zone from being exposed. Also, around a circuit pattern region of a mask, usually there are provided alignment marks for effecting alignment between a circuit pattern of the mask with a circuit pattern already formed on a wafer, and the exposure apparatus is equipped with detecting means for detecting any positional deviation of such alignment marks. In order to prevent exposure of these alignment marks, preferably an aperture means for limiting the exposure region is provided to cover the alignment marks against the exposure illumination. SUMMARY OF THE INVENTION The present invention, in one aspect, pertains to an exposure apparatus of the type as described above, and it is an object of the present invention to provide a structure for controlling an X-ray exposure region very precisely when the same is to be limited, without blocking path of an optical system of a positional deviation detecting system and without exposing an alignment mark on a wafer. These and other objects, features and advantages of the present invention will become more apparent upon a consideration of the following description of the preferred embodiments of the present invention taken in conjunction with the accompanying drawings. |
description | This is a continuing application, under 35 U.S.C. § 120, of copending international application No. PCT/EP2004/001515, filed Feb. 18, 2004, which designated the United States; this application also claims the priority, under 35 U.S.C. § 119, of German patent application No. 103 09 742.2, filed Mar. 6, 2003; the prior applications are herewith incorporated by reference in their entirety. The invention relates to a spacer for a fuel assembly of a nuclear reactor cooled by light water, as is disclosed, for example, in published European patent application EP 0 237 064 A2 (corresponding to U.S. Pat. No. 4,726,926). The known spacer is constructed from a multiplicity of crisscrossing webs that form a grid with a multiplicity of meshes. Each web is formed by two thin sheet-metal strips welded to one another. The sheet-metal strips are provided in each case with raised corrugations that extend into the interior of the grid mesh bounded in each case by the sheet-metal strip. Neighboring corrugations, respectively opposite one another, of the sheet-metal strips assembled to form a web form an approximately tubular flow sub-channel extending in a vertical direction. The flow sub-channels are inclined relative to the vertical and produce a flow component of the cooling liquid that is oriented parallel to the web and directed to a crossing point of the webs. The component produces a swirl flow around the fuel rods respectively penetrating the meshes. In the case of the known spacer, these corrugations serve at the same time, moreover, as a bearing for the fuel rods penetrating the meshes. The fuel rod bearing has proved to be particularly advantageous in practice, since only slight fretting defects are observed on the fuel cans when use is made of such spacers. It is accordingly an object of the invention to provide a spacer that overcomes the above-mentioned disadvantages of the prior art devices of this general type, which simultaneously exhibits improved thermohydraulic properties in conjunction with a high level of resistance to fretting. With the foregoing and other objects in view there is provided, in accordance with the invention, a spacer for a fuel assembly of a nuclear reactor cooled by light water. The spacer contains a multiplicity of crisscrossing webs forming a grid. Each of the webs are formed of interconnected first and second sheet-metal strips having corrugations such that in each case neighboring corrugations form a flow sub-channel running oblique to a vertical. The neighboring corrugations impart to cooling water emerging from the flow sub-channel a flow component perpendicular to a vertical middle plane running between the sheet-metal strips. Such a spacer for a fuel assembly of a nuclear reactor cooled by light water is constructed from a multiplicity of crisscrossing webs that form a grid and respectively are formed of interconnected first and second sheet-metal strips. The sheet-metal strips have corrugations in such a way that in each case neighboring corrugations form a flow sub-channel and are fashioned in such a way that they impart to the cooling water emerging from the flow sub-channel a flow component perpendicular to a vertical middle plane running between the sheet-metal strips. An improved lateral mixing of the cooling water is rendered possible thereby. In particular, the cross section of the partial channel formed by a first corrugation decreases in the flow direction of the cooling water, and the cross section of the partial channel formed by a second, neighboring corrugation increases in this direction. In other words, the cross section of the partial channel formed by the first corrugation is greater at the inlet opening than at the outlet opening, and the cross section of the partial channel formed by the second corrugation is smaller at the inlet opening than at the outlet opening. This increase or decrease in the cross section can be affected continuously over the entire web height. However, forms of corrugation in which a middle region of constant cross section is present are also known. Owing to the mutually differing fashioning of the respectively neighboring corrugations, it is possible to produce a flow component directed perpendicular to the web plane by shaping the sheet-metal strips in a particularly simple way as regards production engineering. In a further advantageous refinement of the invention, the first and the second sheet-metal strips in each case have first and second corrugations that are alternately disposed in the longitudinal direction of the first and second sheet-metal strips. The first and second sheet-metal strips are assembled to form the web in such a way that each flow sub-channel is formed by a first and second corrugation. Such a sheet-metal strip and the spacer formed by it are easy to fabricate. In a preferred refinement of the invention, the flow sub-channel runs oblique, or at an inclination, to the vertical at least at its downstream end. In a further preferred embodiment, the cooling water, respectively emerging from the flow sub-channel of a web, which are mutually neighboring and inclined to a crossing point of two webs, has mutually opposed flow components perpendicular to the middle plane such that a swirl flow is produced around the crossing point, in which case, in particular, the swirl flows around mutually neighboring crossing points are respectively directed in an opposed fashion along a web. This prevents the occurrence of an overall torque produced by the swirl flows and acting on the entire fuel element. 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 spacer, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. Referring now to the figures of the drawing in detail and first, particularly, to FIGS. 1 and 2 thereof, there is shown a spacer that is constructed from a multiplicity of crisscrossing webs 2 that form a grid with polygonal meshes 4, square ones in the exemplary embodiment, through which fuel rods 5 are guided. Each web 2 is assembled from a first and second sheet-metal strip 6 and 8, respectively, that are welded to one another at their mutually touching upper and lower longitudinal edges. The first and second sheet-metal strips 6 and 8, respectively, are provided in each case with corrugations 10, 12 and 14, 16, respectively, that extend in each case into an interior of the mesh 4 respectively bounded by the sheet-metal strips 6 and 8. The corrugations 10, 12, 14, 16 serve simultaneously as bearings for the fuel rods 5 penetrating the meshes 4. In this way, there is formed between the corrugations 10, 14 and 12, 16 of the first and second sheet-metal strips 6, 8 respectively forming a web 2, a flow sub-channel 20 in which cooling water flows upward in a vertical direction through the spacer (out of the plane of the drawing). It is to be seen in FIGS. 1 and 2 that over their entire length in the web plane the flow sub-channels 20 are inclined to the vertical, that is to say inclined to a direction that runs perpendicular to the plane of the drawing. This inclination effects a deflection of the flow inclined to the vertical, but still parallel to the web plane, as before. Respectively neighboring flow sub-channels 20 of a web 2 have an opposing inclination. The four flow sub-channels 20 neighboring a crossing point P are oriented in this case such that two flow sub-channels 20 disposed in a common web 2 are inclined toward one another, while the two flow sub-channels 20 belonging to the other web 2 are inclined away from one another. Each flow sub-channel 20 has a shape that is asymmetric in relation to a middle plane 24 located between the sheet-metal strips 6, 8 and oriented perpendicular to the plane of the drawing. The corrugations 10, 16 are provided for this purpose in each case with a lower convex arch 101 and 161, respectively, such that at this location the corrugation 10 or 16 lies closer to the crossing point P. The corrugations 14 and 16 assigned respectively to the corrugations 10 and 12 therefore have upper convex arches 121 and 141, respectively, in their upper region, and so the cross-sectional area of the flow sub-channel 20 remains approximately the same over the entire height of the web 2. Because of the convex arches 101 and 161, respectively, at the inlet of the flow sub-channel 20, the partial channel 110 or 116 respectively formed by the corrugations 10, 16 has a larger cross-sectional area than the partial channel 112 or 114 respectively formed by the corrugations 12, 14. The partial channels 110, 116 therefore branch off a larger quantity of cooling water from the main channel formed by the mesh 4 than do the partial channels 112, 114. Since the cross sections of the partial channels 110, 116 narrow in the flow direction, and the cross sections of the partial channels 112, 114 widen, the cooling water flowing in the flow sub-channel 20 is displaced to the partial channels 112 and 114 and in this way acquires a flow component perpendicular to the web or middle plane 24. In other words, the asymmetric shaping of the corrugations 10, 14 and 12, 16, that is to say the offset configuration of the convex arches 101, 121, 141, 161, additionally lends the cooling water flowing between the corrugations 10, 14 and 12, 16 a velocity component perpendicular to the middle plane 24 of the web 2, since the cooling water experiences a deflection toward the convex arch 121 or 141. As an alternative to the configuration of the corrugations that is illustrated in FIGS. 1 to 3 and in the case of which the convex arches form an enlargement of the corrugation only in a direction of the middle plane 24 (web plane), it is also possible to provide convex arches that extend more deeply into the interior of the mesh 4, as is indicated by dashes, with the aid of a convex arch 200, in the right-hand lower mesh of FIG. 1, and to make better use of the space present in the corners and left free by the fuel rod 5. The effect of the convex arches 121 and 141 is then that, because of the velocity component perpendicular to the longitudinal direction 24, the cooling water flowing out of the flow sub-channel 20 is not directed straight onto the crossing point P but is directed past the latter obliquely. This produces a swirl flow around the crossing point P that leads to an improved heat transfer between the fuel rod and the fluid. Moreover, the corrugations 10, 12, 14, 16 are disposed in such a way that the direction of the swirl flow of respectively neighboring crossing points 22 is opposed. This prevents the torques respectively exerted by the swirl flows from adding up to produce an overall torque acting on the fuel assembly. In the exemplary embodiment, the corrugations 10, 12, 14, 16 fundamentally have the same shape. First and second sheet-metal strips 6 and 8, respectively, are, however, disposed rotated relative to one another about an axis perpendicular to the plane of the sheet-metal strip or middle plane 24. The shape of the flow sub-channels 20 in particular emerges clearly from the diagram of FIG. 3. It is clearly to be seen in FIG. 3 that the majority of the cooling water flowing in from below and branched out of the main channels of the flow sub-channel 20 is taken up by the partial channel 110 that is formed by the corrugation 10 that has a lower convex arch 101. Because of the cross-sectional narrowing of the partial channel 110, the cooling water, which flows upward oblique to the vertical (z direction) in the middle plane because of the flow sub-channel 20 running obliquely over its entire length is directed into the partial channel 114 of the neighboring corrugation 14 and thereby acquires, in addition to a velocity component vx directed toward the crossing point, a velocity component vy perpendicular thereto. As is shown in FIG. 4, the corrugations 10, 14 are respectively provided with longitudinal slots 26 in order to improve the mixing of the cooling liquid between the individual main channels, that is to say in order to increase the lateral mass flow. A further increase in the lateral mass flow can also be achieved by providing windows 28 at the crossing points P. Just as in the case of the known HTP spacer, the corrugations 10, 12, 14, 16 can still have, in the middle of the web 2, elongated convex arches that are on both sides of the slot 26, are orientated into the interior of the mesh 4 and, owing to their shaping, form a linear bearing for the fuel rod such that the latter is held resiliently in the mesh overall on eight lines. This is indicated in the exemplary embodiment in accordance with FIG. 5. Convex arches 30 are illustrated in this FIG. 5 on both sides of the slot 26. Moreover, in a departure from the exemplary embodiment illustrated in FIG. 3, the flow sub-channel 20 formed by the corrugations 14 and 10 is not inclined to the vertical z over its entire length l (=web height), but only over a part a of its length l at its downstream end and, if appropriate, over a part b of its length l upstream. In the remaining part l-a or l-a-b, the flow sub-channel 20 runs substantially parallel to the vertical z. It is possible in this way, given small mesh widths and large web heights (length l of the flow sub-channel 20), to produce a relatively large velocity component vx parallel to the web plane 24, that is to say a velocity component directed away from a crossing point P or toward a crossing point P. This application claims the priority, under 35 U.S.C. § 119, of German patent application No. 103 09 742.2, filed Mar. 6, 2003; the entire disclosure of the prior application is herewith incorporated by reference. |
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062529382 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method and apparatus for making focused and unfocused grids and collimators which are movable to avoid grid shadows on an imager, and which are adaptable for use in a wide range of electromagnetic radiation applications, such as x-ray and gamma-ray imaging devices and the like. More particularly, the present invention relates to a method and apparatus for making focused and unfocused grids, such as air core grids, that can be constructed with a very high aspect ratio, which is defined as the ratio between the height of each absorbing grid wall and the thickness of the absorbing grid wall, and that are capable of permitting large primary radiation transmission therethrough. 2. Description of the Related Art Anti-scatter grids and collimators can be used to eliminate the scattering of radiation to unintended and undesirable directions. Radiation with wavelengths shorter than or equal to soft x-rays can penetrate materials. The radiation decay length in the material decreases as the atomic number of the grid material increases or as the wavelength of the radiation increases. These grid walls, also called the septa and lamellae, can be used to reduce scattered radiation in ultraviolet, x-ray and gamma ray systems, for example. The grids can also be used as collimators, x-ray masks, and so on. For scatter reduction applications, the grid walls preferably should be two-dimensional to eliminate scatter from all directions. For many applications, the x-ray source is a point source close to the imager. An anti-scatter grid preferably should also be focused. Methods for fabricating and assembling focused and unfocused two-dimensional grids are described in U.S. Pat. No. 5,949,850, entitled "A Method and Apparatus for Making Large Area Two-dimensional Grids", referenced above. When an anti-scatter grid is stationary during the acquisition of the image, the shadow of the anti-scatter grid will be cast on the imager, such as film or electronic digital detector, along with the image of the object. It is undesirable to have the grid shadow show artificial patterns. The typical solution to eliminating the non-uniform shadow of the grid is to move the grid during the exposure. The ideal anti-scatter grid with motion will produce uniform exposure on the imager in the absence of any objects being imaged. One-dimensional grids, also known as linear grids and composed of highly absorbing strips and highly transmitting interspaces which are parallel in their longitudinal direction, can be moved in a steady manner in one direction or in an oscillatory manner in the plane of the grid in the direction perpendicular to the parallel strips of highly absorbing lamellae. For two-dimensional grids, the motion can either be in one direction or oscillatory in the plane of the grid, but the grid shape needs to be chosen based on specific criteria. The following discussion pertains to a two-dimensional grid with regular square patterns in the x-y plane, with the grid walls lined up in the x-direction and y-direction. If the grid is moving at a uniform speed in the x-direction, the film will show unexposed stripes along the x-direction, which also repeat periodically in the y-direction. The width of the unexposed strips is the same or essentially the same as the thickness of the grid walls. This grid pattern and the associated motion are unacceptable. If the grid is moving at a uniform speed in the plane of the grid, but at a 45 degree angle from the x-axis, the image on the film or imager is significantly improved. However, strips of slightly overexposed film parallel to the direction of the motion at the intersection of the grid walls will still be present. As the grid moves in the x-direction at a uniform speed, the grid walls block the x-rays everywhere, except at the wall intersection, for the fraction of the time EQU 2d/D, where d is the thickness of the grid walls and D is the periodicity of the grid walls. At the wall intersection, the grid walls blocks the x-rays for the fraction of the time EQU 2d/D<t.ltoreq.d/D, depending on the location. Thus, stripes of slightly overexposed x-ray film are produced. Methods for attempting to eliminate the overexposed strips discussed above are disclosed in U.S. Pat. Nos. 5,606,589, 5,729,585 and 5,814,235 to Pellegrino et al., the entire contents of each patent being incorporated herein by reference. These methods attempt to eliminate the overexposed strips by rotating the grid by an angle A, where A=atan(n/m), and m and n are integers. However, these methods are unacceptable or not ideal for many applications. Accordingly, a need exists for a method and apparatus for eliminating the overexposed strips associated with two-dimensional focused or unfocused grid intersections. SUMMARY OF THE INVENTION An object of the present invention, therefore, is to provide a method and apparatus for manufacturing a focused or unfocused grid which is configured to minimize overexposure at its wall intersections when the grid is moved during imaging. Another object of the present invention is to provide a method and apparatus for moving a focused or unfocused grid so that no perceptible areas of variable density are cast by the grid onto the film or other two-dimensional electronic detectors. A further object of the present invention is to provide a method and apparatus for assembling sections of a two-dimensional, focused or unfocused grid. Still another object of the present invention is to provide a method and apparatus for joining stacked layers of two-dimensional focused or unfocused grids. These and other objects of the present invention are substantially achieved by providing a grid, adaptable for use with electromagnetic energy emitting devices, comprising at least metal layer, formed by electroplating. The grid comprises top and bottom surfaces, and a plurality of solid integrated walls. Each of the solid integrated walls extends from the top to bottom surface and having a plurality of side surfaces. The side surfaces of the solid integrated walls are arranged to define a plurality of openings extending entirely through the layer. For some applications, all the walls are 90.degree. with respect to the top and bottom surfaces. For some other applications, at least some of the walls extend at an angle other than 90.degree. with respect to the top and bottom surfaces such that the directions in which the walls extend all converge at a point in space at a predetermined distance from the front surface of the at least one layer. These and other objects of the present invention are also substantially achieved by providing a grid, adaptable for use with electromagnetic energy emitting devices. The grid comprises at least one solid metal layer, formed by electroplating. The solid metal layer comprises top and bottom surfaces, and a plurality of solid integrated, intersecting walls, each of which extending from the top to bottom surface and having a plurality of side surfaces. The side surfaces of the walls are arranged to define a plurality of openings extending entirely through the layer, and at least some of the side surfaces have projections extending into respective ones of the openings. |
claims | 1. A system for collecting metal in an electrorefining process, the system comprising:a. a hollow cathode; andb. a container defining an upwardly extending surface adapted to slidably communicate with interior aspects of the hollow cathode;wherein the container encircles the bottom and laterally facing surfaces of the cathode. 2. The system as recited in claim 1 wherein the extending surface is coaxial with the longitudinal axis of the container and of the cathode. 3. The system as recited in claim 1 wherein the cathode does not contact the container. 4. The system as recited in claim 1 wherein the cathode is not in electrical communication with the container. 5. The system as recited in claim 1 adapted to contain metal in two annular spaces between the cathode and the container. 6. The system as recited in claim 1 wherein the cathode and the container are adapted to be imbedded in molten salt in a first position. 7. The system as recited in claim 6 wherein the container is adapted to be imbedded in molten salt in a second position. 8. The system as recited in claim 1, wherein annular spaces defined between the cathode and the container are adapted to retain metal formed during the electrorefining process, while displacing an electrolyte out of the annular spaces. 9. A system for collecting metal in an electrorefining process, the system comprising:a. a rotatable, hollow cathode;b. a container defining an upwardly extending surface adapted to oppose interior surfaces of the hollow cathode;c. a first plate adapted to support the container;d. a second plate superior from the first plate and connected thereto via a plurality of vertically extending struts; ande. a cathode-scraper third plate disposed between the first plate and the second plate and attached to the struts, each of the first, second, and third plates defining a central aperture to slidably receive the cathode, wherein the cathode is collinearly aligned with the apertures of the first, second and third plates. 10. The system as recited in claim 9 further comprising:f) radially extending protuberances on the surface of the cathode; andg) diametrically opposed notches formed in the central aperture of the third plate, said notches in slidable communication with the protuberances upon rotation of the cathode. 11. The system as recited in claim 9 wherein the container defines a first annular region and a central protuberance coaxial with the annular region. 12. The system as recited in claim 11 wherein the cathode defines a tube adapted to receive the central protuberance of the container to define a second, inner annular region between the central protuberance and in the interior surfaces of the hollow cathode. 13. The system as recited in claim 12, wherein the first and second annular regions are adapted to retain metal formed during the electrorefining process, while displacing an electrolyte from the annular regions. |
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047160063 | abstract | A method of operating a pressurized water nuclear reactor comprising determining the present core power and reactivity levels and predicting the change in such levels due to displacer rod movements. Groups or single clusters of displacer rods can be inserted or withdrawn based on the predicted core power and reactivity levels to change the core power level and power distribution thereby providing load follow capability, without changing control rod positions or coolant boron concentrations. |
description | The present invention relates to the field of ensuring, enhancing and maintaining safety in safety critical systems. Safety critical systems, such as, for example, nuclear power stations and civilian aircraft are designed to safety standards. Safety standards may be set by national or international regulators, standard-setting bodies or certification agencies, for example. Safety standards may be defined for industries as a whole, for system classes or for individual systems, for example. Even in the absence of formal safety standards, equipment in a system may be designed with safety rules when this is seen as desirable, for example to protect biodiversity. A fission reactor, for example, must be designed and constructed in a way that enables operators to control its functioning. Such controlling may comprise, if necessary, causing the reactor to transition to a managed idle state when instructed. Such an idle state may comprise a state where fission reactions are subcritical and decay heat is removed from the reactor core to prevent its overheating, which might otherwise damage the core of the reactor, potentially leading to release of radionuclides. A civilian aircraft, on the other hand, may be safely operated only in case the aircraft can be reliably flown even when aircraft systems develop fault conditions. For example, in case a flight computer develops a fault, pilots must be able to continue providing meaningful control inputs to the aircraft to continue its safe flight. To obtain safe operability in safety-critical systems, components comprised in such systems may be associated with safety conditions. For example, a flight computer may be made redundant, wherein an aircraft may be furnished with a plurality of flight computers, each individually being capable of controlling the flight. In this case, redundancy is a safety condition associated with the flight computer. In case of a fault condition in one of the flight computers, another one of the flight computers may assume the task of controlling the flight, the faulty flight computer being set to an inactive state. In conventional systems, in case a component is replaced and the component is associated with a safety condition, then the replacement component becomes associated with the safety condition as well. This occurs since the safety condition operates on the component level. In some systems, replacing a component may be constrained to a replacement part that to a maximum extent possible resembles the replaced part. The invention is defined by the features of the independent claims. Some specific embodiments are defined in the dependent claims. According to a first aspect of the present invention, there is provided a method, comprising defining a task category information element, the task category information element being associated with at least one functional requirement and at least one design principle, associating the task category information element with at least one architecture definition information element, associating each of the at least one architecture definition information element with at least one system-level information element, and verifying the system described by the at least one architecture definition information element and associated system-level information elements is compliant with the at least one design principle. Various embodiments of the first aspect may comprise at least one feature from the following bulleted list: each system-level information element is associated with at least one equipment information element the method further comprises defining a second task category information element, and associating the second task category information element with at least one architecture definition information element the at least one functional requirement is comprised in the following list: reactivity control, core cooling, confinement of radioactive substances, controlling flight, communications, preventive protection, reactor protection, automatic back-up, manual back-up, manual accident management, back-up of manual accident management, safe shut-down at least one of the at least one architecture definition information element is comprised in the following list: functional architecture definition information element, automation architecture definition information element, process and electrical architecture definition information element, control room architecture definition information element and layout architecture definition information element at least one of the at least one system-level information element is comprised in the following list: preventive protection system information element, preventive actuation and indication system information element, safety injection system information element, emergency power supply system information element, digital HMI system information element and automatic backup system information element the method further comprises storing the task category information element, each of the at least one architecture definition information element and each of the at least one system-level information element in a database system the verified system comprises a nuclear power station or an aircraft the method is performed, at least in part, in the database system. According to a second aspect of the present invention, there is provided a method, comprising recording a first change in an information element in a database system, the database system storing a task category information element, at least one architecture definition information element and at least one system-level information element, and verifying the system described by the at least one architecture definition information element and system-level information elements associated with the at least one architecture definition information element is compliant with at least one design principle, wherein the task category information element is associated with at least one functional requirement and at the least one design principle. Various embodiments of the second aspect may comprise at least one feature from the following bulleted list: the database further comprises at least one equipment information element the task category information element is associated with the at least one architecture definition information element and each of the at least one architecture definition information element is associated with at least one of the at least one system-level information element responsive to the verification indicating the system does not comply with the at least one design principle, the method comprises recording a second change in the database system and performing a second verification as to whether the system complies with the at least one design principle the second change does not modify the same information element as the first change at least one of the at least one architecture definition information element is comprised in the following list: functional architecture definition information element, automation architecture definition information element, process and electrical architecture definition information element, control room architecture definition information element and layout architecture definition information element at least one of the at least one system-level information element is comprised in the following list: preventive protection system information element, preventive actuation and indication system information element, safety injection system information element, emergency power supply system information element, digital HMI system information element and automatic backup system information element the at least one design principle: comprises at least one of the following: redundancy, diversity, separation, isolation, quality level, reliability level, seismic qualification and environmental condition qualification the at least one functional requirement is comprised in the following list: reactivity control, core cooling, confinement of radioactive substances, controlling flight, communications, preventive protection, reactor protection, automatic back-up, manual back-up, manual accident management, back-up of manual accident management, safe shut-down. According to a third aspect of the present invention, there is provided a database system, comprising a task category database configured to store at least one task category information element, each task category information element being associated with at least one functional requirement and at least one design principle, an architecture database configured to store at least one architecture definition information element, a system database configured to store at least one system-level information element, wherein each of the at least one task category information element is associated with at least one architecture definition information element and each of the at least one architecture definition information element is associated with at least one system-level information element. Various embodiments of the third aspect may comprise at least one feature from the following bulleted list: an equipment database configured to store at least one equipment information element the task category database is interfaced with the architecture database via a first database relation, and the architecture database is interfaced with the system database via a second database relation the system database is interfaced with the equipment database via a third database relation at least one of the at least one architecture definition information element is comprised in the following list: functional architecture definition information element, automation architecture definition information element, process and electrical architecture definition information element, control room architecture definition information element and layout architecture definition information element least one of the at least one system-level information element is comprised in the following list: preventive protection system information element, preventive actuation and indication system information element, safety injection system information element, emergency power supply system information element, digital HMI system information element and automatic backup system information element the at least one design principle comprises at least one of the following: redundancy, diversity, separation, isolation, quality level, reliability level, seismic qualification and environmental condition qualification the at least one functional requirement is comprised in the following list: core, cooling, controlling flight and communications the database system stores sequence information elements, each sequence information element describing a sequence of actions, each sequence information element being associated with a triggering event and each sequence information element being associated with a task category information element. According to a fourth aspect of the present invention, there is provided a computerized nuclear power station monitoring system, comprising a memory configured to store a database comprising a task category database comprising plurality of task category information elements comprising a safety function information element and a reactor protection information element, each task category information element being associated with at least one technical functional requirement and at least one technical design principle, each technical design principle being comprised in a technical design principle list, the technical design principle list comprising redundancy, diversity, separation and isolation, each functional requirement being comprised in a functional requirement list, the functional requirement list comprising reactivity control, core cooling and safe shut-down, and an equipment database configured to store at least one equipment information element, and at least one processor configured to, responsive to receipt in the computerized monitoring system of a failure notification concerning a first equipment information element, determine, using the database, a set comprising each technical design principle associated with each task category information element associated, via database relations, with the first equipment information element, and to identify, based on each technical design principle comprised in the set, a technical constraint of an action compensating, at least partly, effects of the failure identifier in the failure notification. Various embodiments of the fourth, sixth or eighth aspect may comprise at least one feature from the following bulleted list: the at least one processor is configured to determine a constraint of increased unit count responsive to the set comprising the technical design principle redundancy the at least one processor is configured to determine a constraint of principle of action responsive to the set comprising the technical design principle diversity the at least one processor is configured to determine a constraint of location responsive to the set comprising the technical design principle separation the at least one processor is configured to determine a constraint of physical separation responsive to the set comprising the technical design principle isolation the at least one processor is further configured to provide an indication of the determined constraints. According to a fifth aspect of the present invention, there is provided a method in a computerized nuclear power station monitoring system, comprising storing a database comprising a task category database comprising a plurality of task category information elements comprising a safety function information element and a reactor protection information element, each task category information element being associated with at least one technical functional requirement and at least one technical design principle, each technical design principle being comprised in a technical design principle list, the technical design principle list comprising redundancy, diversity, separation and isolation, each functional requirement being comprised in a functional requirement list, the functional requirement list comprising reactivity control, core cooling and safe shut-down, and an equipment database configured to store at least one equipment information element, determining, responsive to receipt in the computerized monitoring system of a failure notification concerning a first equipment information element, using the database, a set comprising each technical design principle associated with each task category information element associated, via database relations, with the first equipment information element, and identifying, based on each technical design principle comprised in the set, a technical constraint of an action compensating, at least partly, effects of the failure identifier in the failure notification. According to a sixth aspect of the present invention, there is provided a computerized communication network monitoring system, comprising a memory configured to store a database comprising a task category database comprising plurality of task category information elements comprising a switching failure elimination information element and a privacy information element, each task category information element being associated with at least one technical functional requirement and at least one technical design principle, each technical design principle being comprised in a technical design principle list, the technical design principle list comprising redundancy, diversity, separation and isolation, each functional requirement being comprised in a functional requirement list, the functional requirement list comprising subscription uniqueness, data encryption, subscriber identification and radio resource management, and an equipment database configured to store at least one equipment information element, and at least one processor configured to, responsive to receipt in the computerized monitoring system of a failure notification concerning a first equipment information element, determine, using the database, a set comprising each technical design principle associated with each task category information element associated, via database relations, with the first equipment information element, and to identify, based on each technical design principle comprised in the set, a technical constraint of an action compensating, at least partly, effects of the failure identifier in the failure notification. According to a seventh aspect of the present invention, there is provided a method in a computerized communication network monitoring system, comprising storing a database comprising a task category database comprising a plurality of task category information elements comprising a switching failure elimination information element and a privacy information element, each task category information element being associated with at least one technical functional requirement and at least one technical design principle, each technical design principle being comprised in a technical design principle list, the technical design principle list comprising redundancy, diversity, separation and isolation, each functional requirement being comprised in a functional requirement list, the functional requirement list comprising subscription uniqueness, data encryption, subscriber identification and radio resource management, and an equipment database configured to store at least one equipment information element, determining, responsive to receipt in the computerized monitoring system of a failure notification concerning a first equipment information element, using the database, a set comprising each technical design principle associated with each task category information element associated, via database relations, with the first equipment information element, and identifying, based on each technical design principle comprised in the set, a technical constraint of an action compensating, at least partly, effects of the failure identifier in the failure notification According to an eighth aspect of the present invention, there is provided a computerized power distribution network monitoring system, comprising a memory configured to store a database comprising a task category database comprising a plurality of task category information elements comprising a frequency control information element and a safe installation information element, each task category information element being associated with at least one technical functional requirement and at least one technical design principle, each technical design principle being comprised in a technical design principle list, the technical design principle list comprising redundancy, diversity, separation and isolation, each functional requirement being comprised in a functional requirement list, the functional requirement list comprising production ramp-up, demand ramp-up and electrical isolation, and an equipment database configured to store at least one equipment information element, and at least one processor configured to, responsive to receipt in the computerized monitoring system of a failure notification concerning a first equipment information element, determine, using the database, a set comprising each technical design principle associated with each task category information element associated, via database relations, with the first equipment information element, and to identify, based on each technical design principle comprised in the set, a technical constraint of an action compensating, at least partly, effects of the failure identifier in the failure notification. According to a ninth aspect of the present invention, there is provided a method in a power distribution network monitoring system, comprising storing a database comprising a task category database comprising a plurality of task category information elements comprising a frequency control information element and a safe installation information element, each task category information element being associated with at least one technical functional requirement and at least one technical design principle, each technical design principle being comprised in a technical design principle list, the technical design principle list comprising redundancy, diversity, separation and isolation, each functional requirement being comprised in a functional requirement list, the functional requirement list comprising production ramp-up, demand ramp-up and electrical isolation, and an equipment database configured to store at least one equipment information element, determining, responsive to receipt in the computerized monitoring system of a failure notification concerning a first equipment information element, using the database, a set comprising each technical design principle associated with each task category information element associated, via database relations, with the first equipment information element, and identifying, based on each technical design principle comprised in the set, a technical constraint of an action compensating, at least partly, effects of the failure identifier in the failure notification. According to a tenth aspect of the present invention, computer programs are provided to cause a method in accordance with the second, fifth, seventh or ninth aspect to be performed. At least some embodiments of the present invention find application in optimizing safety critical systems such as nuclear power generation and/or aircraft systems. Further examples of suitable systems include communication networks and power distribution networks. By assigning design principles to functional requirements, more efficient implementation and maintenance of safety critical systems may be obtained. Where design principles, such as redundancy or diversity, are assigned to individual equipment rather than higher-level functional requirements, over-implementation or degradation of a safety level may occur and/or refitting existing safety critical systems may be more constrained by equipment-specific requirements. By assigning design principles to functional task categories rather than individual equipment, more flexible implementation is enabled. FIG. 1 illustrates an example system capable of supporting at least some embodiments of the present invention. The system of FIG. 1 is a nuclear power station operating a fission-based reactor, although in other embodiments of the invention, other kinds of systems may be envisioned. The system of FIG. 1 comprises building 100, which houses reactor 110. Building 100 is arranged to draw water for cooling from source 300, which may comprise an ocean, lake, river or other stable source of cooling water, for example. The system of FIG. 1 further comprises building 200, which houses systems not housed in building 100. A nuclear power station may comprise a large number of systems, a subset of which is illustrated in FIG. 1 to serve the purpose of illustrating the principles underlying the present invention. Systems comprised in a safety critical system may embody at least one design principle, such as for example a safety-related design principle. Examples of design principles include redundancy, diversity, separation, isolation, quality level, reliability level, seismic qualification and environmental condition qualification. System 120A, which may comprise, for example a pump system, has a redundant system 120B. In other words, system 120A and system 120B are similar and enabled to perform a similar task. Either one, system 120A or system 120B, may alone be capable of performing its task. Systems 120A and 120B may be configured to operate on the same, or similar, principles of action. In general where a similar redundant system or equipment is provided for a given system or equipment, this system or equipment is said to embody redundancy. A system embodying redundancy is more dependable than a system without redundancy, as a redundancy-embodying system can continue operation in case one system develops a fault, since the faulty system, for example system 120A, may be switched off and the task may be assigned to the redundant system, for example system 120B. System 130A, which may comprise, for example, a safety system, has a hard-wired backup diversity system 130B. In other words, system 130A and system 130B are enabled to perform a similar task. Either one, system 130A or system 130B, may alone be capable of performing its task. Systems 130A and 130B are configured to operate on different principles of action. Since systems 130A and 130B are configured to operate using different principles of action, they are less likely to fail at the same time as a response to an unusual operating condition. For example where these systems comprise safety systems, they may be based on different physical processes having the same overall functional specifications. In other words, designs may be developed independently for system 130A and system 130B. Such independent development may comprise using different design teams, subcontractors, materials and/or principles of action, for example. As a consequence, if system 130A encounters an error in a certain unusual operating condition of the nuclear power station of FIG. 1, it is unlikely that system 130B encounters an error in the same operating condition. In this case, responsibility can be re-assigned from system 130A to system 130B, to obtain uninterrupted and secure operation of the power station. In general, where a system or equipment embodies diversity this system or equipment may be seen to comprise more than one subsystem, the subsystems being configured to operate on different principles and each being capable of performing the task of the system or equipment. Herein the term “system” may generally be used to refer to an equipment, system, architecture or installation. System 140A, which may comprise, for example, a control system, has a diversity system 140B. System 140A and system 140B are enabled to perform a similar task. Either system 140A or system 140B may alone be capable of performing its task. System 140B, which may operate based on a same or a different principle as system 140A, is housed in building 200 while system 140A is housed in building 100. That the systems are housed in different buildings, or more generally separate from each other, means the systems embody a separation design principle. Situating the systems separately from each other increases the dependability of the aggregate system comprising system 140A and system 140B, since a problem affecting, say, building 200 may leave building 100 and systems housed therein unaffected. Additionally or alternatively to physical separation, systems may be separated electrically and/or functionally. The intent in separation overall is to avoid failures from progressing from a system to its back-up system, or from one task category to another task category. Electrical separation, for example, may be accomplished by either not connecting the separated systems to each other electrically, or by suitably filtering electrical connections arranged between the systems. Examples of suitable filtering include over-voltage protection, current protection and fibre-optic filters. Where system 140A and system 140B are based on the same, or a similar, operating principle the system comprising system 140A and system 140B may be considered to embody separation and redundancy. Where system 140A and system 140B are based on different operating principles the system comprising system 140A and system 140B may be considered to embody separation and diversity. A system embodying the design principle isolation may comprise a system wherein the system, including equipment comprised in the system, is isolated from its surroundings. For example, being disposed inside a reactor containment vessel and/or hardened building provides isolation. Isolation may be defined in various ways, for example, ability to withstand an impact of a passenger aircraft and/or ability to contain a molten reactor core. Among further design principles, a quality level may comprise that a system embodying that design principle meets a standardized quality level. Further, a reliability level, a seismic qualification level and an environmental condition qualification are examples of design principles that may be embodied by systems comprised in safety critical systems. When designing, maintaining or refitting a safety critical system, it may be advantageous to associate design principles with functional requirements. This association may take place in a task category, which may comprise a database structure, such as an information element, which comprises or is associated with both the functional requirement and at least one design principle. The task category may be associated with hierarchically lower levels of a design in such a way that the design principles associated with the task category are embodied by the aggregate system that performs the functional requirement associated with the task category. The functional requirement associated with the task category may be referred to simply as the functional requirement of the task category. At least in some embodiments, a task category information element does not define structure but is associated, directly or indirectly, with information elements that do define structure. Examples of information element types that define structure include an architecture definition information element and a system-level information element. When a design principle is associated with a task category, implementing systems to perform the functional requirement of the task category becomes more flexible, allowing more intelligent implementation that may result in a simpler and safer. Requiring that each system and equipment in the system performing the functional requirement separately comply with the design principle is a more restrictive model, where equipment may be duplicated excessively. For example, where an equipment, such as for example a pump, is comprised in a system that performs a functional requirement of a task category, it may be assigned another role in a system that performs a functional requirement of another task category. The pump, for example, may embody diversity with respect to more than one system or task category. In general, a system may embody a design principle with respect to more than one system and/or task category. In a complex system such as a nuclear power station, or an aircraft, the number of systems and equipments may be very large. To enable use of task categories and associated design principles, a database system may be employed. By database system it is herein meant a physical system configured to store a database, by which it is in turn meant an organized storage or assembly of information elements, which may be interrelated within the database system via associations and/or database relations. The database system may be use a suitable system, such as for example a computer system, and suitable magnetic, solid-state, holographic or other kind of memory. Such a database is illustrated in FIG. 1 as database 150. In the database, functional requirements and design principles may take the form of information elements. A communication network may be considered a further example of a safety critical system, as it may be employed to communicate critical information, such as emergency telephone calls, radar data and/or environmental sensor network data. Examples of communication networks include cellular and fixed networks. A further example of a safety critical system is a power arranged to convey electrical energy from generating stations to industry and consumers. In database 150, information elements may be arranged in a hierarchical structure which is illustrated in FIG. 2. FIG. 2 illustrates an example database hierarchy in accordance with at least some embodiments of the present invention. The example database hierarchy of FIG. 2 involves a database hierarchy of a nuclear power station. At the top level are disposed nuclear safety design requirements 210. These requirements may be derived from and/or be based on regulatory requirements, codes and/or standards. In some embodiments, nuclear safety design requirements 210 are absent from the database, for example where their content is taken into account, implicitly or explicitly, in other layers. The requirements of requirement layer 210 may be associated with, or comprised in, task categories in layer 220, which corresponds to the level of the entire nuclear power station. Layer 220 may be termed the plant layer. Each task category may be associated with at least one design principle and at least one functional requirement, as described above. In detail, each task category may, in some embodiments, be associated with one and only one functional requirement and at least one design principle. Task categories included in the example of FIG. 2 are task categories 220A, 220B and 220C. In the database, task categories may be present as task category information elements. Examples of task categories comprise preventive safety function, reactor protection and automatic back-up. Examples of functional requirements comprise accident frequency, pollutant leakage rate, maintenance intervals and emergency landing frequency of an aircraft. Under plant layer 220 is disposed architecture layer 230. Layer 230 may comprise architecture definition information elements, at least some such information elements being associated with at least one task category information element on level 220. The architecture definition information elements may comprise indications as to the way in which the architecture therein defined contributes to embodiment of the design principles associated with associated task categories. In other words, the architecture definition information elements may comprise information as to how the design principles of the higher-level task categories are implemented in the architecture level. Architecture definition information elements included in the example of FIG. 2 are architecture definition information elements 230A, 230B and 230C. Examples of architectures include functional architecture, automation architecture, process and electrical architecture, control room architecture and layout architecture. Functional architecture may describe how the main processes of the plant are implemented. Automation architecture may describe how automation mechanisms are arranged in the plant. Process and electrical architecture may describe how processes and electrical systems are designed on the high level. Control room architecture may describe how the control room is arranged to control functioning of the plant, and layout architecture may describe how systems of the plant are distributed among buildings comprised in the plant. Under architecture layer 230 is disposed system layer 240. System layer 240 may comprise system-level information elements, each such information element being associated with at least one architecture definition information element on level 230. The system-level information elements may comprise indications as to the way in which the systems therein defined contribute to embodiment of the design principles associated with associated task categories, wherein task categories are associated with system-level information elements via architecture definition information elements on architecture level 230. System-level information elements included in the example of FIG. 2 are system-level information elements 240A, 240B, 240C and 240D. Examples of systems include preventive protection system, preventive actuation and indication system, safety injection system, emergency power supply system, digital human-machine interface, HMI, system and automatic backup system. Under system layer 240 is disposed equipment layer 250. Equipment layer 250 may comprise equipment-level information elements, each such information element being associated with at least one system-level information element on level 240. The equipment-level information elements may comprise indications as to the way in which the equipment therein defined contribute to embodiment of the design principles associated with associated task categories, wherein task categories are associated with equipment-level information elements via system-level information elements on system layer 240 and architecture definition information elements on architecture level 230. Equipment-level information elements are illustrated in FIG. 2 collectively as elements 250A. In some embodiments, regulatory requirements may be assigned to individual systems or pieces of equipment. Such system-level or equipment-level regulatory requirements may be recorded in system-level or equipment-level information elements and used as additional constraints in implementing methods in accordance with the present invention. Using the hierarchical database system described above it can be determined, which pieces of equipment in the plant contribute to which plant-level functional requirements and design principles. As a consequence, when a piece of equipment is replaced with a new type of equipment, it can be assessed, what the implications are for the plant or aircraft overall in terms of design principles. For example, where a computer is replaced with a new kind of computer, for example, a computer based on a complex instruction set computing, CISC, processor is replaced with a computer based on a reduced instruction set computing, RISC, processor, the new computer may be able to perform as a diversity computer to an already present CISC computer in the plant. In this case, installing the RISC computer to replace an older computer may enable removal of a further computer from the plant, the diversity role of the further computer being thereafter performed by the new computer. The further computer may be comprised in a different system or architecture, and it may be associated with a different task category than the old computer the new RISC computer replaces, for example. Running the database system as described above may at least in part automate such design considerations of the safety critical system. Each equipment-level information element storing or being associated with information describing each role the described equipment performs in the plant, a user may interact with the database system to identify whether replacing the piece of equipment with a new piece of equipment enables a simplification in the overall system, or whether characteristics of the new piece of equipment necessitate a further modification to the overall system. A further modification may be necessary where, for example, the new piece of equipment is unable to perform a role the previous piece of equipment performed, for example as a redundancy or diversity element to a further piece of equipment, which may be comprised in a different system or architecture. Thus instead of assigning requirements to individual pieces of equipment, design principles are associated with task categories to enable smart plant management and selection of replacement pieces of equipment in such a way that the plant overall may be simplified. A simplified plant provides the technical effect that running it consumes less energy, for example. Further modification may be required in cases where regulatory requirements have changed between initial construction and refit of a safety critical system. Similarly, using the database system enables a fuller understanding of fault conditions, since the database system identifies the roles each piece of equipment registered therein performs. Therefore, when a piece of equipment develops a fault, it can be identified, using the database system, which other systems have less redundancy as a consequence of the fault. FIG. 3 illustrates an example apparatus capable of supporting at least some embodiments of the present invention. Illustrated is device 300, which may comprise, for example, a device such as database system 150 of FIGURE. Comprised in device 300 is processor 310, which may comprise, for example, a single- or multi-core processor wherein a single-core processor comprises one processing core and a multi-core processor comprises more than one processing core. Processor 310 may comprise a Xeon or Opteron processor, for example. Processor 310 may comprise more than one processor. A processing core may comprise, for example, a Cortex-A8 processing core manufactured by ARM Holdings or a Bulldozer processing core produced by Advanced Micro Devices Corporation. Processor 310 may comprise at least one application-specific integrated circuit, ASIC. Processor 310 may comprise at least one field-programmable gate array, FPGA. Processor 310 may be means for performing method steps in device 300. Processor 310 may be configured, at least in part by computer instructions, to perform actions. Device 300 may comprise memory 320. Memory 320 may comprise random-access memory and/or permanent memory. Memory 320 may comprise at least one RAM chip. Memory 320 may comprise magnetic, optical and/or holographic memory, for example. Memory 320 may be configured to store information elements of a database system, for example. Memory 320 may be at least in part accessible to processor 310. Memory 320 may be means for storing information. Memory 320 may comprise computer instructions that processor 310 is configured to execute. When computer instructions configured to cause processor 310 to perform certain actions are stored in memory 320, and device 300 overall is configured to run under the direction of processor 310 using computer instructions from memory 320, processor 310 and/or its at least one processing core may be considered to be configured to perform said certain actions. Memory 320 may be at least in part comprised in processor 310. Memory 320 may be at least in part external to device 300 but accessible to device 300. Device 300 may comprise a transmitter 330. Device 300 may comprise a receiver 340. Transmitter 330 and receiver 340 may be configured to transmit and receive, respectively, information in accordance with at least one communication standard. Transmitter 330 may comprise more than one transmitter. Receiver 340 may comprise more than one receiver. Transmitter 330 and/or receiver 340 may be configured to operate in accordance with wireless local area network, WLAN, Ethernet and/or worldwide interoperability for microwave access, WiMAX, standards, for example. Device 300 may comprise user interface, UI, 360. UI 360 may comprise at least one of a display, a keyboard and a touchscreen. A user may be able to operate device 300 via UI 360, for example to interact with a database system comprised in, or controlled by, device 300. Device 300 may comprise or be arranged to accept a user identity module 370. User identity module 370 may comprise, for example, a secure element. A user identity module 370 may comprise cryptographic information usable to verify the identity of a user of device 300 and/or to facilitate encryption and decryption of database contents. Processor 310 may be furnished with a transmitter arranged to output information from processor 310, via electrical leads internal to device 300, to other devices comprised in device 300. Such a transmitter may comprise a serial bus transmitter arranged to, for example, output information via at least one electrical lead to memory 320 for storage therein. Alternatively to a serial bus, the transmitter may comprise a parallel bus transmitter. Likewise processor 310 may comprise a receiver arranged to receive information in processor 310, via electrical leads internal to device 300, from other devices comprised in device 300. Such a receiver may comprise a serial bus receiver arranged to, for example, receive information via at least one electrical lead from receiver 340 for processing in processor 310. Alternatively to a serial bus, the receiver may comprise a parallel bus receiver. Processor 310, memory 320, transmitter 330, receiver 340, UI 360 and/or user identity module 370 may be interconnected by electrical leads internal to device 300 in a multitude of different ways. For example, each of the aforementioned devices may be separately connected to a master bus internal to device 300, to allow for the devices to exchange information. However, as the skilled person will appreciate, this is only one example and depending on the embodiment various ways of interconnecting at least two of the aforementioned devices may be selected without departing from the scope of the present invention. FIG. 4 illustrates an example database structure in accordance with at least some embodiments of the present invention. Layer 410 corresponds in terms of FIG. 2 to the task category layer, storing task category information elements. Layer 420 corresponds in terms of FIG. 2 to the architecture layer, storing architecture definition information elements. Layer 430 corresponds in terms of FIG. 2 to the system layer, storing system-level information elements. Finally, layer 440 corresponds in terms of FIG. 2 to the equipment layer, storing equipment-level information elements. A database relation layer may be disposed between layer 410 and layer 420, between layer 420 and layer 430, and/or between layer 430 and layer 440. FIG. 5 is a first flow chart of a first method in accordance with at least some embodiments of the present invention. The phases of the illustrated method may be performed in database 150 of FIG. 1 or on device 300 of FIG. 3, for example. Phase 510 comprises defining a task category information element, the task category information element being associated with at least one functional requirement and at least one design principle. Phase 520 comprises associating the task category information element with at least one architecture definition information element. Phase 530 comprises associating each of the at least one architecture definition information element with at least one system-level information element. Finally, phase 540 comprises verifying the system described by the at least one architecture definition information element and associated system-level information elements is compliant with the at least one design principle. FIG. 6 is a second flow chart of a second method in accordance with at least some embodiments of the present invention. The phases of the illustrated method may be performed in database 150 of FIG. 1 or on device 300 of FIG. 3, for example. Phase 610 comprises recording a first change in an information element in a database system, the database system storing a task category information element, at least one architecture definition information element and at least one system-level information element. Phase 620 comprises verifying the system described by the at least one architecture definition information element and system-level information elements associated with the at least one architecture definition information element is compliant with at least one design principle, wherein the task category information element is associated with at least one functional requirement and at the least one design principle. In general, there is provided a method, comprising defining a task category information element, the task category information element being associated with at least one functional requirement and at least one design principle, associating the task category information element with at least one architecture definition information element, associating each of the at least one architecture definition information element with at least one system-level information element, and verifying the system described by the at least one architecture definition information element and associated system-level information elements is compliant with the at least one design principle. The method may be performed using a database system, for example. The associating phases comprised in the method may comprise defining information element association properties in the database system. The verifying may comprise running a verification algorithm on the information elements comprised in the database system. The verifying may comprise checking that for each design principle, the architecture, systems and pieces of equipment associated with the design principle together embody the design principle. The verifying does not, in some embodiments, require that each information element directly or indirectly associated with the design principle embodies the design principle. For example, where the design principle comprises redundancy, not all pieces of equipment directly or indirectly associated with a task category associated with redundancy need be made redundant in the sense of installing duplicate pieces of equipment. In these embodiments, it suffices that the function defined by the associated information elements as a whole is redundant. In other words, should any individual piece of equipment comprised in this function fail, its purpose may be served by another piece of equipment which need not be identical to it, and need not be comprised in the function in question. The safety critical system may comprise systems or pieces of equipment that do not need redundancy or diversity, for example. The information elements describing these pieces of equipment may comprise information indicating the way in which the design principle is implemented with respect to the functions of these pieces of equipment. In at least some embodiments, the database system stores sequence information elements, each sequence information element describing a sequence of actions, each sequence information element being associated with a triggering event and each sequence information element being associated with a task category information element. The sequence may control the consequences of the occurrence of the event. For example, an event may comprise a point failure in a system or an interruption in cooling water supply, and the sequence of actions may comprise a pre-planned response to the event whereby the consequences of the event are controlled. FIG. 7 illustrates example design verification in accordance with at least some embodiments of the present invention. Of the W-shaped FIG. 7, the left-most prong corresponds to architecture of a power plant, the mid prong corresponds to verification of the plant as designed, and finally the right-most prong corresponds to verification of a completed, built plant. Unit 710 is an accident management plan, which is verified in a plan for accident management plan review 7100, leading to a preliminary independent review 710A in the as-designed phase. A corresponding final independent review 710B is conducted in the as-built phase. Unit 720 is a functional architecture, which forms a basis for other architectures. Functional architecture 720 can be divided, for example, into short term and long term event progression architectures, describing plant specific safety functions on a general level. Functional architecture is verified in a review 7200, resulting in a preliminary safety analysis 720A in the as-designed phase and a final safety analysis 720B in the as-built phase. Unit 730 is an automation architecture, which may describe how task category functional requirements and/or safety functions are divided to automation systems so that design principles are met. Automation architecture is verified in a review 7300, resulting in an interface analysis 730A in the as-designed phase and interconnected tests 730B in the as-built phase. The interconnected tests 730B may be derived in part jointly with final safety analysis 720B. Unit 740 is a control room and procedures architecture, which may describe safety and ergonomics requirements for human-machine interfaces, HMI, control room and procedures. Control room and procedures architecture 740 is verified in a review 7400, resulting in task support verification 740A in the as-designed phase and integrated systems validation 740B in the as-built phase. Process architecture 750 corresponds to interface analysis 750A in the as-designed phase and plant start-up tests 750B in the as-built phase. System performance criteria 760 are developed into system-specific tests 760A in the as-designed phase and system-specific tests 760B in the as-built phase. System performance criteria 760 together with system-specific tests 760A and 760B correspond to system-level verification. 700A denotes the left-most and center prongs, which correspond to plant-level architecture and design, and 700B corresponds to basic design, detailed design and realisation. Overall, the database system comprising task category information elements, architecture definition information elements, system-level information elements and equipment-level information elements enables verifying the design correctly embodies the design principles associated with the task category information elements. At at least in some embodiments, where a piece of equipment is associated with two task categories having different design principles, the more stringent design principle, safety class or quality requirement may be arranged to prevail concerning the function of the piece of equipment. For example, diversity may be seen as more stringent than redundancy, since in addition to another unit, an additional requirement of different operating principle is assigned to the units. As another example, where differing environmental safety requirements apply, the more stringent requirement may be arranged to prevail. In some embodiments of the invention, a computerized monitoring system is provided, wherein the computerized monitoring system is configured to receive, from the nuclear power station or aircraft, failure notifications. Each failure notification may relate to a failure of an item of equipment, for example one represented by an equipment information element, a system-level information element and/or an architecture information element in a database arranged in accordance with the principles of the present invention. The failure notifications may be automatically generated from sensors arranged to monitor how equipment comprised in the nuclear power station or aircraft perform, for example. The computerized monitoring system may be configured to, responsive to a failure notification, determine, using a database such as one described above, an effect of the failure on how a design principle is complied with. A design principle may comprise at least one of the following: redundancy, diversity, separation, isolation, quality level, reliability level, seismic qualification and environmental condition qualification. For example, where an item of equipment fails, and the failed equipment played a role in providing for a design principle with respect to another item of equipment, a visual or other kind of indication may be provided, the indication conveying that the design principle is not sufficiently provided for as it relates to the another item of equipment. Thus, for example, where a first equipment fails and the first equipment provided, prior to the failure, at least partly, redundancy for a second equipment, the computerized monitoring system may determine the redundancy effect of the failure of the first equipment, using the task category associated with the functional requirement and the at least one design principle to determine the systems and/or equipments the redundancy of which is affected by the failure. An indication may be provided of a reduced redundancy level, and the systems and/or pieces of equipment that the reduced redundancy level affects. The reduced redundancy level is a technical characteristic of the nuclear power station or aircraft and the equipments comprised therein. The first equipment corresponds in the database to a corresponding first equipment information element. Where the first equipment information element is associated, via database relations, to more than one task category information element in the database, the computerized monitoring system may be configured to determine a set comprising each design principle associated with each task category information element associated, via database relations, with the first equipment information element, and to identify, based on each design principle comprised in the set, a technical constraint of an action compensating, at least partly, effects of the failure identifier in the failure notification. Thus, for example, where a pump in a nuclear power station develops a failure, a sensor comprised in the pump may provide a failure notification to the computerized monitoring system. Responsive to the failure notification, the computerized monitoring system may determine that an equipment information element in the database corresponding to the pump is associated, via database relations, with the task category information elements corresponding to the task categories safety function and reactor protection. In this example, task category safety function is associated with design principles redundancy and diversity, and task category reactor protection is associated with design principles redundancy, diversity and separation. Thus, the set of design principles comprises redundancy, diversity and separation. The computerized monitoring system may further be configured to identify, based on each technical design principle comprised in the set, a technical constraint of an action compensating, at least partly, effects of the failure associated with the failure notification. In the example above, compensating actions would be constrained with respect to equipment unit count to meet the design principle redundancy, equipment principle of action to meet the design principle diversity, and equipment location to meet the design principle separation. Thus in accordance with the invention, personnel are enabled to become aware of which aspects of a failed piece of equipment are relevant for safe operation of the nuclear power station, for example. Expressed in other words, the computerized monitoring system is configured to provide information concerning the operational status of the nuclear power station, and deviations from a nominal operational status that result from the failure. A technical effect provided by the computerized monitoring system and associated database lies in enabling reaction to the actually relevant aspects of an equipment that has developed a failure. In prior systems, a decision tree may be employed, for example. However, a decision tree in the case of a nuclear power station is very difficult to maintain due to the highly complex nature of such a station. Furthermore, a decision tree does typically not provide information on the actual aspects of a failed equipment that are of significance, rather, a decision tree simply informs concerning actions needed to replace the failed equipment with an identical one. The constraints described herein, on the other hand, enable reacting to a failure in a way that addresses the technical situation, rather than requires simple duplication of an original design and like-for-like replacement of a failed piece of equipment. A like-for-like replacement may be unavailable in case the type of equipment is no longer in production. Furthermore, a like-for-like replacement may be less preferable than a more modern piece of equipment, which may be more energy-efficient, for example. A modern piece of equipment may furthermore be enabled to assume more roles in terms of redundancy or diversity in the overall system. In a communication network, task categories may include, for example, a switching failure elimination task category and a privacy task category. The switching failure elimination may relate to connecting connections, such as telephone calls, for example, correctly, while privacy may relate, for example, to confidentiality of communicated information and/or subscriber data. Further, resource management may be a task category, whereby radio or trunk resources are allocated in a suitable way to communications routed through the network. It is to be understood that the embodiments of the invention disclosed are not limited to the particular structures, process steps, or materials disclosed herein, but are extended to equivalents thereof as would be recognized by those ordinarily skilled in the relevant arts. It should also be understood that terminology employed herein is used for the purpose of describing particular embodiments only and is not intended to be limiting. Reference throughout this specification to “one embodiment” or “an embodiment” means that a particular feature, structure, or characteristic described in connection with the embodiment is included in at least one embodiment of the present invention. Thus, appearances of the phrases “in one embodiment” or “in an embodiment” in various places throughout this specification are not necessarily all referring to the same embodiment. As used herein, a plurality of items, structural elements, compositional elements, and/or materials may be presented in a common list for convenience. However, these lists should be construed as though each member of the list is individually identified as a separate and unique member. Thus, no individual member of such list should be construed as a de facto equivalent of any other member of the same list solely based on their presentation in a common group without indications to the contrary. In addition, various embodiments and example of the present invention may be referred to herein along with alternatives for the various components thereof. It is understood that such embodiments, examples, and alternatives are not to be construed as de facto equivalents of one another, but are to be considered as separate and autonomous representations of the present invention. Furthermore, described features, structures, or characteristics may be combined in any suitable or technically feasible manner in one or more embodiments. In the following description, numerous specific details are provided, such as examples of lengths, widths, shapes, etc., to provide a thorough understanding of embodiments of the invention. One skilled in the relevant art will recognize, however, that the invention can be practiced without one or more of the specific details, or with other methods, components, materials, etc. In other instances, well-known structures, materials, or operations are not shown or described in detail to avoid obscuring aspects of the invention. While the forgoing examples are illustrative of the principles of the present invention in one or more particular applications, it will be apparent to those of ordinary skill in the art that numerous modifications in form, usage and details of implementation can be made without the exercise of inventive faculty, and without departing from the principles and concepts of the invention. Accordingly, it is not intended that the invention be limited, except as by the claims set forth below. |
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048470066 | summary | The present invention relates to a solid bitumen product having embedded or encapsulated therein granular and/or pulverulent ion-exchange resin which is at least partially saturated with radioactive ions; and to a method for manufacturing the product; and to the use of the product for the long-term storage of radioactive waste of low and intermediate activity. Energy producing power stations generate large quantities of radioactive waste, which must be converted to a form suitable for long-term storage. The major part of this waste, measured in volume, comprises waste of low and intermediate radioactivity. Most of this waste is concentrated in ion-exchangers, while a minor part is concentrated in evaporators. There are obtained in this way large quantities of radioactive ion-exchangers in granular and/or pulverulent form. The evaporation residues can also be converted to granular or pulverulent form. When practising known methods, the resultant ion-exchangers are dried and then mixed with liquid bitumen, normally at a minimum temperature of 130.degree. C. The resultant mixture is normally transferred into barrels, e.g. having a volumetric capacity of 200 l, in which the mixture is allowed to solidify and cool to ambient temperature, whereafter the barrels are sealed. The barrels are then placed in long-term storage locations of particular construction, e.g. rock cavities. The known method of embedding dry ion-exchange resin in bitumen at high temperatures is encumbered with several drawbacks, of which the most serious reside in the risk of fire when using bitumen at high temperatures, and in the fact that the ion-exchangers are in a dry state. Dry ion-exchangers swell considerably when coming into contact with water. Consequently, should the dry ion-exchanger embedded in the bitumen come into contact with water, which is at least theoretically possible, there would be generated an extremely high swelling pressure sufficient to explode the encasing barrel and therewith spread the radioactivity throughout the surroundings. This risk, together with that of fire, has been a subject of criticism on the part of the authorities. Since the liquid bitumen has a high temperature, normally higher than 150.degree. C., water present in the ion-exchange resin will depart upon contact of the resin with the bitumen, the resin therewith losing the major part of its water content. Water is also given off when pre-heating the ion-exchange resin prior to said mixing process. Consequently, when practicing known techniques, it is impossible for the ion-exchange resin embedded in solidified bitumen to be moist. Other known methods and processes for treating radioactive material are found described in GB-A-959 751, CH-A-549 265, and FR-A-2 289 034, the radioactive material in these cases being mixed with a bitumen and water emulsion. The mixture is then heated to remove residual water, and hence the radioactive material is present in the bitumen in a dry state. GB-A-2 116 355 describes a method in which ion-exchange resin having radioactive ions absorbed therein is encapsulated in bitumen. The radioactive ion-exchange resin and the bitumen are heated to extract water therefrom. The objective of the present invention is to avoid the aforesaid drawbacks and to provide a novel and improved method of encapsulating ion-exchange resin in a solid bitumen matrix, in which subsequent to being encapsulated the ion-exchanger is in a wet, swollen form and with which there is no risk of fire during the actual working operation, and to enable the resultant product to be used for the long-term storage of radioactive waste of low and intermediate levels of activity. This objective is achieved in accordance with the invention by mixing/combining the ion-exchange resin with a bitumen-water emulsion; by adding the ion-exchange resin, and optionally the waste material, to a given quantity of emulsion in an amount such that the break point of the emulsion is reached and the mixture transforms to a solid product, in which the ion-exchange resin is present in a swollen, aqueous form. The method according to the invention affords many advantages in relation to known techniques, in that it is possible to use a moist ion-exchanger, thereby rendering it unnecessary to dry the ion-exchanger prior to its embedment. It is also possible, however, to use dry ion-exchanger that will swell when coming into contact with the water present in the aqueous emulsion. The aqueous emulsion can be used at ambient temperatures, therewith obviating the need to heat the bitumen. A further advantage is that the ion-exchanger and bitumen can be mixed with the use of existing apparatus and equipment, although in this case without supplying thermal energy or minor forms of energy to the system, so as to ensure that no major evaporation of the water content of the emulsion takes place. The use of mixing appliances is not necessary, however, since the ion-exchanger can be added to the emulsion without mixing the two together. Preferred embodiments of the invention are set forth in the depending claims. The invention will now be described with reference to a number of preferred embodiments. |
abstract | An optical axis adjusting mechanism for an X-ray lens, an X-ray analytical instrument and a method of adjusting an optical axis of an X-ray lens, capable of enhancing detection efficiency of an X-ray while preventing degradation of the device performance are provided. An optical axis adjusting mechanism for an X-ray lens to be implemented in an X-ray analytical instrument, includes an exit side adjusting mechanism for adjusting an exit side focal point of the X-ray lens to focus on an X-ray detector, and an entrance side adjusting mechanism for adjusting an entrance side focal point of the X-ray lens to focus on an analytical point of a sample, and the entrance side adjusting mechanism is disposed with a greater distance from the X-ray lens than a distance between the exit side adjusting mechanism and the X-ray lens. |
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abstract | A charged particle beam irradiation apparatus includes: a transport line configured to transport a charged particle beam; and a rotating gantry rotatable around a rotation axis, wherein the transport line has an inclined section configured to make the charged particle beam advancing in a direction of the rotation axis advance to be inclined so as to become more distant from the rotation axis, and is formed so as to turn the charged particle beam advanced in the inclined section to a rotational direction of the rotation axis and bend the charged particle beam turned to the rotational direction to the rotation axis side, the rotating gantry is formed of a tubular body which can accommodate an irradiated body and supports the transport line, and the inclined section is disposed to pass through the inside of the tubular body of the rotating gantry. |
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description | The present application claims the benefit of U.S. Provisional Patent Application Ser. No. 62/357,603, filed Jul. 1, 2016, the entirety of which is hereby incorporated by reference. The present invention relates generally to a container for storing and/or transporting spent nuclear fuel, and more specifically to such a container having a collapsible trunnion or lifting lug. Heavy casks containing hazardous materials such as high level nuclear waste and fissile materials are typically handled by a set of trunnions. The trunnions are generally made of a cylindrical bar stock welded to a hard location near the top of the cask. The trunnion must project out sufficiently to provide an engagement shoulder for a lift yoke to engage it. This projection, however, is a problem where the cask must be designed to withstand a free fall event such as that required for transport casks containing used nuclear fuel. The federal regulations and the IAEA standards require the cask to be qualified under a free fall event from a height of 30 feet onto an essentially unyielding surface under any orientation of impact. In such a case, the cask may be equipped with an impact limiter at each extremity to absorb the kinetic energy of impact by crushing. The projection of the trunnion, made of a high strength steel or other alloy material, however, interferes with the crushing action of the impact limiter if the impact orientation of the cask is aligned with the plane of the trunnion. The solution to this design problem thus far has been to tap the trunnions and thread them into the cask's flange. The trunnion is removed when not in use to eliminate the threat of trunnion penetration during the above-mentioned design basis accident event. This approach has three major shortcomings: (1) The threaded joint sometimes freezes under the bending moment from the lifted load making the trunnion's subsequent removal problematic; (2) It may not be possible to handle the cask without the trunnions in place (after all, their sole purpose is to enable cask's handling); and (3) The trunnions are restricted to be located in the neck of the cask so that its projection beyond the cask's body is minimized. The above limitations make the conventional trunnion design a rather unsatisfactory embodiment. Thus, a need exists for a trunnion design that overcomes the aforementioned deficiencies. The present invention, in one aspect, is a container for storing and/or transporting spent nuclear fuel. The container includes a body that defines an internal cavity that holds the spent nuclear fuel and an outer surface. The outer surface has holes formed therein into which trunnions are positioned. The container can be lifted by a lift yoke by coupling the lift yoke to the trunnions. The trunnions may include first and second components such that the first component is slidable in its axial direction relative to the second component when a force that exceeds a threshold acts on the second component. Thus, the second component may be slidable between a protruded state in which a portion of the second component protrudes from the outer surface of the body and a retracted state in which the second component does not protrude from the outer surface of the body. In one aspect, the invention can be a container for storing and/or transporting spent nuclear fuel comprising: a body having an outer surface and an inner surface defining an internal cavity; a plurality of blind holes formed into the body, each of the blind holes defined by a floor and a sidewall extending from the floor to an opening in the outer surface of the body; a plurality of trunnions coupled to the body, each of the trunnions comprising: a first component located within one of the blind holes, the first component extending from a first end to a second end and having an inner surface defining a hollow interior; and a second component at least partially located within the hollow interior of the first component, the second component extending from a first end to a second end along a longitudinal axis; and wherein for each of the plurality of trunnions, the second component is axially slidable relative to the first component between: (1) a protruded state in which a portion of the second component protrudes from the outer surface of the body; and (2) a retracted state in which the second component does not protrude from the outer surface of the body In another aspect, the invention can be a container for storing and/or transporting spent nuclear fuel comprising: a body having an outer surface and an inner surface defining an internal cavity configured to hold spent nuclear fuel; a hole formed into the the body, the hole defined by a floor and a sidewall extending from the floor to an opening in the outer surface of the body; a trunnion coupled to the body within the hole, the trunnion comprising: a first component welded to the body within the hole, the first component having a first end that is in contact with the floor of the hole, a second end that is flush with or recessed relative to the outer surface of the body, an outer surface that is in contact with the sidewall of the hole, and an inner surface that defines a hollow interior; and a second component located within the hollow interior of the first component, the second component extending from a first end to a second end along a longitudinal axis, the second component comprising a first portion that is located within the hollow interior of the first component and spaced from the floor of the hole by a gap and a second portion that protrudes from the outer surface of the body; and wherein upon application of an axial force that exceeds a predetermined threshold onto the second end of the second component, the second component slides relative to the first component in an axial direction into the gap In yet another aspect, the invention can be a container for storing and/or transporting spent nuclear fuel comprising: a body having an outer surface and an inner surface defining an cavity; a lid enclosing a top end of the cavity, the lid having a bottom surface facing the cavity and an opposite top surface; at least one lifting lug coupled to the lid, the lifting lug comprising: a first component coupled to the lid and protruding from the top surface of the lid, the first component having a top surface and an inner surface that defines a hollow interior; and a second component coupled to the first component and extending from a first end to a second end along a longitudinal axis, the second component having a first portion located within the hollow interior of the first component and a second portion protruding from the top surface of the first component; and wherein upon application of a compression force that exceeds a predetermined threshold onto the second component, the second component axially slides relative to the first component until a top surface of the second component is flush with or recessed relative to the top surface of the first component. In a further aspect, the invention can be a container for storing and/or transporting spent nuclear fuel comprising: a body having an outer surface and an inner surface defining an internal cavity; a lid enclosing a top end of the cavity, the lid having a bottom surface facing the cavity and an opposite top surface; at least one lifting lug coupled to the lid, the lifting lug comprising: a first component coupled to the lid and protruding from the top surface of the lid, the first component having an inner surface that defines a hollow interior and a top surface; and a second component coupled to the first component and at least partially located within the hollow interior of the first component, the second component extending from a first end to a second end along a longitudinal axis; and wherein the second component of the lifting lug is axially slidable relative to the first component of the lifting lug between: (1) a protruded state in which a portion of the second component protrudes from the top surface of the first component; and (2) a retracted state in which the second portion of the lifting lug does not protrude from the top surface of the first component of the lifting lug. In a still further aspect, the invention can be a container for storing and/or transporting spent nuclear fuel comprising: a body having an outer surface and an inner surface defining an internal cavity; a plurality of blind holes formed into the body, each of the blind holes defined by a floor and a sidewall extending from the floor to an opening in the outer surface of the body; a plurality of trunnions, each of the trunnions located within one of the blind holes and extending from a first end to a second end along a longitudinal axis; and wherein at least one of the plurality of trunnions is axially slidable relative to the body between: (1) a protruded state in which a portion of the trunnion protrudes from the outer surface of the body; and (2) a retracted state in which the trunnion does not protrude from the outer surface of the body. In a yet further aspect, the invention may be a lifting lug comprising: a first component having an inner surface that defines a hollow interior and a top surface; a second component at least partially located within the hollow interior of the first component, the second component extending from a first end to a second end along a longitudinal axis; and wherein the second component of the lifting lug is axially slidable relative to the first component of the lifting lug between: (1) a protruded state in which a portion of the second component protrudes from the top surface of the first component; and (2) a retracted state in which the second portion of the lifting lug does not protrude from the top surface of the first component of the lifting lug. In another aspect, the invention may be a container for storing and/or transporting spent nuclear fuel comprising: a body having an outer surface and an inner surface defining an internal cavity; a canister containing radioactive materials located in the internal cavity; at least one lifting device coupled to the body; wherein the lifting device is collapsible between: (1) a non-collapsed state; and (2) a collapsed state; and wherein in the non-collapsed state the lifting device protrudes a greater distance from the outer surface of the body than in the non-collapsed state. Further areas of applicability of the present invention will become apparent from the detailed description provided hereinafter. It should be understood that the detailed description and specific examples, while indicating the preferred embodiment of the invention, are intended for purposes of illustration only and are not intended to limit the scope of the invention. The following description of the preferred embodiment(s) is merely exemplary in nature and is in no way intended to limit the invention, its application, or uses. The description of illustrative embodiments according to principles of the present invention is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. In the description of embodiments of the invention disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,” “above,” “below,” “up,” “down,” “top” and “bottom” as well as derivatives thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation unless explicitly indicated as such. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. Moreover, the features and benefits of the invention are illustrated by reference to the exemplified embodiments. Accordingly, the invention expressly should not be limited to such exemplary embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features; the scope of the invention being defined by the claims appended hereto. Referring to FIGS. 1-4 concurrently, a container 100 for storing and/or transporting spent nuclear fuel is illustrated according to an embodiment of the present invention. The container 100 may be configured to hold any type of radioactive materials, including radioactive waste such as spent nuclear fuel, nuclear waste, or the like, and also including other types of materials. The container 100 may be a cask that is intended for the dry storage of spent nuclear fuel after the spent nuclear fuel has been cooled in a spent fuel pool to reduce the heat and radiation to a sufficiently low level so that the spent nuclear fuel can be transported with safety. In that regard, in some embodiments the container 100 may be a steel structure that is welded and/or bolted closed that provides a leak-tight confinement of the spent nuclear fuel. The spent nuclear fuel within the container 100 may be surrounded by an inert gas. During long-term storage, the container 100 may be surrounded by additional steel, concrete, or other material to provide radiation shielding to workers and members of the public. The containers 100 described herein are vertical dry storage casks that are intended to be placed vertically in a concrete vault during storage. However, the containers 100 may be of the horizontal type as well. While the container 100 is discussed herein as being used to store spent nuclear fuel, it is to be understood that the invention is not so limited and that, in certain circumstances, the container 100 can be used to transport spent nuclear fuel from location to location if desired. Moreover, the container can be used in combination with any other type of high level radioactive waste such as high level nuclear waste, fissile materials, or the like. The container 100 may be designed to accept one or more canisters for storage at an Independent Spent Fuel Storage Installation (“ISFSI”). All canister types engineered for the dry storage of spent nuclear fuel can be stored in the container 100. Suitable canisters include multi-purpose canisters (“MPCs”) and, in certain instances, can include thermally conductive casks that are hermetically sealed for the dry storage of high level radioactive waste. Typically, such canisters may include a honeycomb basket or other structure to accommodate a plurality of spent nuclear fuel rods in spaced relation. In the exemplified embodiment, the container 100 comprises a body 110 having an outer surface 111 and an inner surface 112 that defines an internal cavity 113 within which spent nuclear fuel may be contained for storage and/or transport. The internal cavity 113 extends from a bottom end to a top end along a longitudinal axis A-A. In the exemplified embodiment, the body 110 has a rectangular shape formed by a first sidewall 101, a second sidewall 102 that is opposite the first sidewall 101, a third sidewall 103, and a fourth sidewall 104 that is opposite the third sidewall 103. Each of the third and fourth sidewalls 103, 104 extends between the first and second sidewalls 101, 102. The first, second, third, and fourth sidewalls 101-104 collectively form a sidewall of the body 110. Of course, shapes other than rectangular are possible in other embodiments including cylindrical (FIG. 8) and other polygonal/prismatic shapes. Furthermore, the body 110 comprises a base plate 105 that connects to a bottom end of each of the first, second, third, and fourth sidewalls 101-104 and closes a bottom end of the internal cavity 113. The body 110 also comprises a lid flange 106 extending from a top end of each of the first, second, third, and fourth sidewalls 101-104. An outer diameter of the lid flange 106 may be slightly less than an outer diameter of the sidewall as best seen in FIG. 3. A lid 107 is coupled to the lid flange 106 to close a top end of the internal cavity 113. Furthermore, in the exemplified embodiment a cover (also known as a secondary lid) 108 is coupled to a top end of the lid flange 106 above the lid 107. In the exemplified embodiment, each of the sidewalls 101-104 and the base plate 105 comprises two layers. The inner layer is a wall formed of stainless steel that forms the containment structure of the container 100. The outer layer is a dose blocker plate that may also be formed of stainless steel and may form a secondary containment structure of the container 100. Each of the inner and outer layers may form a distinct hermetically sealed vessel thus providing dual-walled protection against radiation to prevent such radiation from exiting the internal cavity 113 and entering the atmosphere. Although described herein as being formed of stainless steel, the inner and outer layers can be formed of other materials, such as austenitic stainless steel and other metal alloys including Hastelloy™ and Inconel™. The dose blocker plate may include additives or the like for radiation shielding purposes to prevent radiation emanating from the spent nuclear fuel from exiting the internal cavity 113. Of course, in other embodiments the sidewalls 101-104 may have just a single layer. Furthermore, in still other embodiments the sidewalls 101-104 may be formed of concrete or the like instead of stainless steel. The container 100 also comprises a plurality of trunnions 120 coupled to the body 110. Specifically, in the exemplified embodiment there are four trunnions 120 on the first sidewall 101 and four trunnions 120 on the second sidewall 102. Of course, there may be more or less than four of the trunnions 120 on each of the first and second sidewalls 101, 102 in alternative embodiments. Furthermore, although in the exemplified embodiment there are no trunnions 120 on the third and fourth sidewalls 103, 104, in other embodiments there may be one or more trunnions 120 on the third and fourth sidewalls 103, 104 in addition to or instead of the trunnions 120 on the first and second sidewalls 101, 102. In some embodiments, all of the trunnions 120 are located on one or more of the sidewalls 101-104 and there are no trunnions 120 located on the lid flange 106. In the exemplified embodiment, each of the trunnions 120 on the first sidewall 101 is aligned with one of the trunnions 120 on the second sidewall 102. More specifically, the trunnions 120 on the first and second sidewalls 101, 102 are aligned along a reference axis B-B that is perpendicular (possibly without intersecting) to the longitudinal axis A-A of the internal cavity 113. Thus, each of the trunnions 120 on the first sidewall 101 is spaced the same distance from the lid 107 and the same distance from the third sidewall 103 as one of the trunnions 120 on the second sidewall 102. The purpose of the trunnions 120 is to enable a lift yoke to engage the portion of the trunnion 120 that protrudes out from the outer surface 111 of the body 110 to transport the container 100. Thus, the lift yoke can raise the container 100 off the ground via engagement between the lift yoke and the trunnions 120 to move the container 100 from one location to another. The lift yoke serves as the interface between a plant crane and the container 100 to maneuver the container 100 as desired. An exemplary lift yoke is illustrated in FIG. 10 and will be described briefly below. The container 100 is a hermetically sealed apparatus that has no openings or penetrations therein when the lid 107 is coupled to the top ends of the sidewalls 101-104 to close the internal cavity 113. Thus, there is no passageway extending from the internal cavity 113 to the external atmosphere, which is necessary to prevent radiation from entering the atmosphere. In that regard, it is important to ensure that under no circumstances can the trunnions pierce the body 110 and form a passageway into the internal cavity 113. Thus, even if the container 100 were to fall on one of its first or second sidewalls 101, 102, it is important that the contact between the trunnions 120 and the ground or hard surface upon which the container 100 falls does not cause the trunnions 120 to pierce the body 110 or otherwise penetrate the internal cavity 113 of the body. As mentioned in the background, previously this was accomplished by coupling the trunnions to the body via a threaded engagement so that the trunnions could be removed when not being used to maneuver the container 100. However, this conventional technique has disadvantages that are overcome in the present invention whereby the trunnions 120 are formed as a collapsible structure so that during a fall event as described above, the trunnions 120 will collapse into the body 110 of the container 100 rather than piercing the body 110 of the container. Thus, the trunnions 120 need not be removed from the body 110 during periods of non-use because there is no possibility that the trunnions 120 can pierce the body 110 of the container 100 even during an undesirable or unplanned fall or tip-over event. In the exemplified embodiment, the trunnions 120 are in a protruded state such that a portion of the trunnions 120 extends/protrudes from the outer surface 111 of the body 110. The trunnions 120 must be in the protruded state in order for the lift yoke to be able to engage the trunnions 120 to move the container 100. Specifically, as described below with reference to FIG. 10, the hooks of the lift yoke must be able to be positioned between a flange located at a distal end of the trunnions 120 and the outer surface 111 of the body 110 in order for the lift yoke to lift the container 100 by the trunnions 120. Thus, in normal operation and use of the container 100, the trunnions 120 are in the protruded state. The trunnions 120 are configured to be alterable from the protruded state to a retracted state whereby the trunnions 120 no longer protrude from the outer surface 111 of the body 110 (see FIGS. 7A and 7B, described in more detail below). The trunnions 120 only move into the retracted state when they are forced into that state, such as by the container 100 tipping over and landing on one of the trunnions 120. Specifically, if the container 100 were to tip over, contact of the trunnions 120 with the ground or some other hard surface combined with the force created by the weight of the container 100 pressing the trunnions 120 into the ground or other hard surface will move the trunnion 120 from the protruded state to the retracted state without causing any damage to the body 110 of the container 100. This is because, as discussed below, the trunnion 120 or a portion thereof will simply slide within a hole in the body 110 from which it extends, as described in more detail below. When in the retracted state (illustrated in FIG. 7B), the lift yoke is no longer capable of engaging the trunnions 120 because there is no surface area of the trunnions 120 protruding from the outer surface 111 of the body 110 to be engaged by the lift yoke. Thus, when the trunnions 120 are in the retracted state, the container 100 cannot be moved via engagement between the lift yoke and the trunnions 120 due to a lack of an engagement surface on the trunnions 120. As stated above, in certain embodiments the trunnions 120 may collapse from the protruded state into the retracted state upon an axial force being applied onto the portions of the trunnions 120 that are protruding from the outer surface 111 of the body 110. Thus, if there were a tip-over event and the container 100 were to fall onto one of its sidewalls 101-104, the weight of the container 100 against the ground or other hard surface would cause the trunnions 120 to be moved from the protruded state to the retracted state without the trunnions 120 penetrating the inner surface 112 of the body 110. Thus, by forming the trunnions 120 to be collapsible as described herein, the sidewalls 101-104 of the body 110 that have the trunnions 120 thereon are protected from being damaged by the trunnions 120 without ever requiring removal of the trunnions. Referring to FIGS. 5-7A, the container 100 comprises a plurality of blind holes 115 formed into the outer surface 111 of the body 110. In the exemplified embodiment, the number of blind holes 115 corresponds with the number of trunnions 120 because one of the trunnions 120 is positioned within each of the blind holes 115 and secured to the body 110 within the blind hole 115. Thus, the blind holes 115 are formed into the first and second sidewalls 101, 102 of the body 110 in the exemplified embodiment, but they may be located at any position along the body 110 at which a trunnion 120 is desired to facilitate lifting and maneuvering of the container 100. As will be understood from the description below in conjunction with FIGS. 7A and 7B, the trunnions 120 remain coupled to the body 110, or at least located within the blind hole 115 of the body 110, in both the protruded and retracted states. Each of the blind holes 115 is defined by a floor 116 and a sidewall 117 extending from the floor 116 to an opening 118 in the outer surface 111 of the body 110. The blind holes 115 do not penetrate the entire thickness of the sidewall of the body 110 in which they are formed but rather extend a distance into the sidewall to the floor 116 of the blind hole 115. A portion of the sidewall of the body 110 remains between the floor 116 of the blind hole 115 and the interior cavity 113 of the body 110. In the exemplified embodiment, the blind holes 115 have a circular cross-sectional shape. However, the shape of the blind holes 115 should correspond with the shape of the trunnions 120 and thus although the blind holes 115 and the trunnions 120 have a circular cross-sectional shape in the exemplified embodiment, this is not required in all embodiments and other prismatic cross-sectional shapes (i.e., triangular, rectangular, or the like) may be used without affecting the function described herein. Still referring to FIGS. 5-7A concurrently, the trunnions 120 will be described in greater detail in accordance with an exemplary embodiment. FIGS. 5 and 6 illustrate different cross-sectional views of the container 100 that illustrate the structural details of the each of the trunnions 120 and FIG. 7A illustrates a close-up view of one of the trunnions 120 in cross-section. Due to the enlarged size of the trunnion 120 in FIG. 7A, some of the reference numerals used below to describe the structure and function of the trunnion 120 may only be provided in FIG. 7A for clarity and to avoid clutter. However, the features referred to by these reference numerals are also illustrated in FIGS. 5 and 6 and thus those figures can be viewed in conjunction with FIG. 7A to gain a full understanding and appreciation of the teachings set forth herein. In the exemplified embodiment, each of the trunnions 120 has the same structure and thus the description below will be made with regard to one of the trunnions 120, it being understood that the description is applicable to each trunnion 120. Of course, in other embodiments some of the trunnions 120 may have a structure different than that which is shown in the drawings and described herein. For example, some of the trunnions 120 may be collapsible as described herein and others may be of a more conventional type that are detachable from the body 110 via a threaded connection or the like. Additionally or alternatively, some of the trunnions may be of the type that are welded directly to the body 110 and not detachable therefrom or collapsible. Of course, it may be preferable in some embodiments for each of the trunnions 120 to be collapsible. In the exemplified embodiment, the trunnions 120 have a two-part structure comprising a first component 130 and a second component 150. As will be discussed in greater detail below, in some embodiments the first component 130 may be omitted and the trunnions 120 may have a one-part structure without changing the function described herein. In the exemplified embodiment, both the first and second components 130, 150 are formed of a metal such as stainless steel, although other metals and metal alloys may be used in other embodiments. The first component 130 extends from a first end 131 to a second end 132 and has an outer surface 133 and an inner surface 134. The inner surface 134 is smooth and free of bumps, ridges, protuberances or the like and defines a hollow interior 135. In the exemplified embodiment, the hollow interior 135 is open at both the first and second ends 131, 132 and thus the hollow interior 135 forms a passageway through the first component 130 from the first end 131 to the second end 132. However, the invention is not to be so limited in all embodiments and the first end 131 of the first component 130 may be closed while an opening remains in the second end 132 of the first component 130. The first component 130 is positioned within one of the blind holes 115 so that the first end 131 of the first component 130 is in contact with the floor 116 of the blind hole 115 and the outer surface 133 is in contact with the sidewall 117 of the blind hole 115. The outer diameter of the first component 130 and the diameter of the blind hole 115 may be selected to ensure that the first component 130 fits snugly within the blind hole 115 so that it can not be readily removed from the blind hole 115 once positioned therein. In some embodiments, the first component 130 may be welded to the body 110 to strengthen the attachment between the first component 130 and the body 110. For example, an annular weld joint may be formed at the location where the inner surface 134 of the first component 130 meets the floor 116 of the blind hole 115. Alternatively, an annular weld joint may be formed at the location where the second end 132 of the first component 130 meets the outer surface 111 of the body 110. The connection between the first component 130 and the body 110 may be further reinforced by radial gussets extending between the inner surface 134 of the first component 130 and the floor 116 of the blind hole 115. In certain embodiments, the first component 130 remains fixed to the body 110 regardless of whether the trunnion 120 is in the protruded or retracted states. The second end 132 of the first component 130 comprises a first portion 136 and a second portion 137. The first portion 137 of the second end 132 of the first component 130 is flush with the outer surface 111 of the body 110 when the first component 130 is inserted into the blind hole 115 as illustrated. The second portion 137 of the second end 132 of the first component 130 is recessed relative to the first portion 136 and recessed relative to the outer surface 111 of the body 110 when the first component 130 is inserted into the blind hole 115 as illustrated. Thus, the second end 132 of the first component 130 has a stepped surface. The second portion 137 of the second end 132 of the first component 130 forms a nesting groove 138 in the second end 132 of the first component 130 within which a portion of the second component 150 may nest when in the retracted state as described in more detail below with reference to FIG. 7B. In the exemplified embodiment, the second portion 137 of the second end 132 of the first component 130 is annular or ring-like in shape and the first portion 136 of the second end 132 of the first component 130 surrounds the second portion 137 of the second end 132 of the first component 130. Of course, the invention is not to be limited by the stepped surface of the second end 132 of the first component 130 as illustrated and the second end 132 of the first component 130 may be flat or planar rather than stepped in other embodiments. Although the first portion 136 of the second end 132 of the first component 130 is illustrated as being flush with the outer surface 111 of the body 110 in the exemplified embodiment, it may be recessed relative to the outer surface 111 of the body 110 in other embodiments. While the first component 130 of the trunnion 120 is a hollow structure, the second component 150 of the trunnion 120 is a solid structure extending from a first end 151 to a second end 152 along a longitudinal axis C-C. Although illustrated as being an entirely solid body, the second component 150 may have an internal cavity in other embodiments to reduce material costs so long as it does not detract from the ability 150 of the second component 150 to support the weight of the container 100. Specifically, during maneuvering of the container 100, the hooks of the lift yoke are coupled to the second component 150 of the trunnion 120. Thus, the second component 150 of the trunnion 120 must be sufficiently rigid and strong to support the entire weight of the container 100 without bending or breaking. Forming the second component 150 from solid steel is therefore preferable. The second component 150 is positioned at least partially within the hollow interior 135 of the first component 130 and protrudes from the outer surface 111 of the body 110. In that regard, the second component 150 has an outer surface 153 that may be in direct contact with the inner surface 134 of the first component 130. The second component 150 may be coupled to the first component 130 via a seal weld (i.e., a fillet weld) in some embodiments. In such an embodiment, the weld will create an axial load retention that will prevent the second component 150 from sliding in its axial direction (i.e., in the direction of the longitudinal axis C-C) until a force is applied onto the second component 150 that causes the weld to break. Thus, the weld provides the trunnion 120, and more specifically the second component 150 of the trunnion 120, with a limited axial load bearing capacity such that an axial load up to a predetermined threshold will not break the weld and cause the second component 150 to slide axially. However, a force that breaks the weld will enable the second component 150 to slide axially relative to the first component 130. The container 100 tipping over and falling on the trunnion 120 would create a sufficient force to break the weld. In other embodiments, the second component 150 may be shrunk-fit in the hollow interior 135 of the first component 130. Specifically, the second component 150 may be shrunk by cooling the second component 150 and then inserting the cooled/shrunk second component 150 into the hollow interior 135 of the first component 130. Then, when the second component 150 returns to its normal temperature, the second component 150 will expand back to its original size. This expansion will cause the second component 150 to be tight-fit within the hollow interior 135 of the first component 130. As a result, the outer surface 153 of the second component 150 will be in intimate surface contact with the inner surface 134 of the first component, thereby creating an interface pressure between the first and second components 130, 150. This interface pressure will provide the trunnion 120, and more specifically the second component 150 of the trunnion, with a limited axial load bearing capacity such that an axial load up to a predetermined threshold will not cause the second component 150 to slide axially due to the interface pressure between the first and second components 130, 150. However, a force greater than the predetermined threshold will enable the second component 150 to slide axially relative to the first component 130. Again, the container 100 tipping over and falling on the trunnion 120 would create a sufficient force to overcome the interface pressure and cause the second component 150 of the trunnion 120 to slide axially into the hollow interior 135 of the first component 130. In some embodiments, the second component 150 may be welded to the first component 130 and shrunk-fit in the hollow interior 135 of the first component 130. The second component 150 is positioned within the hollow interior 135 of the first component 130 so that the first end 151 of the second component 150 is spaced apart from the floor 116 of the blind hole 115 by a gap 155. Furthermore, the second component 150 has a first portion 156 that is located within the hollow interior 135 of the first component 130 and a second portion 157 that protrudes from the outer surface 111 of the body 110. The lift yoke is able to hook onto the protruding portion of the second component 150 during lifting/maneuvering of the container 100. Thus, in order to move the container 100 via engagement between the lift yoke and the trunnions 120, the second component 150 must protrude from the outer surface 111 of the body 110 to provide an engagement shoulder for the lift yoke. The gap 155 has a length measured from the floor 116 of the blind hole 115 to the first end 151 of the second component 150 and the second portion 157 of the second component 130 has a length measured along the longitudinal axis C-C of the second component 150 such that the length of the gap 155 is equal to or greater than the length of the second portion 157 of the second component 150. The gap 155 having an equal or greater length than the second portion 157 of the second component 150 (i.e., the portion of the second component 150 that protrudes from the outer surface 111 of the body 110 when the trunnion 120 is in the protruded state) enables the second component 150 to slide into the gap 155 during a collapsing procedure a sufficient amount so that the second portion 157 of the second component 150 no longer protrudes from the outer surface 111 of the body 110. The second component 150 comprises a body portion 160 and a flange portion 161 extending radially from the body portion 160 at the second end 152 of the second component 150. When the lift yoke is being used, the hooks of the lift yoke become trapped between the flange portion 161 of the second component 150 and the outer surface 111 of the body 110 of the container (or the second end 132 of the first component 130). The flange portion 161 prevents the hook of the lift yoke from readily sliding off the trunnion 120 during a moving operation. Referring to FIGS. 7A and 7B concurrently, the collapsible nature of the trunnion 120 will be described. Due in part to the fact that the second component 150 is located within the hollow cavity 135 of the first component 130 in a spaced apart manner from the floor 116 of the blind hole 115, when a sufficient axial force is applied onto the second component 150 towards the internal cavity 113 of the body 110, the second component 150 can slide axially relative to the first component 130 into the gap 155. The only things hindering free axial movement of the first component 130 into the gap 155 are the possible weld joint coupling the first and second components 130, 150 together and the possible interface pressure between the inner surface 134 of the first component 130 and the outer surface 153 of the second component 150. However, if a force acts on the second component 150 that overcomes the weld joint and/or the interface pressure, the second component 150 will slide axially into the gap 155. Thus, the second component 150 of the trunnion 120 is axially slidable relative to the first component 130 of the trunnion 120 between a protruded state (FIG. 7A) in which the second portion 157 of the second component 150 protrudes from the outer surface 111 of the body 110 and a retracted state (FIG. 7B) in which the second component 150 does not protrude from the outer surface 111 of the body 110. In the exemplified embodiment, in the protruded state the first portion 156 of the second component 150 is located within the hollow interior 135 of the first component 130 and the second portion 157 of the second component 150 protrudes from the hollow interior 135 of the first component 130. Stated another way, in the protruded state the first portion 156 of the second component 150 is located within the blind hole 115 and the second portion 157 of the second component 150 protrudes from the outer surface 111 of the body 110. Furthermore, in the exemplified embodiment, in the retracted state the first and second portions 156, 157 of the second component 150 are both located within the hollow interior 135 of the first component 130 such that no part of the second component 150 protrudes out of the hollow interior 135 and from the outer surface 111 of the body 110. Stated another way, in the retracted state the first and second portions 156, 157 of the second component 150 are both located within the blind hole 115. Thus, in both the protruded and retracted states, at least a portion of the second component 150 of the trunnion 120 is located within the blind hole 115. As has been described above, sliding the second component 150 from the protruded state to the retracted state occurs when a sufficient axial force F (i.e., one that is greater than the axial load retention capacity of the second component 150 relative to the first component 130) acts upon the second component 150. This axial force F causes the second component 150 to slide axially relative to the first component 130 (and relative to the body 110) into the gap 155. When the second component 150 is in the retracted state, the flange portion 161 of the second component 150 nests within the nesting groove 138 formed by the first and second portions 136, 137 of the second end 132 of the first component 130. Both the nesting groove 138 and the flange portion 161 may be annular shaped in some embodiments. As a result, when the second component 150 is in the retracted state, the second end 152 of the second component 150 is flush with the outer surface 111 of the body 110. Of course, in other embodiments in the retracted state the second end 152 of the second component 150 may be recessed relative to the outer surface 111 of the body 110. This can be dictated by the length of the blind hole 115 and the length of the second component 150 of the trunnion 120. During normal use of the container 100, the second component 150 of the trunnion 120 is in the protruded state such that it can be used at any time for movement of the container 100. In order to ensure that the second component 150 is sufficiently strong to enable it to support the weight of the container 100, it is desirable for an adequate amount of the second component 150 to be located within the hollow interior 135 of the first component 130. Stated another way, the second component 150 is configured to project sufficiently inside the first component 130 such that it develops the full stiffness of a cantilevered beam with the first component 130 serving as the anchor of the cantilever. In that regard, the first portion 156 of the second component 150 (which is the portion located within the hollow interior 135 of the second component 130) has a length L and a diameter D. A ratio of the diameter D to the length L is between 1:1 and 2:1. Thus, a length of the second component 150 equal to between ½ and 1.0 of the diameter D of the second component 150 is located within the hollow interior 135 of the first component 130 when the second component 150 is in the protruded state. In the exemplified embodiment, when the second component 150 is in the protruded state at least two-thirds of the length of the second component 150 measured from the first end 151 to the second end 152 is located within the hollow interior 135 of the first component 130 (and also within the blind hole 115). Thus, a ratio of the length of the first portion 156 of the second component 150 to the length of the second portion 157 of the second component is between 2:1 and 3:1 in some embodiments, and more specifically approximately 2.5:1. The collapsible trunnion 120 configured in this manner will have a limited axial load bearing capacity without any reduction in its load bearing capacity which derives from its bending rigidity, which is not impaired by the reduction in its axial load sustaining capacity. The collapsible trunnion 120 is a structurally competent member in bending but a relatively weak member in axial tension or compression. In order to facilitate the axial sliding of the second component 150 relative to the first component 130, the outer surface 153 of the second component 150 and the inner surface 134 of the first component 130 are preferably smooth and without ridges, protuberances, or the like. Furthermore, although in the exemplified embodiment the trunnion 120 is illustrated and described as including both of the first and second components 130, 150, in some alternative embodiments the first component 130 may be omitted. In such an embodiment, the portions of the trunnion 120 illustrated as forming the first component 130 will instead be formed directly by the body 110 of the container 100 and the second component 150 will form the entirety of the trunnion 120. Thus, as used herein “trunnion” may refer to the combination of the first and second components 130, 150 or just the second component 150. In embodiments that omit the first component 130, the second component 150 may be positioned in intimate surface contact with the sidewall 117 of the blind hole 115 so as to be axially slidable relative to the body 110 in the manner described herein. Thus, the first component 130 serves as an intermediary between the body 110 and the second component 150 in the exemplified embodiment, but it may not be required in all embodiments. In some embodiments the trunnion 120 may collapse rather than slide axially. Thus, the trunnion 120 may be configured to collapse upon itself. Furthermore, in some embodiments the trunnion 120 may be collapsible between a non-collapsed state and a collapsed state such that in the non-collapsed state the trunnion 120 protrudes a greater distance from the outer surface 111 of the body 110 than in the collapsed state. Thus, the trunnion 120 may protrude a first distance from the outer surface 111 of the body 110 in the non-collapsed state and a second distance from the outer surface 111 of the body 110 in the collapsed state, the first distance being greater than the second distance. In some embodiments the second distance may be zero, or may be negative whereby the trunnion 120 is recessed relative to the outer surface 111 of the body 110 in the collapsed state. Thus, the non-collapsed state may be equivalent or similar to the protruded state described above and the collapsed state may be equivalent or similar to the retracted state described above. Furthermore, the non-collapsed state and the protruded state may be referred to herein as a first state and the collapsed state and the retracted state may be referred to herein as a second state. Referring to FIGS. 8-10, an alternate embodiment of a container 200 is illustrated in accordance with an embodiment of the present invention. The container 200 is cylindrical in shape and defines a cylindrical shaped cavity for storing spent nuclear fuel. The container 200 has a body portion 201 and a lid flange 202 extending from the body portion 201. The lid flange 202 has a smaller diameter than the body portion 201 as shown. A lid 203 is coupled to the lid flange 202. Furthermore, a plurality of trunnions 220 are coupled to the body portion 201 of the container 200. The trunnions 220 are identical to the trunnions 120 described above and thus a detailed description of their structure and function will not be described herein. Rather, the description of the trunnions 120 above is applicable to the trunnions 220. FIG. 10 illustrates the container 200 with a lift yoke 250 coupled thereto in preparation for moving the container 200. Thus, the lift yoke 250 has a first hook arm 251 and a second hook arm 252. The first hook arm 251 engages a first one of the trunnions 220 and the second hook arm 252 engages a second one of the trunnions 220. In that regard, in the exemplified embodiment the first and second ones of the trunnions 220 are spaced apart by approximately 180° about the circumference of the cylindrical body 201 of the container 200. The purpose of FIGS. 8-10 is to illustrate that trunnions of the type described herein can be used with containers of varying shapes and having varying shaped cavities. Referring briefly to FIGS. 1 and 3, in the exemplified embodiment the container 100 also includes a lifting lug 180 coupled to the lid 107. The lifting lug 180 is the structural component of the container 100 that facilitates coupling of the lid 107 to the body 110 of the container 100. Specifically, a crane can be operably coupled to the lifting lug 180 to lift the lid 107 off from the body 110 or place the lid 107 onto the body 110. Referring now to FIGS. 11 and 12A, the lifting lug 180 is illustrated by itself and in cross section, respectively. The lifting lug 180 comprises a first component 181 and a second component 190. Similar to the concepts described above with regard to the trunnions 120, in this embodiment the second component 190 is axially slidable relative to the first component 181. In that regard, the first component 181 of the lifting lug 180 is coupled to the lid 107 and the second component 190 of the lifting lug 180 is coupled to the first component 181 in a manner that permits the second component 190 to axially slide relative to the first component. The second component 190 has a high tensile load bearing capability so that it can support the weight of the lid 107 without breaking while having a low compression load bearing capability such that if a force F2 (FIG. 12B) is applied onto the second component 190 it will cause the second component 190 to slide axially as described further herein below. The first component 181 of the lifting lug 180 protrudes from a top surface of the lid 107 and extends from a first end 182 to a second end 183. The first component 181 also has an inner surface 184 that defines a hollow interior 185. The first component 181 comprises a body portion 186 and a flange portion 187 extending from a top end of the body portion 186 inwardly towards the hollow interior 185. The first end 182 of the first component 180 defines a first opening 188 having a first cross-sectional area and the flange portion 187 defines a second opening 189 having a second cross-sectional area, the first cross-sectional area being greater than the second cross-sectional area. The second component 190 is coupled to the first component 181 and extends from a first end 191 to a second end 192 along a longitudinal axis D-D. The second component 190 is axially slidable between a protruded state illustrated in FIG. 12A and a retracted state illustrated in FIG. 12B. Specifically, in the protruded state a first portion of the second component 190 is located within the hollow interior 185 of the first component 181 and a second portion of the second component 190 protrudes from the second end 183 of the first component 181. Upon application of a sufficient downward or compression force, the second component 190 will axially slide relative to the first component 181 form the protruded state into the retracted state illustrated in FIG. 12B. As shown in FIG. 12A, the second component 190 has a body portion 193 and a flange portion 194 extending from the body portion 193 in a direction away from the outer surface of the body portion 193. When the second component 190 is located within the hollow interior 185 of the first component 181, the flange portion 194 of the second component 190 engages the flange portion 187 of the first component 181 so that a tensile load coupled to the second component 190 will not separate the second component 190 from the first component 181. Moreover, the body portion 193 of the second component 190 has a third cross-sectional area and the flange portion 194 of the second component 190 has a fourth cross-sectional area that is greater than the third cross-sectional area. Furthermore, the fourth cross-sectional area is equal to or less than the first cross-sectional area of the first opening 188 and is greater than the second cross-sectional area of the second opening 189. Thus, the second component 190 can be inserted into the hollow interior 185 of the first component 181 through the first opening 188 in the first end 182 but not through the second opening 189 in the second end 183 because the flange portion 194 will not fit through the second opening 189. This difference in cross-sectional areas of the various regions of the first and second components 181, 190 also maintains the coupling of the first and second components 181, 190 when the second component is carrying a tensile load. Thus, the flange portions 187, 194 of the first and second components 181, 190 of the lifting lug 180 interact to prevent the second component 190 from sliding relative to the first component 181 in a first axial direction. However, a compression force acting on the second component 190 will cause the second component 190 to axially slide relative to the first component 181 in a second axial direction, which is inwardly into the hollow interior 185 of the first component 180. The second component 190 may be capable of being axially slid relative to the first component 181 until the second end 192 of the second component 190 is flush with or recessed relative to the second end 183 of the first component 181. Although described herein as it relates to the lid of a nuclear storage container, the lifting lug 180 described herein can be used for other applications. Specifically, lifting lugs are generally used by serving as a robust tension member and they have uses outside of the nuclear storage industry for lifting a wide variety of apparatuses and equipment. Thus, the lifting lugs 180 described herein with a retractability feature may be used in any instance in which a lifting lug is desired. In that regard, in some embodiments the invention may be directed to the lifting lug 180 itself without regard to its specific application or the specific device to which it is coupled. Thus, the trunnion 120, 220 and lifting lug 180 designs described herein and illustrated in the drawings rely on the concept of direction-dependent stiffness that is engineered into the structure of the component. Specifically, the trunnions 120, 220 can carry a heavy load in a direction perpendicular to their axes but a load or force applied axially will cause the trunnions 120, 220 to collapse as described herein. Thus, the trunnions 120, 220 are a structurally competent member in bending but a weak one in axial tension or compression, which facilitates the collapsible functionality described herein. Similarly, the lifting lug 180 can carry a heavy tensile load, but a compression load will cause the lifting lug 180 to collapse as described herein. The term lifting device may be used herein to refer to either the lifting lug 180 or the trunnion 120, 220. There may be applications where the linear member is required to have a modest load carrying capacity in tension or compression but an assured-to-fail configuration if the applied load is large. For example, fasteners used to support a mattress of crushable material used to serve as an impact mitigator can be made of calibrated axial load carrying capacity to enable an efficient impact limited design. While the invention has been described with respect to specific examples including presently preferred modes of carrying out the invention, those skilled in the art will appreciate that there are numerous variations and permutations of the above described systems and techniques. It is to be understood that other embodiments may be utilized and structural and functional modifications may be made without departing from the scope of the present invention. Thus, the spirit and scope of the invention should be construed broadly as set forth in the appended claims. |
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claims | 1. An extreme ultraviolet light source apparatus, comprising:a collector included in a vessel, the collector configured to reflect extreme ultraviolet light;a baffle assembly included in the vessel, the baffle assembly configured to allow the extreme ultraviolet light reflected from the collector to pass through an internal transmissive region of the baffle assembly, the baffle assembly including a baffle body and a discharge plate, the discharge plate provided in a first end portion of the baffle body adjacent to the collector, the discharge plate configured to collect the target material debris within the baffle body;a guide structure configured to guide target material debris collected in the baffle assembly to a collection tank; anda first heating member and a second heating member, the first heating member provided in the guide structure, the second heating member provided in the discharge plate, the first and second heating members configured to heat the target material debris. 2. The extreme ultraviolet light source apparatus of claim 1, wherein the first heating member comprises at least one heating line which extends from the baffle assembly to the collection tank, and the at least one heating line is included in the guide structure. 3. The extreme ultraviolet light source apparatus of claim 1, wherein the guide structure has a length between 100 mm to 300 mm. 4. The extreme ultraviolet light source apparatus of claim 1, wherein the guide structure extends along a direction of gravity. 5. The extreme ultraviolet light source apparatus of claim 1, wherein the guide structure comprises a guide plate which extends from the baffle assembly to the collection tank. 6. The extreme ultraviolet light source apparatus of claim 1, whereinthe baffle body is between the collector and an outlet port of the vessel. 7. The extreme ultraviolet light source apparatus of claim 1, further comprising:a discharge nozzle installed in a discharge hole formed in the discharge plate, the discharge nozzle configured to be connected to the guide structure. 8. The extreme ultraviolet light source apparatus of claim 7, further comprising:a third heating member included in the discharge nozzle, the third heating member configured to heat the target material debris. 9. An extreme ultraviolet light source apparatus, comprising:a collector included in a vessel, the collector configured to reflect extreme ultraviolet light; anda target debris collection device configured to collect target material debris within the vessel,the target debris collection device comprises,a baffle body between the collector and an outlet port of the vessel, the baffle body configured to allow the extreme ultraviolet light reflected from the collector to pass through an internal transmissive region of the baffle body,a discharge plate provided in a first end portion of the baffle body adjacent to the collector, the discharge plate configured to collect the target material debris within the baffle body,a guide structure configured to guide the target material debris collected in the discharge plate to a collection tank, the guide structure having a length between 100 mm and 300 mm,a first heating member provided in the guide structure, the first heating member configured to heat the target material debris, anda second heating member provided in the discharge plate, the second heating member configured to heat the target material debris. 10. The extreme ultraviolet light source apparatus of claim 9, wherein the first heating member comprises at least one heating line which extends from the discharge plate to the collection tank. 11. The extreme ultraviolet light source apparatus of claim 9, wherein the guide structure extends along a direction of gravity. 12. The extreme ultraviolet light source apparatus of claim 9, wherein the guide structure comprises a guide plate which extends from the discharge plate to the collection tank. 13. The extreme ultraviolet light source apparatus of claim 9, wherein the target debris collection device further comprises a discharge nozzle installed in a discharge hole formed in the discharge plate, the discharge nozzle configured to be connected to the guide structure. 14. The extreme ultraviolet light source apparatus of claim 9, wherein the discharge plate has an annular shape extending along the first end portion of the baffle body. 15. The extreme ultraviolet light source apparatus of claim 9, further comprising:a purge gas supply portion configured to supply a purge gas into the collector. 16. A target debris collection device for extreme ultraviolet light source apparatus, comprising:a baffle body extending within an extreme ultraviolet (EUV) vessel between a collector and an outlet port of the EUV vessel, the baffle body configured to allow extreme ultraviolet light reflected from the collector to pass through an internal transmissive region of the baffle body;a discharge plate provided in a first end portion of the baffle body adjacent to the collector, the discharge plate configured to collect target material debris on an inner surface of the baffle body;a guide structure configured to guide the target material debris collected in the discharge plate to a collection tank;a first heating member provided in the guide structure, the first heating member configured to heat the target material debris; anda second heating member provided in the discharge plate, the second heating member configured to heat the target material debris. 17. The target debris collection device for extreme ultraviolet light source apparatus of claim 16, further comprising:a plurality of vanes extending from the baffle body on an inner wall of the baffle body. 18. The target debris collection device for extreme ultraviolet light source apparatus of claim 16, wherein the first heating member comprises at least one heating line which extends from the discharge plate to the collection tank. 19. The target debris collection device for extreme ultraviolet light source apparatus of claim 16, wherein the guide structure comprises a guide plate which extends from the discharge plate to the collection tank. 20. The target debris collection device for extreme ultraviolet light source apparatus of claim 16, further comprising:a discharge nozzle installed in a discharge hole formed in the discharge plate, the discharge nozzle configured to be connected to the guide structure. |
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claims | 1. A radiation induced growth indication apparatus for the determination of radiation induced growth of a pressurized water reactor nuclear fuel assembly positioned in a reactor core having an upper core support plate, the nuclear fuel assembly including an upper tie plate, said apparatus comprising: an inelastically compressible structural member configured to be disposed between and to be in contact with the upper tie plate and upper core support plate, wherein said compressible structural member is configured to be compressed due to a movement of the upper tie plate relative to the upper core support plate as a result of radiation induced growth of the fuel assembly. 2. The radiation induced growth indication apparatus as in claim 1 wherein the inelastically compressible structural member is a hollow tube. claim 1 3. The radiation induced growth indication apparatus as in claim 2 wherein the hollow tube compresses in response to the application of a compressible force of 10 pounds or less. claim 2 4. The radiation induced growth indication apparatus as in claim 3 wherein the hollow tube has a circumference and further comprises a plurality of holes through the circumference. claim 3 5. The radiation induced growth indication apparatus as in claim 2 wherein the inelastically compressible structural member is a rod. claim 2 6. The radiation induced growth indication apparatus as in claim 5 wherein the rod is positioned at one end into an aperture in an upper surface of the fuel assembly upper tie plate. claim 5 7. The radiation induced growth indication apparatus as in claim 6 wherein the rod compresses in response to the application of a compressive force of 10 pounds or less. claim 6 8. A radiation induced growth indication apparatus for the determination of radiation induced growth of a pressurized water reactor nuclear fuel assembly positioned in a reactor core having an upper core support plate, the nuclear fuel assembly including an upper tie plate, comprising a rod disposed between the upper tie plate and upper core support plate and positioned at a proximate end in an aperture formed into a top of the upper tie plate and a distal end of the rod extending into contact with the upper core support plate for compressing the distal end further into the aperture due to the movement of the fuel assembly against the upper core support plate as a result of radiation induced growth. 9. The radiation induced growth indication apparatus as in claim 8 wherein the aperture in the top of the upper tie plate extends into the upper tie plate a predetermined depth to allow the rod to be pressed into the aperture a distance corresponding to the movement of the fuel assembly toward the upper core support plate as a result of radiation induced growth. claim 8 10. A system comprising: a reactor core having an upper core support plate; a pressurized water reactor nuclear fuel assembly positioned in said reactor core and having an upper tie plate; and an inelastically compressible structural member disposed between and in contact with the upper tie plate and upper core support plate, wherein said compressible structural member is configured to be compressed due to a movement of the upper tie plate relative to the upper core support plate as a result of radiation-induced growth of the fuel assembly. 11. The system as in claim 10 wherein the inelastically compressible structural member is a hollow tube. claim 10 12. The system as in claim 2 wherein the hollow tube compresses in response to the application of a compressible force of 10 pounds or less. claim 2 13. The system as in claim 3 wherein the hollow tube has a circumference and further comprises a plurality of holes through the circumference. claim 3 |
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048636754 | abstract | A nuclear power system comprises a plurality of modules disposed in below-grade pits to provide a compact, self-contained nuclear power supply. The modules are preferably individually transportable so that they may be substantially preassembled prior to installation. The system operates at relatively low temperatures and pressures, and includes various safety features which would prevent radioactive contamination of the surrounding environment in the event of a disturbance causing rupture of one or more of the odules or the pipes interconnecting the modules. The system also provides a low resistance flow path for vapor discharged from the turbine to improve efficiency. |
046541900 | summary | FIELD OF THE INVENTION The present invention relates to an emergency feedwater system to provide emergency feedwater to the steam generators of a pressurized water reactor so as to cool the reactor in the event of a failure of a main feedwater system to a steam generator. BACKGROUND OF THE INVENTION In pressurized water reactors for the nuclear production of power, a pressurized fluid is passed through the reactor core and, after being heated in the core, is passed through heat transfer tubes that are positioned in a secondary side of a steam generator. In the secondary side of the steam generator, the heat transfer tubes transfer heat to a secondary fluid to produce steam that is then used to operate a turbine for production of electrical power. The provision of emergency feedwater systems for the secondary side of the steam generators of a pressurized water reactor is made in order to supply feedwater to the steam generators following an accident or transient conditions when the main feedwater system is not available, thereby maintaining the capability of the steam generators to remove plant stored heat and reactor core decay heat by converting the emergency feedwater to steam which may then either be discharged to the condensor or to the atmosphere. Such emergency feedwater systems generally comprise a source of emergency feedwater, such as a supply of water contained in a storage tank, and associated lines, pumps and valving systems to direct the emergency feedwater, when necessary, to the steam condensors. In order to assure operation of the system under various adverse conditions, means must be provided to effect operation of the emergency feedwater system for example in the event of loss of electrical power, or in the event of a passive failure such as a pipe rupture or an active failure such as a failure of a valve to respond to a signal to open or close. In addition, provision should be made to address even remote possibilities of interruption of an emergency feedwater system, such as fires or other external events, such as air craft impacts, explosions, or the like, which might impair the operability of the system. SUMMARY OF THE INVENTION An emergency feedwater system for the steam generators of a pressurized water nuclear reactor power plant contains two separately located subsystems. Each subsystem comprises an emergency feedwater supply tank, and a pair of emergency feedwater lines for discharge of water from the tank to the inlet line of a steam generator. An electrically operated motor driven pump is located in one of said pair of emergency feedwater lines, and a steam driven turbine pump is located in the other of said pair of emergency feedwater lines. A cavitating venturi is provided in the emergency feedwater line between the pumps and the inlet line to a steam generator. In the embodiment for a four loop system, a cavitating venturi is provided in each of the emergency feedwater lines, with one feedwater line provided for each of four steam generators, and connecting lines are provided between each pair of emergency feedwater lines. In the embodiment for a three loop system, one of each of said pair of emergency feedwater lines of a subsystem charges emergency feedwater to separate first and second steam generators, while the other of each of said pair of emergency feedwater lines combine to form a common discharge line to a third steam generator. A cavitating venturi is provided in each of said feedwater lines to the first and second steam generators and a further cavitating venturi is provided in the common discharge line. In the embodiment for a two loop system, each pair of emergency feedwater lines of a subsystem combine to form a common discharge line to separate ones of the two steam generators, and a cavitating venturi is provided in each of the two common discharge lines. |
abstract | An improved retention system for retaining fuel rods in a fuel assembly is disclosed. The retention system includes a plurality of first engagement surfaces on the bottom nozzle of a fuel assembly. There is at least one engagement surface for each fuel rod. A second engagement surface is formed on the bottom end plug of each fuel rod. The first and second engagement surfaces are configured for engagement with each other for axially and laterally retaining each fuel rod within the fuel assembly. Debris deflectors may also be provided to deflect debris from coolant channels surrounding the fuel rods. |
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051620979 | summary | TECHNICAL FIELD This invention relates generally to nuclear reactors, and, more particularly, to a reactor having improved fuel arrangements in a reactor core. BACKGROUND ART Fission reactors rely on fissioning of fissile atoms such as uranium isotopes (U.sup.233, U.sup.235) and plutonium isotopes (Pu.sup.239, Pu.sup.241). Upon absorption of a neutron, a fissile atom can disintegrate, yielding atoms of lower atomic weight and high kinetic energy along with several high-energy neutrons. The kinetic energy of the fission products is quickly dissipated as heat, which is the primary energy product of nuclear reactors. Some of the neutrons released during disintegration can be absorbed by other fissile atoms, causing a chain reaction of disintegration and heat generation. The fissile atoms in nuclear reactors are arranged so that the chain reaction can be self-sustaining. To facilitate handling, fissile fuel is typically maintained in modular units. These units can be bundles of vertically extending fuel rods. Each rod has a cladding which encloses a stack of fissile fuel pellets. Generally, each rod includes a space or "plenum" for accumulating gaseous byproducts of fission reactions which might otherwise unacceptably pressurize the rod and lead to its rupture. The bundles are arranged in a two-dimensional array in the reactor to form a "core". Neutron-absorbing control rods are inserted between or within fuel bundles to control the reactivity of the core. The reactivity of the core can be adjusted by incremental insertions and withdrawals of the control rods. Both economic and safety considerations favor improved fuel utilization, which can mean less frequent refueling and less exposure to radiation from a reactor interior. In addition, improved fuel utilization generally implies more complete fuel "burnups", or fissioning. A major obstacle to obtaining long fuel element lifetimes and complete fuel burnups is the inhomogeneities of the neutron flux both radially and axially throughout the core. For example, fuel bundles near the center of the core are surrounded by other fuel elements. Accordingly, the neutron flux at these central fuel bundles exceeds the neutron flux at peripheral fuel bundles which have one or more sides facing away from the rest of the fuel elements. Therefore, peripheral fuel bundles tend to burn up more slowly than do the more central fuel bundles. The problem of flux density variations with radial core position has been addressed by repositioning fuel bundles between central and peripheral positions. This results in extended fuel bundle lifetimes at the expense of additional refueling operations. Variations in neutron flux density occur in the axial direction as well as the radial direction. For example, fuel near the top or bottom of a fuel bundle is subjected to less neutron flux than is fuel located midway up a fuel bundle. These axial variations are not effectively addressed by radial redistribution of fuel elements. In addition to the variations in neutron flux density, variations in spectral distribution affect burnup. For example, in a boiling-water reactor (BWR), neutrons released during fissioning move too quickly and have too high an energy to readily induce the further fissioning required to sustain a chain reaction. These high energy neutrons are known as "fast" neutrons. Slower neutrons, referred to ask "thermal neutrons", most readily induce fission. In BWRs, thermal neutrons are formerly fast neutrons that have been slowly primarily through collisions with hydrogen atoms in the water (moderator) used as the heat transfer medium. Between the energy levels of thermal and fast neutrons are "epi-thermal" neutrons. Epithermal neutrons exceed the desired energy for inducing fission but promote resonance absorption by many actinide series isotopes, converting some "fertile" isotopes to "fissile" (fissionable) isotopes. For example, epithermal neutrons are effective at converting fertile U.sup.238 to fissile Pu.sup.239. Within a core, the percentages of thermal, epithermal and fast neutrons vary over the axial extent of the core. Axial variations in neutron spectra are caused in part by variations in the density or void fraction of the water flowing up the core. In a boiling-water reactor (BWR), water entering the bottom of a core is essentially completely in the liquid phase. Water flowing up through the core boils, so most of the volume of water exiting the top of the core is in the vapor phase, i.e., steam. Steam is less effective than liquid water as a neutron moderator due to the lower density of the vapor phase. Therefore, from the point of view of neutron moderation, core volumes occupied by steam are considered "voids"; the amount of steam at any spatial region in the core can be characterized by a "void fraction". Within a fuel bundle, the void fraction can vary from about zero at the base to about 0.7 near the top. Continuing the example for the BWR, near the bottom of a fuel bundle, neutron generation and density are relatively low, but the percentage of thermal neutrons is high because of the moderation provided by the low void fraction water at that level. Higher up, neutron density reaches its maximum, while void fraction continues to climb. Thus, the density of thermal neutrons peaks somewhere near the lower-middle level of the bundle. Above this level, neutron density remains roughly stable while the percentages of epithermal and fast neutrons increase. Near the top of the bundle, neutron density decreases across the spectrum since there are no neutrons being generated just above the top of the bundle. The inhomogeneities induced by this spectral distribution can cause a variety of related problems. Focusing on the upper-middle section, problems of inadequate burnup and increased production of high-level transuranic waste are of concern. Since the upper-middle section has a relatively low percentage of thermal neutrons, a higher concentration of fissile fuel is sometimes used to support a chain reaction. If the fuel bundle has a uniform fissile fuel distribution, this section could fall below criticality (the level required to sustain a chain reaction) before the other bundle sections. The fuel bundle would have to be replaced long before the fissile fuel in all sections of the bundle were depleted, wasting fuel. The problem with waste disposal is further aggravated at this upper-middle section since the relatively high level of epithermal neutrons results in increased production of actinide-series elements such as neptunium, plutonium, americium, and curium, which end up as high level-waste. One method of dealing with axial spectral variations is using a control rod. For the BWR, control rods typically extend into the core from below and contain neutron-absorbing material which robs the adjacent fuel of thermal neutrons which would otherwise be available for fissioning. Thus, control rods can be used to modify the distribution of thermal neutrons over axial position to achieve more complete burnups. However, control rods provide only a gross level of control over spectral density. More precise compensation for spectral variations can be implemented using enrichment variation and burnable poisons. Enrichment variation using, for example, U.sup.235 enriched uranium, can be used near the top of a fuel bundle to partially compensate for a localized lack of thermal neutrons. Similarly, burnable poisons such as gadolinium oxide (Gd.sub.2 O.sub.3), can balance the exposure of bundle sections receiving a high thermal neutron flux. Over time, the burnable poisons are converted to isotopes which are not poisons so that more thermal neutrons become available for fissioning as the amount of fissile material decreases. In this way, fissioning can remain more constant over time in a section of the fuel bundle. By varying the amount of enrichment and burnable poisons by axial position along a bundle, longer and more complete burnups can be achieved. In addition, the enrichment and poison profiles can be varied by radial position to compensate for radial variations in thermal neutron density. Nonetheless, taken together, the use of control rods, radial positional exchange of bundles, selective enrichment and distribution of burnable poisons still leave problems with axial variations in burn rates and neutron spectra. Furthermore, none of these employed methods effectively address the problem of the high level of fissile material produced and left in the upper-middle sections of the bundle due to the high level of epithermal neutrons and the low level of thermal neutrons. What is needed is a system that deals more effectively with axial spectral variations in neutron flux so that higher fuel burnups are provided and so that high-level waste is minimized. OBJECTS OF THE INVENTION A major objective of the present invention is to provide for more thorough fuel burnups to enhance fuel utilization and minimize active waste products. Another object of the present invention is to provide a new and improved reactor effective for using axial variations in neutron flux density and in neutron spectral distribution for both converting fertile fuel to fissile fuel, and providing more uniform and complete fuel fissioning during the life of the fuel in the reactor core. DISCLOSURE OF INVENTION In accordance with the present invention, a nuclear reactor with a recirculating heat transfer fluid has a bi-level core which provides enhanced flexibility in fuel arrangement. The bi-level core includes two sets of fuel units, one set arranged on a first level, the other set arranged on a second level. Preferably, fuel units of the second level are arranged in vertical alignment with fuel units of the first level. This permits a fuel unit of the first level to be accessed by removing only the adjacent fuel unit of the second level. During refueling operations, fuel units can be shifted from one level to the other, providing additional flexibility in arranging units at various stages of burnup. Preferably, fuel units of the first level are inverted relative to the fuel units of the second level. The inversion provides for placing plenum sections of fuel rods in different levels away from each other so that the plenums do not introduce a discontinuity in neutron generation, and allows for more uniform axial fuel burnup. The bi-level core allows fuel to be initially positioned in the second level for conversion of fertile fuel to fissile fuel, and then repositioned to the first level for more complete axial burnup. In a preferred embodiment the first level boils water to generate saturated steam, and the second level is cooled by the saturated steam and generates superheated steam. |
051075261 | abstract | A high resolution x-ray microscope for imaging microscopic structures within biological specimens has an optical system including a highly polished primary and secondary mirror coated with identical multilayer coatings, the mirrors acting at normal incidence. The coatings have a high reflectivity in the narrow wave bandpass between 23.3 and 43.7 angstroms and have low reflectivity outside of this range. The primary mirror has a spherical concave surface and the secondary mirror has a spherical convex surface. The radii of the mirrors are concentric about a common center of curvature on the optical axis of the microscope extending from the object focal plane to the image focal plane. The primary mirror has an annular configuration with a central aperture and the secondary mirror is positioned between the primary mirror and the center of curvature for reflecting radiation through the apertture to a detector. An x-ray filter is mounted at the stage end of the microscope, and film sensitive to x-rays in the desired band width is mounted in a camera at the image plane of the optical system. The microscope is mounted within a vacuum chamber for minimizing the absorption of x-rays in air from a source through the microscope. |
051436906 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS In accordance with the present invention, a natural-circulation boiling-water reactor 100 comprises a vessel 102, a core 104, a chimney 106, a steam separator 108, and a dryer 110. Control rod drive housings 112 extend through the bottom of vessel 102 and support control rod guide tubes 113. Control rod guide tubes 113 extend to the bottom of core 104 so that control blades therein can be inserted into and retracted from core 104 to control its power output. Water flows, as indicated by arrows 114, into core 104 from below. This subcooled water is boiled within core 104 to yield a water/steam mixture which rises through core 104 and chimney 106, as indicated by arrow 115. Steam separator 108 helps separate steam from water, and the released steam exits through a steam exit 116 near the top of vessel 102. Before exiting, any remaining water entrained in the steam is removed by dryer 110. Water is returned down peripheral downcomer 118 by the force of the driving steam head provided by chimney 106. Feedwater enters vessel 102 through a feedwater inlet nozzle 120 and feedwater sparger 122 to replenish and to help cool the recirculating water in downcomer 118. Core 104 is bounded from below by a core support plate 124, along with associated orificed support stubs 126, and bounded from above by a top guide 128. These structures support and aid in the installation of fuel assemblies 130 that constitute core 104. Fuel assemblies 130 are arranged in a two-dimensional array, as shown in FIG. 2. Spaces are left between groups of four fuel assemblies for control rods 232 with cruciform cross sections to move vertically to regulate power output. As schematically indicated in FIG. 3, each fuel bundle 130 comprises a channel wall 310 and a fuel bundle 301. Fuel bundle 301 comprises a top tie plate 306, a bottom tie plate 308, and fuel rods 302. Channel wall 310 defines a coolant channel 304 through which coolant for rods 302 can flow. Fuel rods 302 are arranged in a 15.times.15 array within channel 304 of bundle 130. Channel 304 has a square cross section, which defines the square cross section of fuel assembly 130 (as shown in FIG. 2). Not all positions within the array are filled with fuel rods 302. Some positions are left open to provide additional coolant throughput and to optimize the fission characteristics of bundle 301. Channel wall 310 extends between tie plates 306 and 308. wall 310 defines a channel for coolant to flow through. Tie plates 306 and 308 maintain fuel rods 302 in their respective array positions, yet allow for vertical expansion, which occurs as temperatures rise. Tie plates 306 and 308 have respective ridges 316 and 318. These ridges are designed to seat securely on support stub 126. Thus, assembly 130 can be supported in either an inverted orientation (in which case ridge 316 engages support stub 126) or an uninverted orientation (in which case, ridge 318 engages support stub 126, as shown). Each tie plate 306 and 308 includes respective pairs of radial holes 326 and 328, for admitting prongs of a refueling tool 330, shown about to engage holes 326 in FIG. 3. Refueling tool 330 aids in the insertion, removal, movement and inversion of fuel assembly 130. Refueling tool 330 is inserted from above while reactor 100 is shut down and the top of vessel 102 is removed. Thus, both tie plates 306 and 308 provide for both seating on support stub 126 and for manipulation by tool 300. This achieves a measure of symmetry required for the inversions of the present invention. The symmetry corresponding to bundle inversion is a 180.degree. rotational symmetry about a line perpendicular to the vertical axis of bundle 130. It is generally desirable to have a greater pressure drop at the downstream end of a fuel bundle than at the upstream end. To this end, prior art fuel bundles including a constrictive orifice at their base. However, if built into the fuel bundle, such an orifice would not meet the symmetry requirement. Accordingly, support stub 126 defines the required constriction 332, which remains in place however fuel bundle 130 is oriented. Despite the smoothing effects of inversion, different positions within a fuel bundle 301 are exposed to different conditions. The fuel near the top of the bundle 301 undergoes a different history than does the fuel at the bottom. Accordingly, fuel 140 in rods 302 is distributed non-uniformly. For example, more fertile fuel can be located near the top of bundle 301. Fertile fuel at the bottom of bundle 301 is not converted during a first cycle, and does not contribute significantly to power during the second cycle. Bottom fertile fuel is converted during the second cycle and is available during an optional third cycle. However, since further cycles are not anticipated, it is not desirable to generate too much fissile actinide products, since those that do not burn up wind up as long-term radioactive waste. Mechanically, each fuel rod 302 has a top plenum 342 and a bottom plenum 344 on both sides of fuel 346. These plenums 342 and 344 are designed to accommodate gaseous fission products that escape the fuel. Plenums are incorporated at both ends to provide symmetry in the fuel position for fuel bundle 130. Top plenum 342 houses a spring 348 that helps compact fuel 346 while permitting thermal expansion. Bottom plenum 346 houses a ventilated pedestal 350 that keeps fuel 346 out. Pedestal 350 also opposes the force of spring 348 to maintain compression of fuel 346. The use of a spring in one plenum and a pedestal in the opposing plenum, and the non-uniform distribution of fuel 356 are examples of acceptable asymmetry in fuel assembly 130. In accordance with the present invention, a fuel management method 400 for reactor 100 involves inverting fuel assemblies between operational cycles. Preparatory steps include assembly of fuel rods and of fuel assembly. During a refueling operation, a fuel assembly is inserted, at step 401, top side up into core 104. Once refueling is complete, reactor 100 is operated, at step 402, for a first cycle. After this cycle and any additional refueling operations and operating cycles not involving inversion of this specific fuel assembly, the reactor is shut down. During the next refueling operation, the fuel assembly is inverted, at step 403, and replaced in its original position in the core. The present invention also provides for changing the array position of the fuel assembly in accordance with other refueling strategy considerations. A second operational cycle is implemented, at step 404, followed by a respective refueling shutdown. At this point, there are two major alternatives provided by the present invention. The fuel assembly can be disposed of, branch 410, or it can be inverted again and reinstalled, branch 411. In the latter case, the second inversion, at step 405, restores the original orientation of the fuel assembly. One purpose of this reversion is the burn up of actinide fissile material generated during the second operating cycle. After this inversion, a third operating cycle and shutdown, step 406, is implemented. During a following refueling operation, the fuel assembly is removed and disposed of, step 407. A new fuel assembly is added to the core in its place. Where the reversion is not implemented, steps 405 and 406 are skipped and step 407 follows step 404. The branches of method 400 can be practiced in the alternative or together with respect to different assemblies in the same core. The election of two versus three cycles depends on core position of a fuel assembly and the distribution of fuel in the bundle. A major consideration is the amount of fissile fuel generated during the second operating cycle. These and other modifications to and variations upon the described embodiments are provided for by the present invention, the scope of which is limited only by the following claims. |
description | The invention relates to repairing or replacing a bimetallic weld in a jet pump diffuser assembly and, more particularly, to a jet pump diffuser clamping assembly that structurally replaces/repairs the weld. In a boiling water nuclear reactor, hollow tubular jet pumps positioned within the shroud annulus provide the required reactor core flow. The lower portion of the jet pump is a long conical tubular section. The function of the diffuser is to slow the high velocity flow stream and thus convert the dynamic head of the flow stream into static pressure. The jet pump diffuser assemblies in boiling water reactors can be one of two differing design configurations. The first configuration incorporates an adapter component, which is comprised of an upper austenitic stainless steel ring and a lower inconel alloy 600 ring. These two rings were shop-welded, as this was a bimetallic weld (a weld of dissimilar metals) and more difficult to perform than a weld joining similar metals. This adapter component then facilitated the assembly of the diffuser in the reactor, as the lower section of the adapter was welded to the inconel alloy 600 shroud support plate. The diffuser tail pipe being constructed of austenitic stainless steel was then joined to the upper section of the adapter, which was constructed of austenitic stainless steel. As such, at the time of field construction of the reactor these two welds of similar metals were performed. The second design configuration consists of a lower ring constructed of inconel alloy 600, shop-welded to the diffuser tail pipe. Since the diffuser tail pipe is fabricated from austenitic stainless steel, this was a bimetallic weld. This design approach results in the welding of the lower ring to the shroud support plate as the only weld required to be performed at the time of field construction of the reactor. Regardless of diffuser assembly design configuration under consideration, a bimetallic weld is present in the diffuser assembly. In the event that the structural integrity of the bimetallic weld of the diffuser assembly should become degraded, a means of reinforcing or structurally replacing this weld is desired. In an exemplary embodiment of the invention, a clamp assembly connects a diffuser adapter or lower ring to a diffuser tail pipe of a jet pump diffuser in a boiling water nuclear reactor. The clamp assembly includes at least two clamp segments shaped generally corresponding to an exterior circumference of the diffuser, a swivel link affixed at each end of each of the clamp segments, and at least two connecting bands pivotably secured to the swivel links between the ends of the clamp segments. The clamp segments each includes a locking assembly engageable with the diffuser adapter or lower ring and the diffuser tail pipe. In another exemplary embodiment of the invention, a method of repairing a weld joining a diffuser adapter or lower ring to a diffuser tail pipe of a jet pump diffuser in a boiling water nuclear reactor includes the steps of forming a plurality of holes in the diffuser adapter or lower ring and the diffuser tail pipe; positioning two clamp segments each including a clamp body and upper and lower pin inserts on the diffuser such that the upper pin inserts engage the holes in the diffuser tail pipe and such that the lower pin inserts engage the holes in the diffuser adapter or lower ring; connecting a swivel link at each end of each of the clamp segments; securing two connecting bands to the swivel links between the ends of the clamp segments; and tightening the connection between the clamp segments and the swivel links to ensure proper fit-up and substantially equal distribution of loads. In still another exemplary embodiment of the invention, the clamp assembly includes two clamp segments shaped generally corresponding to an exterior circumference of the diffuser. Each clamp segment includes a clamp body housing an upper pin insert and a lower pin insert. The upper and lower pin inserts have pins engageable with corresponding holes formed in the diffuser adapter or lower ring and the diffuser tail pipe. A swivel link is affixed at each end of each of the clamp segments, and at least two connecting bands are pivotably secured to the swivel links between the ends of the clamp segments. The jet pump diffuser clamp assembly 10 is shown installed on a jet pump diffuser PD in FIGS. 1-3. Components of the clamp assembly 10 include clamp segments 12 shaped generally corresponding to an exterior circumference of the diffuser PD, swivel links 14 affixed at each end of the clamp segments 12, and connecting bands 16 pivotably secured to the swivel links 14 between the ends of the clamp segments 12. As shown in FIGS. 4, 6 and 7, the clamp segments 12 are comprised of a clamp body 18 and a pair of pin inserts, including an upper pin insert 20 and a lower pin insert 22. The clamp bodies 18 each include a pair of circumferential channels 24 on an inside surface thereof that are sized to receive the upper and lower pin inserts 20, 22. The pin inserts 20, 22 include a plurality of pins 26 engageable with corresponding holes H formed in the diffuser adapter or lower ring DA and the diffuser tailpipe TP. Preferably, the pins 26 are conically shaped, and the holes H formed in the diffuser adapter or lower ring and the diffuser tailpipe are correspondingly tapered. The preferably conical pins and mating tapered holes are sized such that the pins seat in the tapered holes (i.e. the pin inserts and clamp bodies do not bottom-out or touch the diffuser when the clamp assembly bolts are tightened). As such, leakage of coolant is minimized, since the conical pins theoretically seal the tapered holes. Although cylindrical pins could be used, and the invention is not meant to be limited to conical pins per se, if the pins were cylindrical and they fit into cylindrical holes, there would need to be a clearance between the pins and holes in order to facilitate assembly. The radial clearance would therefore provide a leakage path at every pin-hole interface. In an attempt to minimize leakage, the conical pins and tapered holes are provided. With continued reference to FIGS. 1 and 6, each of the clamp segments 12 includes a bolt collar 28 at each end thereof including at least one aperture 30 for receiving a bolt 32 engageable with a corresponding one of the swivel links 14. Ends of the connecting bands 16 include openings 34 therein that are aligned between openings 36 in the swivel links 14. A pin 38 is fit through the aligned openings to secure the connecting bands 16 within the swivel link. Each of the bolts 32 and the pins 38 preferably also include crimp collars 40 to ensure fastener retention. Since the diffuser tailpipe TP and the diffuser adapter or lower ring DA are typically formed of different materials having different coefficients of thermal expansion, it is desirable to match the coefficients of thermal expansion between any interfacing components. As such, the upper pin insert 20, which is engageable with the diffuser tailpipe TP, is formed of a first material having a coefficient of thermal expansion substantially matching that of the diffuser tailpipe TP. Correspondingly, the lower pin insert 22, which is engageable with the diffuser adapter or lower ring DA, is formed of a second material having a coefficient of thermal expansion substantially matching that of the diffuser adapter or lower ring DA. In one exemplary arrangement, the first material of the upper pin insert is type 316 austenitic stainless steel, and the second material of the lower pin insert is inconel alloy 600. In a similar context, since the clamp body 18 spans across both the diffuser tailpipe TP and the diffuser adapter or lower ring DA, it is desirable to engineer the thermal expansion of the clamp body to be substantially equivalent to the combined thermal expansion of the diffuser tailpipe TP and the diffuser adapter or lower ring DA at operating temperatures. As such, the clamp body 18 is formed of a material having a coefficient of thermal expansion that is intermediate between the coefficients of thermal expansion of austenitic stainless steel (upper section of the adapter and diffuser tail pipe) and inconel alloy 600 (lower section of the adapter or lower ring). In an exemplary embodiment of the invention, the material of the clamp body 18 is type XM-19 austenitic stainless steel (nitronic 50). As a consequence, as the reactor heats up from ambient to operating temperature, the pins 26 expand in concert with the expanding holes H, and since the geometry of the pins 26 substantially matches the geometry of the holes H, leakage is minimized and thermal stresses are not imposed on clamp and diffuser components. Moreover, thermal expansion of the clamp body 18 is engineered to be equivalent to the combined thermal expansion of the jet pump tailpipe TP and adapter piece or lower ring. DA. As a result, thermal stresses at the bimetallic weld of the adapter are alleviated. The holes H, which are conical in a preferred exemplary embodiment of the invention, may be formed via electric discharge machining (EDM) in the tailpipe TP and the adapter or lower ring DA components, as shown in FIG. 5. The holes H include a row of holes in the tailpipe TP and a row of holes directly below in the adapter or lower ring DA. The center lines of the holes H in each row of holes are preferably equidistant from each other and lie in a common plane. Mirrored holes are also provided in the opposite side of the tailpipe TP and adapter or lower ring DA as shown in FIG. 6. The relative distance of the holes above and below the bimetallic weld is also a function of the coefficients of thermal expansion of the various materials. In an exemplary embodiment, the holes H are approximately ¾ (three fourths) inch in diameter with a total taper angle of 10°. Twenty-eight holes are shown, fourteen of which are machined in one side of the diffuser, and the other fourteen of which are machined in the other side of the diffuser. The holes H in any given side of the jet pump are preferably machined simultaneously in order to preserve the hole spacing above and below the weld line. The distance of these holes above and below the weld and the material selection of the clamp body components minimizes the introduction of additional thermal stress in the weld. The repair clamp installation process is shown in FIG. 7. Once the holes H are machined in the jet pump diffuser, clamp components are installed. In a preferred exemplary embodiment, pre-assembled subassemblies including the clamp segments 12, swivel links 14, bolts 32, and bolt crimp collars 40, are installed onto each side of the jet pump diffuser PD such that the pins of the upper and lower pin inserts 20, 22 interface with the holes H in the jet pump diffuser. Subsequently, the connecting bands 16 are installed and secured to the swivel links 14 at each end of the connecting bands by virtue of the pins 38. The bolts 32 are then tightened sequentially to ensure proper fit-up and equal distribution of loads in the clamp components. Finally, the pin and bolt crimp collars 40 are crimped to ensure fastener retention. The clamp assembly described herein serves to structurally repair/replace the weld connecting the jet pump adapter or lower ring to the jet pump tailpipe. Since the clamp assembly is designed to structurally replace the attachment bimetallic weld, it is not necessary that the existing weld be accessible for visual inspection after the repair clamp has been installed. The clamp assembly is remotely installable in the reactor. The clamp assembly is simplified in design and installation, resulting in a superior alternative for jet pump diffuser repair installations. While the invention has been described in connection with what is presently considered to be the most practical and preferred embodiments, it is to be understood that the invention is not to be limited to the disclosed embodiments, but on the contrary, is intended to cover various modifications and equivalent arrangements included within the spirit and scope of the appended claims. |
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abstract | A water jet peening apparatus includes: a nozzle, which is arranged in water and has a mouth from which water is jetted out; a detecting device, which is arranged in the water and detects sound in at least a part of a period during which the water is being jetted out from the mouth; and a processing device, which determines, based on a result of the detection by the detecting device, presence or absence of abnormality in the nozzle. |
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description | The invention relates to maintenance schedules and costs. More particularly, the invention relates to a system and process for determining schedules that minimize maintenance costs of fleet management programs. An important aspect of cost-effectively executing a Fleet Management Program (hereinafter “FMP”) is determining maintenance schedules (MS) that minimize maintenance costs such as the cost of maintenance work and of parts of an engine over a longer time interval. Maintenance schedules depend upon a number of parameters such as maintenance time intervals, thresholds for changing or upgrading engine components, etc. The challenge lies in determining which parameter values minimize the resultant maintenance costs of an engine during an interval of the lifetime of the engine. This determination is generally characterized as a multi-objective global optimization problem, where the objectives are the mentioned costs. For improved cost predictions it may be useful to take into account not only the current state of an engine to be serviced but to consider also an estimated cost incurred at future shop visits during the lifetime of the engine. Moreover, the cost predictions may be further improved by taking into account future probabilities of failure during the service life of the engine. These aspects are usually handled by simulations of the service life of the engine taking into account stochastic events. Thus, the multi objective global optimization problem of optimizing a maintenance schedule has in addition both a time dependent characteristic and a stochastic characteristic. Moreover, a maintenance schedule by itself may be a quite complex and long running program that considers different combinations of parts to be replaced and different types of work that may be implemented and that takes into account the usage of the parts and other aspects of maintenance to propose a current best set of maintenance decisions. Such maintenance decisions may include what parts have to be replaced and what maintenance work has to be performed. The optimization problem is nonlinear and has constraints. Generally, the multi-objective global optimization problem is a mixed integer problem since some optimization parameters may be integer variables, such as decisions to be taken or not taken, or continuous variables such as time limits. All these aspects of the multi-objective global optimization problem usually make the optimization problem hard to solve. Difficulties with classical optimization approaches when facing such problems are known to one of ordinary skill in the art. For example, classical exhaustive optimization techniques can provide high confidence that the best solutions are found, such as by enumerations; grid searches, or graph searches; however, these techniques require too large computational time especially when considering a larger number of parameters. In contrast, branch and bound techniques are known to have difficulties handling multiple objectives. Stochastic approaches, such as evolutionary algorithms, simulated annealing, genetic algorithms, tend to find local optima and may provide limited confidence that global optima are found, while gradient based or pattern search approaches tend to find a local optimum, etc. Therefore, for the optimization of maintenance schedules of aircraft engines, there exists a need for a multi objective optimization procedure that takes into account the time dependent and stochastic nature of the problem and that is designed to overcome the above mentioned difficulties. In accordance with one aspect of the present disclosure, a process for optimizing maintenance work schedules for at least one engine broadly comprises retrieving at least one set of data for an engine from a computer readable storage medium; selecting at least one scheduling parameter for the engine; selecting a set of maintenance rules for the engine; selecting at least one maintenance work decision; selecting at least one objective for said engine; optimizing the at least one objective to generate at least one optimal maintenance work decision; and generating at least one optimal maintenance work schedule for the engine. In accordance with another aspect of the present disclosure, a system comprising a computer readable storage device readable by the system, tangibly embodying a program having a set of instructions executable by the system to perform the following steps for optimizing maintenance work schedules for at least one engine, the system broadly comprising means for retrieving at least one set of data for an engine from a computer readable storage device; means for selecting at least one scheduling parameter for the engine; means for selecting a set of maintenance rules for the engine; means for selecting at least one maintenance work decision; means for selecting at least one objective for the engine; means for optimizing the at least one objective to generate at least one optimal maintenance work decision; and means for generating at least one optimal maintenance work schedule for the at least one engine. In accordance with yet another aspect of the present disclosure, a process for optimizing maintenance work schedules for at least one industrial system broadly comprises retrieving at least one set of data for at least one industrial system from a computer readable storage medium; selecting at least one scheduling parameter for the at least one industrial system; selecting a set of maintenance rules for the at least one industrial system; selecting at least one maintenance work decision for the at least one industrial system; selecting at least one objective for the at least one industrial system; optimizing the at least one objective to generate at least one optimal maintenance work decision for the at least one industrial system; and generating at least one optimal maintenance work schedule for the at least one industrial system. 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. Like reference numbers and designations in the various drawings indicate like elements. Maintenance schedules for industrial systems of parts, such as engines, are plans or programs for repairing or replacing system parts at given time intervals or when certain events (e.g., failures) happen. The MS module decides what parts may be replaced and what maintenance work may be performed on an engine, at a specified shop visit. Generally, a MS receives as input a system or engine in a given state, defined by measures of the life of the engine parts, by engine measurements, and by a maintenance history of the engine. Based on MS parameters X, the MS produces as output another state of the engine, providing parts to be upgraded or changed, and maintenance work to be performed on the engine. In addition, MS returns a scheduled maintenance date for the output engine, and other information such as costs of parts and work performed. Referring to FIG. 1 as an example, a multi-objective global maintenance schedule optimizer system and process described herein may be embodied tangibly in a computer readable medium 101, loaded within a computer 100 as discussed later in greater detail, and/or stored in a computer readable storage medium 102 and executed using the computer 100. Generally, the exemplary system and process may receive as input 108, 112 an engine 106 or industrial system 104, respectively, in the aforementioned given states. Based on MS parameters X, the exemplary system and method produces as output 110, 114, another state of the engine 118 or industrial system 116, respectively, as described above. In addition, the exemplary system and process returns a scheduled maintenance date for the output engine 118 or industrial system 116 and the aforementioned information. To achieve these results an MS may use different algorithms such as stochastic optimization, combinatorial searches, exhaustive searches, heuristic decision rules, or other optimization approaches known to one of ordinary skill in the art. The particular ways by which an MS obtains its decisions are not particularly relevant. The decision parameters X may be discrete or continuous such as: given time thresholds for future shop visits, engine operation variables or thresholds, part life thresholds, part upgrade thresholds, or parameters of decision rules used for maintenance. A main purpose of the optimization of the MS described herein is to find the values of the parameters X that will produce an MS such that the total expected engine maintenance cost, including part costs and work costs incurred in a given time interval such as a contract interval, should be minimal. For fixed values of X, such costs may be computed in an average way, for example using Monte Carlo techniques to simulate a set of engine life realizations and averaging the costs of said realizations. The averaging may take into account the probabilities that a given life realization will happen, for example, taking into account the probabilities of failure associated with events that happened in a particular life realization of the engine. Such formulations are known to one of ordinary skill in the art, for example, multiplying the costs incurred at a given shop visit by the probability by which said shop visit happened. The simulation of the engine life realizations may be performed by an engine life simulator. More particularly, the engine life simulator module may simulate time sequences of shop visits in the life of an engine. Such shop visits may be either scheduled by the MS or may be random, for example, determined by failure probabilities. The estimation of the different costs may be performed by the cost evaluator. Hence, the cost evaluator may estimate one objective or several objectives O1(X), . . . , On(X) for the given parameters X, using the costs and probabilities obtained from a set of engine life simulations. The set of objectives O1(X), . . . , On(X) may represent, for example, the cost(s) of part(s), and the cost(s) of work performed over a time interval of the life of an engine and associated costs, contract related costs, and the like. The optimization module may incorporate a multi-objective optimizer described herein that finds values of parameters X that minimize the objectives O1(X), . . . , On(X). An exemplary single-level multi-objective global optimization module described herein may construct a set of parameter values X; and for each X the optimization module may evaluate an MS(X) on a set of engine life realizations generated by the engine life simulator. The single-level multi-objective global optimization module may obtain the values of the objectives O1(X), . . . , On(X), and then select the best X based on these objectives. A multi-level or hierarchical MS optimizer described herein may be a process comprising a sequence of single level MS optimizers, the levels going from low to high fidelity. A lower fidelity module may be a faster, coarser, simpler representation of a higher level module. Lower fidelity models may be approximations of high fidelity models, often being reduced order models, surrogate models, or lumped models. For example, a low fidelity MS may lump together subsets of parts and use pre-computed solutions or heuristic rules for faster decisions. Lower fidelity simulators may use larger time steps, coarser probability steps, or smaller numbers of simulation steps. Coarse level optimizers may use larger search steps, a smaller number of optimization parameters, and find approximate solutions indicating the regions of the finer level solutions. The types of optimizers may differ on different levels, for example a coarse level optimizer may be a robust global search, while a fine level optimizer may be a gradient based optimization acting only locally close to the best solutions found by the coarse level optimizer. The multi-objective global optimization process and system and variations thereof described herein may be implemented using any one of a number of exemplary embodiments, some of which will now be discussed. Referring now to FIG. 2, a representation of a top level scheme of the multi-objective global optimization system utilizing the exemplary multi-objective global optimization module and employing the process described herein is shown. An exemplary single level schedule optimizer 10 may comprise an optimization module 12, a simulation and MS module 14 having a maintenance scheduler module 16 and an engine life simulator module 18, and a cost evaluator module 20. The optimization module 12 generally ascertains at least one scheduling parameter X 26 that may optimize a set of Objectives O1(X), . . . , On(X). The decision parameters X may be discrete or continuous such as: given time thresholds for future shop visits, engine operation thresholds, part life thresholds, part upgrade thresholds, or parameters of decision rules used for maintenance, and the like. The set of Objectives O1(X), . . . , On(X) may represent, for example, the cost(s) of part(s), cost(s) of work over a time interval of the life of an engine and associated costs, contract related costs, and the like. The exemplary global optimization module 12 may be implemented by any one of a number of classic optimization techniques such as mixed integer optimization algorithms (e.g., branch and bound) when only one objective is present, e.g., when a single compound objective is chosen as a weighted sum of the n objectives as: O(X)=O1(X)w1+ . . . +On(X)wn. Here w1, . . . , wn are weights that multiply the objectives O1(X), . . . , On(X). A multi-objective optimization may be used when more than one objective is present. Any of the multi-objective optimizations techniques described in literature or implemented in commercial optimization software such as in Matlab or iSIGHT may be used. Deterministic approaches (e.g., mesh searches, graph searches, and the like) or stochastic techniques (Monte Carlo, genetic/evolutionary algorithms, simulated annealing, and the like) may also be utilized. The simulation and MS module 14 may include a maintenance schedule module 16 and an engine life simulator module 18. The maintenance scheduler module 16 is a decision tool that may be used to decide what maintenance work may be performed on a given engine, at a given shop visit. Generally, the maintenance schedule module 16 may be a program that incorporates sets of maintenance rules as known to one of ordinary skill in the art and accepts as parameters the aforementioned scheduling parameters X. The MS may encompass different levels of fidelity and may incorporate optimizations by itself to determine the best maintenance decision for each particular shop visit of a given engine. For example, the MS may incorporate combinatorial or stochastic algorithms in order to carry out the determination(s). The engine life simulator module 18 may simulate sequences of shop visits in the life of a part. Such shop visits may be scheduled by the maintenance scheduler 16 or may be unscheduled, e.g., a shop visit may occur randomly due to different causes with certain probability distributions. Generally, the engine life simulator module 18 may, for example, generate sequences of deterministic and random time events in the life of an engine, take decisions using the maintenance scheduler 16 with at least one parameter X, update engine states as a result of the maintenance performed, generate data for use by the cost evaluator module 20, and the like. In addition, the engine life simulator module 18 may send engine and event data 22 to the maintenance scheduler 16 and receive maintenance decisions and costs 24 from the maintenance scheduler 16. The cost evaluator module 20 may utilize data generated by the engine life simulator module 18 to compute the aforementioned set of objectives O1(X), . . . , On(X) for the given values of at least one parameter X. Such data may include different costs and probabilities of events incurred during a set of life realizations of a given part. Referring now to FIG. 3, the multi-objective global optimization system and process described herein may be exemplified in another embodiment. The exemplary multi-objective optimization system and process may be embodied in a hierarchical MS optimizer 30 having at least one MS optimizer per each level 32, 33, 36, each of which is an MS optimizer 10 as described in FIG. 2. In the hierarchy, as each level progress higher (Level k, . . . , Level 1), the fidelity of the optimization, simulation and decision modules also becomes higher, Level 1 being the highest fidelity level 32. Higher level modules 32 (e.g., Level 1 MS Optimizer) may send information 34 such as engine data, failure probabilities, maintenance data, and the like, to lower level modules 36. The lower level MS optimizers 36 (e.g., Level k Optimizer) may provide information 38 on optimal solutions such as approximate solutions, ranges of solutions, and the like, to the higher level modules. For example, where k=2 and a two level hierarchical optimizer is being utilized, the lower level simulation and MS module 16 may use larger time simulation steps and/or a smaller number of simulations for each given parameter X. In this exemplary embodiment, a level k=2 maintenance scheduler 16 may be a simplified maintenance scheduler, e.g., utilizing fewer look ahead steps, larger steps; using combinations of parts; utilizing heuristic decision rules; or using databases of previous solutions for approximating decisions. Likewise, on level k=2, the maintenance scheduler 16 may also comprise a different optimization technique than the optimization techniques used by the level k=1 MS. Similarly, the two MS optimizers 12 used on levels 1 and 2 may be different. For example, the different optimization techniques may include a combinatorial optimization, an exhaustive search, a stochastic decision optimization, and the like. The lower level MS optimizer, e.g., a Level 2 MS Optimizer 33, may send a set of optimal solutions to the higher level MS optimizer, e.g., Level 1 MS Optimizer. The high level MS optimizer may begin with these solutions and improve upon them by further optimization using higher fidelity models. In this manner, a sequence of levels may be employed in the alternative. The lower level MS optimizers would hierarchically reduce the vast search space of all possible solutions and indicate the regions where the best solutions may be found. The higher level optimizers may then further refine the regions of the good solutions, working locally in the regions discovered by the lower level optimizers. For example, the lower level optimizer may constitute the exemplary multi-objective global optimizer described herein while the higher level optimizer may be a local optimizer such as a gradient based optimizer or a pattern search optimizer. Referring now to FIGS. 4A and 4B, the multi-objective global optimization system and process described herein may be exemplified in yet another embodiment. The exemplary multi-objective optimization system and process may be embodied in a hierarchical schedule optimizer 40 having two levels where the lower level optimizer 42 (e.g., Level 2 Optimizer) may comprise two coupled single level MS optimizers 44, 46 (See FIG. 3A). The hierarchical schedule optimizer 40 of FIG. 3A may comprise any number of MS optimizers on level 2 coupled in any desired input-output structure. The embodiment of optimizer 40 shown in FIG. 4B may be expanded to encompass any number of levels larger or equal to 2. On each level, the hierarchical optimizer may have any number of MS optimizers coupled in any proper input-output structure, such as optimizers 43 and 45, and optimizers 44, 46, 49 and 51. The Level 2 Optimizer 45 may contain a single MS optimizer or may contain two or more MS optimizers as the MS Optimizer 42 of FIG. 4A, where MS optimizer 42 contains optimizers 44 and 46 coupled together. The same structure may be applicable to Level 2 Optimizer 43 of FIG. 4B. For example, if three levels are present, i.e., k=3, then FIG. 4B may illustrate an embodiment where the structure of three optimizers 45 and 44, 46 may be similar to the structure of three optimizers 48 and 44, 46 of FIG. 4A. Likewise, the structure of three optimizers 43 and 49, 51 of FIG. 4B may be similar to the structure of three optimizers 48 and 44, 46 of FIG. 4A. For example, FIG. 4B illustrates that subsets of MS optimizers may be grouped together and coupled by input-outputs on any of the levels, for example, optimizers 44, 46, 49 and 51 may all be coupled, respectively, on level k. Each of the MS optimizers on any of the levels may use a different set of optimization parameters Y that are subsets of the parameters X. For example, when considering FIG. 4A, the optimizers 44 and 46 may use the subsets of parameters X1 and X2, respectively, that are subsets of X of the higher level optimizer 48. The set of optimization parameters X1 and X2 may or may not share some common parameters, e.g., if X={x, y, z}, then one embodiment may have two MS optimizers, 44 and 46, with common variables y such as X1={x, y}, X2={y, z}; or the sets of variables of the 2 MS optimizers may not have common variables, such as X1={x, y}, X2={z}. The single level optimizers 44, 46 may communicate with each other and share optimal solution information with one another. For example, when X1={x, y}, X2={y, z}, the optimizer 44 may execute first, identify an optimal solution {x1,y1}, send the solution y1 to optimizer 46 who may execute to find the solution {y2,z2}, then 46 would send the output Y2 to the optimizer 44 which would continue in the same way. This example may be used to illustrate possible interactions between optimizers. Also, the single level optimizers 44, 46 may operate sequentially, as described in the previous example, or simultaneously, e.g., both MS optimizers may act simultaneously. One example to illustrate the simultaneous operation of the optimizers may have X1={x, y}, X2={z}, i.e., the optimizers 44 and 46 act without direct interactions on level 2. The solution {x, y, z} found on level 2 may be sent to the level 1 MS optimizer which may perform an optimization iteration to generate {x2, Y2, Z2} on level 1, then send these solutions to level 2 where the optimizers 44 and 46 would continue the optimization from {x2, Y2, Z2} and so forth. This is a generic way by which a higher level optimizer may couple lower level optimizers. In this case, the lower level MS optimizers may contain a smaller number of variables. This example may also suggest one reason why a hierarchical MS optimizer may be more efficient than the single level 1 MS optimizer, by reducing an expensive high level 1 optimization work to work performed by lower dimensional optimizers on levels k>1. Referring now to FIG. 5, the single and multi-objective optimization techniques and approaches described herein, for example, an optimization module 60 for use with the multi-objective global optimization process and system described herein, may be executed using at least one search 62 coupled with at least one elimination criteria 64 and at least one constraint 66 as illustrated in FIG. 4. The searches 62 may be deterministic approaches such as grid, mesh, and the like, or may be stochastic approaches such as Monte Carlo, evolutionary algorithms, genetic algorithms, and the like. The constraints 66 may be any one of a number of types as known to one of ordinary skill in the art, for example, bounds of variables, or logical constraints imposed by the rules of the maintenance scheduler. In addition, the constraints 66 may be included as penalties in the objectives or treated as additional objectives to be minimized. Elimination criteria 64 may be constructed while the searches 62 are being executed. Regions of poor solutions may be identified and eliminated from the searches 62, that is, the elimination criteria may act as new constraints that may prevent the searches from evaluating solutions in poor regions. Elimination criteria may then increase the efficiency of the optimization by reducing the search space. This may be especially effective if the elimination of the poor regions is performed on a coarse level using faster simulations and simplified maintenance schedules. Moreover, the low level optimizers may use exhaustive techniques to search the whole domain of solutions and identify the regions of possible good solutions. Good solutions may be sought in these regions by the higher fidelity optimizers. Such hierarchical optimizers may provide both the confidence provided by exhaustive searches and the accuracy of the local optimizers. Other types of optimizations may be used as well and each MS optimizer on each level may use a different type of optimization. FIG. 6 illustrates the Pareto surface of optimal solutions for a case that uses two objectives O1(X) and O2(X) The two objectives may be, for example, the cost of parts and the cost of work cumulated during the life of a given engine. The parameters X of the solutions on the Pareto surface are optimal or best in the sense that there is no other set of parameters Y for which both objectives O1(Y) and O2(Y) are smaller, that is, there is no Y better than X, i.e., such that O1(Y)<O1(X) and O2(Y)<O2(X). The Pareto surface for any number of objectives On(X) may be defined in the same way as for 2 objectives, that is, X is optimal if there is no Y such that (O1(Y), O2(Y), . . . , On(Y))<(O1(X), O2(X), . . . , On(X)). In a hierarchical optimization approach, the poor regions, i.e., regions away from the Pareto surface, may be eliminated hierarchically, and the searches may focus on the good regions near the Pareto surface, as described above in the discussion on elimination criteria. The hierarchical optimizer finds the Pareto surface hierarchically. A broad search may be performed using low fidelity models, for example, the simulator and MS may be simplified and used in a search to approximate the region containing the Pareto surface. This region may be further refined at higher levels using higher fidelity models. This approach suggests an efficient hierarchical multi-objective optimization technique. Referring again to FIG. 5 and the exemplary embodiment illustrated, the effectiveness of the searches may come close to the results of exhaustive searches by providing high confidence that the best global solutions have been found. However, searches utilizing elimination criteria are far more effective than exhaustive searches since the largest part of the poor solutions may be eliminated using the elimination criteria (See FIG. 5). In addition, local optimizations, for example, gradient based or pattern searches, may be added locally during highest level optimizations. Referring now to FIG. 7, the multi-objective global optimization process and system may be utilized for tuning at least one parameter X of a maintenance scheduler. Tuning the maintenance scheduler will assist maintenance personnel to determine whether or not an engine requires a shop visit, whether or not any component(s) of the engine should be repaired, replaced or scrapped, and the like. The multi-objective global optimization system and process described herein may comprise an executable program, that is, a set of instructions, embodied tangibly in a computer readable medium 101 and/or stored in a computer readable storage device 102 and executed using a computer 100 (See FIG. 1). The illustrated computer 100 may represent a server, desktop, laptop, personal digital assistant, and the like, and may be networked using a WLAN, Ethernet, and the like, and may also be part of a distributed computing system as known to one of ordinary skill in the art. The various modules, and all their alternative embodiments, may all be embodied tangibly in the same aforementioned computer readable mediums, storage devices, and the like, as known to one of ordinary skill in the art. Another embodiment generalizes the above description to distributed computing systems, as may be understood by the skilled in the art, i.e., multiple connected computers may store different parts of the data and may run any of the mentioned modules, such as the MS, optimizers, engine life simulators, etc. For example, a multi-objective global optimization system 70 may execute the multi-objective global optimization process described herein to tune the aforementioned at least one parameter X of a maintenance scheduler for a fleet management program. A hierarchical schedule optimizer 72 may comprise any one of the exemplary multi-objective global optimization embodiments described herein, along with any and all permutations as recognized by one of ordinary skill in the art. The hierarchical schedule optimizer receives at least one data from a database 74 such as part history data in general, component and systems data specifically, maintenance data, maintenance history, engine performance parameter(s), failure probabilities for the parts and engine, costs associated with the maintenance of the engine, and the like. The hierarchical schedule optimizer 72 receives the data from the database 74 and by utilizing the exemplary multi-objective global optimization process(es) described herein, for example in FIGS. 2-4B, determines those scheduling parameters X 78 that may optimize the objectives O1(X), . . . , On(X). The scheduling parameters X that are identified become at least one optimal scheduling parameter X 78. An optimized maintenance schedule 76 that uses the optimal X is obtained, and it may be used to determine what maintenance work is required for the given engine at the current and future shop visits. One or more embodiments of the present invention have been described. Nevertheless, it will be understood that various modifications may be made without departing from the spirit and scope of the invention. Accordingly, other embodiments are within the scope of the following claims. |
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claims | 1. A method for fabricating nuclear fuel pellets, the method comprising:preparing mixture powder by mixing additive powder comprising MnO and Al2O3 with UO2 powder;compacting the mixture powder to produce green pellets; andsintering the green pellets at a temperature of 1,600° C. to 1,800° C. under a reducing gas atmosphere,wherein when the mixture powder of the additive powder and the UO2 powder is prepared, the additive powder is mixed with UO2 powder in such a way that the (MnO +Al2O3)/UO2 ratio in the mixture powder is 57.17 μg/g to 4205.70 μg/g, and the MnO/Al2O3 ratio of the additive powder is 0.68 to 68.26 by weight, and wherein the melting point of the additive powder is below 1,800° C. 2. The method of claim 1, wherein, in sintering the green pellets, the reducing gas atmosphere is a hydrogen-containing gas atmosphere. 3. The method of claim 2, wherein the hydrogen-containing gas is hydrogen-containing mixture gas, which contains hydrogen gas and at least one selected from the group consisting of carbon dioxide, water vapor and inert gas. 4. The method of claim 2, wherein the hydrogen-containing gas is hydrogen gas. |
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041475887 | summary | The present invention relates to nuclear reactor recharging devices and, more particularly, to recharging devices for fast-neutron reactor with liquid metal coolant, wherein the replacement of spent fuel assemblies and rods of the control and safety system is carried out under conditions when the reactor is stopped, all the recharging operations are shielded, and the core circuit remains sealed. There is known a nuclear reactor recharging device comprising a separate mechanism for replacing fuel assemblies, and a separate mechanism for replacing rods of the control and safety system. The use of two separate mechanisms is due to substantial differences in the configuration and size of the fuel assemblies and rods of the control and safety system. In this specific case, the rods of the control and safety system are round, whereas the fuel assemblies are hexahedral and have a greater length and diameter. There is also a French device, known as Super Phoenix, for recharging fast reactors. In this device, fuel assemblies and rods of the control and safety system are handled by one mechanism. The rods are placed in cartridges so that both make up an assembly similar to a fuel assembly. The latter technical solution envisages the use of special end members for fuel assemblies, rods of the control and safety system, and cartridges. These end members are hollow cylinder-shaped heads with an internal groove to interact with the grip of the recharging mechanism. The device under review is not applicable to nuclear reactors, wherein the recharging mechanism has a grip to interact with profiled heads of fueld assemblies and rods of the control and safety system, which heads are gripped on the outside. However, there is a large group of nuclear reactors whose recharging mechanisms interact with profiled heads of core elements, which are gripped on the outside, because this type of gripping is the most effective. Besides, at present, use is made of a device for recharging a fast-neutron reactor, comprising a storage drum with compartments for new fuel assemblies and rods of the control and safety system, a mechanism for transferring fuel assemblies and rods from the storage drum to a recharging mechanism having a grip to interact with profiled heads of fuel assemblies and rods of the control and safety system to be gripped on the outside, and place new fuel assemblies and rods in the reactor's core and remove spent fuel assemblies and rods therefrom into a storage drum which receives spent fuel assemblies and rods removed by the recharging mechanism from the reactor's core and transferred by the transfer mechanism. In this recharging device, the compartments for rods of the control and safety system are different in shape from the compartments for fuel assemblies, which is due to differences in the dimensions of the rods and fuel assemblies. If rods of the control and safety system are placed by mistake in compartments for fuel assemblies, these rods are out of reach of the recharging mechanism and cannot be withdrawn from the compartments. In order to avoid accidents and damage of recharging mechanisms, the program control system of the recharging device must be designed to rule out placing of rods of the control and safety system in the compartments intended for fuel assemblies, and vice versa. For this purpose, program control systems for controlling recharging devices of known types envisage interlocking to prevent grips with fuel assemblies from approacing compartments in the storage drum intended for rods of the control and safety system, and vice versa. Besides, the presence of special compartments in the storage drum complicates the design and increases the size of the drum. Obviously, the foregoing design of the recharging device is too complicated; on the other hand, it is not reliable enough and does not rule out accidents in the course of recharging. It is the main object of the present invention to provide a recharging device for a fast-neutron reactor, intended for recharging fuel assemblies and rods of the control and safety system with profiled heads, and having a single mechanism for the transfer of fuel assemblies and rods of the control and safety system. It is another object of the present invention to raise the reliability of the recharging process and reduce the recharging time. The foregoing objects are attained by providing a recharging device for a fast-neutron reactor, comprising a storage drum with compartments for new fuel assemblies and rods of the control and safety system, a mechanism for transferring fuel assemblies and rods of the control and safety system from the storage drum to a recharging mechanism having a grip intended to interact with profiled heads of fuel assemblies and rods of the control and safety system to be gripped on the outside, as well as to place new rods and fuel assemblies in the reactor's core and remove spent ones therefrom and a storage drum for spent fuel assemblies and rods of the control and safety system, removed from the reactor's core and received by this second drum from the transfer mechanism, in which device the compartments of the storage drums for rods of the control and safety system are identical in accordance with the invention, with the compartments for fuel assemblies, the rods of the control and safety system being stored and transported from the storage drums to the recharging mechanism in sleeve-type holders, so that when placed in such holders, the dimensions of the rods of the control and safety system are equal to those of the fuel assemblies, the joining of each sleeve-type holder with a respective rod being effected with the aid of a collet provided on the open end of each holder, on the surface of each rod, near its head, there being provided an encircling groove to interact with the collet, the grip of the recharging mechanism being provided with a stop to interact with the internal surface of the collet to open the latter and release the rod of the control and safety system from the holder under the action of a spring arranged in the holder. The recharging device of this invention makes it possible to replace both fuel assemblies and rods of the control and safety system with the aid of the same mechanisms; the device is simple in design and rules out accidents due to placing rods of the control and safety system into compartments intended for fuel assemblies and vice versa. |
abstract | An infrared projector including: a light source which radiates an infrared ray; and an emitter which emits visible light when exposed to infrared light, wherein the emitter is provided on a position where the infrared ray radiated from the light source reaches. |
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059011939 | description | DESCRIPTION OF THE PRESENT INVENTION The invention is based on the realization that a fuel element with a two-layer cladding, in order to obtain a small amount of precipitated hydrides in the boundary zone between the layers and a tangential hydride distribution in both the inner and outer parts, shall be composed of zirconium alloys whose Sn content does not differ more than 0.7%. Since the cladding is built up in such a way that the difference in Sn content between the inner and outer parts does not exceed 0.7%, both parts have a relatively similar recrystallization temperature. This means that a finishing stress-relieve anneal of the cladding results in the outer layer being stress-relieve-annealed or only partially recrystallized. Sn and O increase the recrystallization temperature at a given degree of processing, whereas other alloying elements such as Fe, Cr, Ni and Nb reduce the recrystallization annealing temperature. By a two-layer cladding, good corrosion properties may be utilized in an alloy for the outer layer together with good mechanical and creep properties of a conventional construction material for a fuel cladding such as Zircaloy-2 and Zircaloy-4, which constitute the inner supporting part. Zircaloy-2 contains 1.2-1.7% Sn, 0.07-0.2% Fe, 0.05-0.15% Cr, 0.03-0.08% Ni, 0.07-0.15% O and the total amount of Fe, Cr and Ni is within the interval 0.18-0.38%. zircaloy-4 contains 1.2-1.7% Sn, 0.18-0.24% Fe, 0.07-0.13% Cr and 0.10-0.16% O, and the total amount of Fe and Cr is within the interval 0.28-0.37%. According to the invention, a good corrosion resistance and a small hydrogen pick-up are obtained for a fuel element with a cladding tube whose outer part consists of a zirconium alloy with addition of Sn within the content interval 0.65 to 0.95% and Fe within the interval 0.35 to 0.5%. The inner part of the cladding may consist of a conventional zirconium alloy such as zircaloy-4 and zircaloy-2 with a normal content of Sn (within the interval 1.2 to 1.7%), but may also consist of another zirconium alloy with sufficiently good mechanical and creep resistance properties to be able to constitute a supporting part of a fuel cladding. Of importance, however, is that during stress-relieve annealing of the inner part at 450 to 510.degree. C., the cladding is given an only partially recrystallized outer part. When estimating the recrystallization temperature, all the alloying elements in the inner layer may be taken into consideration, which inner layer may comprise, in addition to Sn, Fe, Cr, Ni and Nb. The Sn content in the outer part and the inner part shall be relatively equal and not differ more than at most 0.7% in order for a final heat treatment of the cladding at 450 to 510.degree. C. to result in the supporting inner part being stress-relieve-annealed and the outer layer being only partially recrystallized. In this way, it may be ensured that hydrogen which is absorbed by the fuel cladding during operation will be precipitated evenly in the cross section of the cladding and preferably not in the boundary zone between the layers and, in addition, a largely tangential precipitation of the hydrides is obtained. Since the Fe content in the outer layer is relatively high and in partially recrystallization-annealed state, a harder outer layer is obtained, which facilitates the mounting of fuel rods in the fuel bundle. During manufacture of a cladding tube for fuel rods included in a fuel element according to the invention, an inner tube of Zircaloy-4 is joined together with an outer part of zirconium with 0.8% Sn and 0.4% Fe. The inner part of zircaloy-4 is chosen so that the Sn content in the material does not exceed the Sn content in the outer layer by more than at most 0.7%. In both Zircaloy-4 and the outer layer, the other substances in the materials are limited to the maximally allowed values for reactor grade zirconium. The parts are joined together by means of extrusion so as to become metallurgically bonded, whereupon tube manufacture in conventional manner using cold-rolling operations and intermediate heat treatments is performed. The final heat treatment is performed at 450-510.degree. C. for 2-5 hours. Measurement of the hydride orientation shows that both the inner and outer parts are given an f.sub.n -value <0.05. By f.sub.n -value is meant the percentage of hydrides oriented within 45.degree. from a radial direction in relation to the total number of hydrides. During a normal hydride test, comprising intentional hydrogenation of the tube so that this contains at least 100 ppm hydrogen, few hydrides visible in a microscope are present in the outer layer or in the bonding zone between the layers. The outer layer constitutes approximately 10-25% of the wall thickness of the cladding tube. During manufacture of the cladding tube, it is advantageous if the so-called annealing parameter A is high. An aim is that log A should be greater than -13. A is a measure of the sum of all the heat treatments during the tube manufacture and is defined as A=.SIGMA..sub.i t.sub.i .multidot.exp(-Q/RT.sub.i), where t.sub.i =annealing time in hours, T.sub.i =annealing temperature in .degree.K, Q=the activation energy=15000 J/mole, R is the general gas constant. |
description | The present application claims benefit under 35 U.S.C. §119(e) of U.S. Provisional Patent Application No. 61/230,544 filed Jul. 31, 2009 entitled “Detecting Pin Diversion from Light Water Reactor Spent Fuel Assemblies,” the disclosure of which is hereby incorporated by reference in its entirety for all purposes. The United States Government has rights in this invention pursuant to Contract No. DE-AC52-07NA27344 between the United States Department of Energy and Lawrence Livermore National Security, LLC for the operation of Lawrence Livermore National Laboratory. Field of Endeavor The present invention relates to light water reactor spent fuel assemblies, and more particularly Pressurized Water Reactor (PWR) spent fuel assemblies with both UO2 and mixed oxide (MOX) types of fuel and more particularly to determining whether fuel rods (pins) within PWR spent fuel assemblies are missing or have been replaced with dummy fuel rods. State of Technology U.S. Pat. No. 7,514,695 for a detector and method for inspecting a sealed nuclear storage container provides the following state of technology information: “Heretofore, spent nuclear fuel has been placed in fuel storage casks which are typically stored above ground, at various locations in the continental United States. These storage areas are typically in restricted areas, and security is provided to protect the casks from possible tampering or the removal of any spent nuclear fuel. At present, the only indication of possible reactor spent fuel diversion from a storage cask is provided by means of tamper-indicating tags and seals which are provided with each of the storage casks. With the increasing risk of terrorist acts within the United States, and the possibility that spent nuclear fuel sources might by sought after and diverted for possible terrorist acts, a renewed effort has been undertaken to identify means by which spent nuclear fuel storage casks may be readily inspected to determine whether spent nuclear fuel which has been stored within same has been removed improperly from the storage cask.” U.S. Pat. No. 4,389,568 for a method for monitoring irradiated nuclear fuel using Cerenkov radiation provides the following state of technology information: “A method is provided for monitoring irradiated nuclear fuel inventories located in a water-filled storage pond wherein the intensity of the Cerenkov radiation emitted from the water in the vicinity of the nuclear fuel is measured. This intensity is then compared with the expected intensity for nuclear fuel having a corresponding degree of irradiation exposure and time period after removal from a reactor core. Where the nuclear fuel inventory is located in an assembly having fuel pins or rods with intervening voids, the Cerenkov light intensity measurement is taken at selected bright spots corresponding to the water-filled interstices of the assembly in the water storage, the water-filled interstices acting as Cerenkov light channels so as to reduce cross-talk. On-line digital analysis of an analog video signal is possible, or video tapes may be used for later measurement using a video editor and an electrometer. Direct measurement of the Cerenkov radiation intensity also is possible using spot photometers pointed at the assembly.” U.S. Pat. No. 4,335,466 for a method and apparatus for measuring irradiated fuel profiles provides the following state of technology information: “In order to comply with various safeguards agreements, inspection organizations such as NRC (Nuclear Regulatory Commission) and IAEA (International Atomic Energy Agency) need a capability of very quickly and accurately monitoring in a non-destructive manner the fissile content of spent fuel assemblies in storage pools. Presently, measurements of the content of residual and produced fissile material are not directly measured but rather are inferred by measuring particular data which is correlated to burnup (which is a measure of nuclear reactor fuel consumption, expressed either as a percent of fuel atoms that have undergone fission or as the amount of energy produced per unit weight of fuel).” Features and advantages of the present invention will become apparent from the following description. Applicants are providing this description, which includes drawings and examples of specific embodiments, to give a broad representation of the invention. Various changes and modifications within the spirit and scope of the invention will become apparent to those skilled in the art from this description and by practice of the invention. The scope of the invention is not intended to be limited to the particular forms disclosed and the invention covers all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the claims. A technical safeguards challenge has remained for decades for the International Atomic Energy Agency (IAEA) to identify possible diversion of nuclear fuel pins from Light Water Reactor (LWR) spent fuel assemblies. As modern nuclear power plants are pushed to higher power levels and longer fuel cycles, fuel failures (i.e., “leakers”) as well as the corresponding fuel assembly repairs (i.e., “reconstitutions”) are commonplace occurrences within the industry. Fuel vendors have performed hundreds of reconstitutions in the past two decades, thus, an evolved know-how and sophisticated tools exist to disassemble irradiated fuel assemblies and replace damaged pins with dummy stainless steel or other type rods. Various attempts have been made in the past two decades to develop a technology to identify a possible diversion of pins and to determine whether some pins are missing or replaced with dummy or fresh fuel pins. However, to date, there are no safeguards instruments that can detect a possible pin diversion scenario to the requirements of the IAEA. The FORK detector system can characterize spent fuel assemblies using operator declared data, but it is not sensitive enough to detect less than 50% missing pins from spent fuel assemblies. In addition, it requires the fuel to be lifted out of the storage location to perform the measurement. Likewise, an emission computed tomography system has been used to try to detect missing pins from a spent fuel assembly and has shown some potential for identifying possible missing pins but this capability has not yet been fully demonstrated. The use of such a device in the future would not be envisaged, especially in an inexpensive, easy to handle setting for field applications. The Cerenkov viewing device is another instrument used in the field. However, it does not work well for fuel that has been cooled for a long time or low burnup fuel or in murky spent fuel pool conditions. It also has an issue with cases where the pins are missing in a random fashion since neighboring fuel pins of a missing one give off radiation potentially giving a false negative indication. The present invention provides a system for determining whether some fuel pins within PWR spent fuel assemblies are missing or replaced with dummy fuel pins. The methodology can detect as low as 10% missing fuel without relying on operator provided data or any movement of fuel from its storage location. The method is also extendable to MOX fresh fuel assemblies or MOX spent fuel assemblies. The apparatus is in the form of a cluster which contains multiple neutron detectors and/or gamma detectors. The apparatus is inserted into spent fuel assemblies through the guide tube holes present in the spent fuel assemblies. The radiation responses, gamma and neutron, of the detectors are simultaneously measured at a location or multiple locations within the guide tubes. The present invention provides methods and apparatus for detecting diversion of spent fuel from PWRs. One embodiment provides a method of determining possible diversion of pins in a PWR spent fuel assembly having guide tube holes, including the steps of providing a detector cluster containing gamma ray detectors, inserting the detector cluster containing the gamma ray detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray radiation responses of the gamma ray detectors in the guide tube holes, processing the gamma ray radiation responses at the guide tube locations by normalizing them to the maximum value among them and producing a signature based on these normalized values, and producing an output that consists of this signature that can indicate possible diversion of the pins from the spent fuel assembly. Another embodiment provides a method of determining possible diversion of pins in a PWR spent fuel assembly having guide tube holes, including the steps of providing a detector cluster containing neutron detectors, inserting the detector cluster containing the neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring neutron radiation responses of the neutron detectors in the guide tube holes, processing the neutron radiation responses at the guide tube locations by normalizing them to the maximum value among them and producing a signature based on these normalized values, and producing an output that consists of this signature that can indicate possible diversion of the pins from the spent fuel assembly. Another embodiment provides a method of determining possible diversion of pins in a PWR spent fuel assembly having guide tube holes, including the steps of providing a detector cluster containing neutron detectors and gamma ray detectors, inserting the detector cluster containing the neutron detectors and the gamma ray detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring neutron radiation responses and gamma ray radiation responses of the neutron detectors and the gamma ray detectors, processing the neutron radiation responses and the gamma ray radiation responses of the neutron detectors and the gamma ray detectors and determining whether pins are missing or have been replaced with dummy or fresh pins, and providing an output indicating possible diversion of the pins in the spent fuel assembly. Another embodiment provides an apparatus for determining whether some fuel pins within PWR spent fuel assemblies are missing or replaced with dummy fuel rods, wherein the spent fuel assemblies have guide tube holes, including a cluster which contains neutron detectors or gamma detectors or neutron detectors and gamma detectors, the cluster inserted into the spent fuel assemblies through the guide tube holes in the spent fuel assemblies; and a measuring and analyzing device, the measuring and analyzing device measuring radiation responses of the detectors simultaneously at a location or multiple locations within the guide tube holes and processing the radiation responses and determining whether pins are missing or have been replaced with dummy or fresh pins. The invention is susceptible to modifications and alternative forms. Specific embodiments are shown by way of example. It is to be understood that the invention is not limited to the particular forms disclosed. The invention covers all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the claims. Referring to the drawings, to the following detailed description, and to incorporated materials, detailed information about the invention is provided including the description of specific embodiments. The detailed description serves to explain the principles of the invention. The invention is susceptible to modifications and alternative forms. The invention is not limited to the particular forms disclosed. The invention covers all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the claims. The present invention provides a system for detecting diversion of spent fuel from Pressurized Water Reactors (PWR) to address a long unsolved safeguards verification problem for international safeguards community such as the International Atomic Energy Agency (IAEA) or the European Atomic Energy Community (EURATOM). The present invention involves inserting tiny neutron and gamma detectors into the guide tubes of a spent fuel assembly and measuring the signals. The guide tubes form a quadrant symmetric pattern in the various PWR fuel product lines and the neutron and gamma signals from these various locations are processed to obtain a unique signature for an undisturbed fuel assembly. Signatures based on the neutron and gamma signals individually or in a combination can be developed. Removal of fuel pins from the assembly will cause the signatures to be visibly perturbed thus enabling the detection of diversion. All of the required signal processing to obtain signatures can be performed on standard equipment such as laptop computers. Monte Carlo simulation studies were performed to develop the present invention and a set of controlled experiments with actual commercial PWR spent fuel assemblies were performed to validate the invention. Based on the simulation studies and benchmarking measurements, the methodology developed promises to be a powerful and practical way to detect partial defects that constitute 10% or more of the total active fuel pins. This far exceeds the detection threshold of 50% missing pins from a spent fuel assembly, a threshold defined by the IAEA Safeguards Criteria. The methodology does not rely on any operator provided data like burnup or cooling time and does not require movement of the fuel assembly from the storage rack in the spent fuel pool. General Introduction to Source Terms Low enriched fuel such as that used in PWRs emit neutrons and photons (gamma rays) that are dependent on the amount of energy produced by this fuel before it is discharged from the core of the reactor and stored as spent fuel in spent fuel pools or other repositories. The amount of energy produced is referred to as burnup and is typically presented in units of MWd/kg or equivalently, GWd/t. The spent fuel emits neutrons and gammas at levels that depend on the burnup. In the case of neutrons, the number emitted per second from typical PWR spent fuel increases as the third to fourth power of the burnup. For gammas, the source is approximately linear with burnup. The neutrons in spent fuel come mainly from the spontaneous fission of actinides such as 244Cm, 242Cm, and to a lesser extent, the even mass number isotopes of Pu. The production of these actinides increases with burnup. To achieve the same burnup, lower enriched fuel needs to operate at a higher neutron fluence than higher enriched fuel while in the core. Therefore, initial enrichment affects the neutron source. In addition, the cooling time (or the total time since the fuel was discharged from the core) is also a parameter that influences the neutron source. This is because of the relatively short half lives of the two dominant neutron sources, 244Cm and 242Cm at 18.1y and 0.45y, respectively. Since 242Cm decays quickly, the principal source for neutrons is 244Cm. A much smaller number of neutrons are produced by (α, n) neutrons produced in the oxide form of the fuel that is typically used in PWRs. The gamma source is linear after 1 year of cooling time and principally depends on the amount of 137Cs. Since the fission yield of this isotope is approximately the same for both 235U and 239Pu, the gamma yield is not very sensitive to initial enrichment. 137Cs has a long half life of 30.2 years and as a result the cooling time is also not much of a factor in the gamma source strength. PDET Concept Pressurized water reactors (PWRs) constitute about 60% of operating reactors in the world. A large number of these have fixed designs with a fuel pin matrix interspersed with guide tubes in a regular quadrant symmetric pattern that are used to insert control rods etc. The drawing figures show a fuel assembly with fuel pins and guide tubes. In the spent fuel pool these guide tubes are generally filled with water and present good locations to insert tiny detectors to measure the population of the neutrons and gammas. The neutrons emitted from the fuel pins are at high energies and migrate and slow down (thermalize) in the water in the pool. The thermal neutrons, and to a lesser extent, fast neutrons go on to produce more fission neutrons by interacting with the fissile and fissionable materials inside the spent fuel. This subcritical multiplication within the spent fuel tends to remove some of the severe dependence of the neutron source on burnup. In addition, neighboring assemblies, depending on burnup gradients and cooling times translating into source gradients, would contribute more neutrons potentially further lessening the burnup dependence of the neutron field. The guide tube locations at the center of an assembly see higher neutron signals since there are contributions from more fuel pins at these locations within the assembly than there are at the guide tube locations in the periphery. The gamma signal is more localized because the gammas are stopped by the high density fuel containing material with high atomic numbers. Gamma signals are also higher at the central locations of guide tubes than at the peripheral locations but the difference much less pronounced than for the neutrons due to the localized nature of the gammas. Contributions to the gamma signal from neighboring assemblies are also small. Ratios of the two signals would cancel out more of the dependence on burnup and would primarily depend on the location of the guide tubes. Thus this signature can be expected to provide insight into the radiation fields inside the assembly with less dependence on the various parameters that vary, such as burnup levels, cooling times, etc. Referring now to the drawings and in particular to FIG. 1, a spent fuel storage pool is illustrated. FIG. 1 shows a partial view of a spent fuel storage pool 100. Three spent fuel assemblies 102 are shown in place at the bottom of the pool 100. The spent fuel assemblies 102 consist of three main elements, a top plate 104 and a bottom plate 106 and a fuel pin array 108 retained by top plate 104 and bottom plate 106. The storage pool 100 has a pool liner 110 and the pool is filled with water 112. The spent fuel assemblies 102 are stored at the bottom of the pool and as indicated by dimension labeled “a.” The assemblies typically are at least 20 feet beneath the surface of the water in the storage pool. Referring now to FIG. 2, the spent fuel storage pool 100 is illustrated with the addition of an insertion fixture 200. FIG. 2 shows the same elements as shown in FIG. 1 with the addition of the insertion fixture 200. The insertion fixture 200 will be shown and explained in greater detail in FIG. 3. The insertion fixture 200 is positioned over the spent fuel assembly 102 that is to be tested by a crane (not shown) that can accurately position the insertion fixture 200 over the desired spent fuel assembly 102. The insertion fixture can then be lowered to engage locating pins into a receptacle on the top plate 104. This will be shown and explained in greater detail in FIG. 4. FIG. 3 is an illustration of the insertion fixture 200. The fixture 200 is comprised of the following elements. There is a bottom plate 208 with an array of guide bushings 206 that will guide the detector rods 204 into the spent fuel array. Protruding from the bottom of plate 208 are guide pins 202. The guide 202 will engage alignment holes in the top plate 104 of spent fuel assembly 102 extending upward from the plate 208 are four slider rods 210. The slider assembly 212 is mounted on the four slider rods 210 and is free to move up and down as indicated by arrow 214. The detector rods 204 are part of the slider assembly 212 and as the slider assembly 212 is lowered the detector rods 204 will slide thru guide bushings 206 and enter guide tubes (control rod tubes) in the spent fuel pin array 108. FIG. 4 is a partial cross sectional view 400 of the insertion fixture 200 in place atop a spent fuel assembly 102. The spent fuel pin array 108 cross section is taken from the fuel lattice illustrated in FIG. 6. The section is taken along section Line A-A. The detector insertion tubes 204 will house the detectors which are not shown here. The detectors can be neutron or gamma detectors or a combination of both types of detectors. FIG. 5 is again a partial cross sectional view 500 of the insertion fixture 200 in place atop a spent fuel assembly 102. In this FIG. the slider assembly 212 has been moved downward thus moving the insertion tube 204 and the detectors contained within the tubes 204. The insertion tubes 204 are shown having entered the guide tubes 404. The slider assembly 212 can position the insertion tubes 204 with their detectors to different depths in the spent fuel array 108 to make multiple measurements. FIG. 6 is an illustration of fuel lattice with guide tubes 602 locations. The fuel lattice shown here in FIG. 6 is a 14×14 array, with sixteen guide tubes 402 with the guide 402 tubes 602 laid out in a symmetrical pattern. FIG. 7 is a view of a 14×14 spent fuel array 108. In FIG. 7 the cylindrical guide tubes (shown as ‘a’, ‘b’, ‘c’ etc.) 402 represent positions that have insertion tubes 204 with neutron or gamma or both types of detectors. The insertion fixture 200 can be repositioned atop the spent fuel array 108 by turning the insertion fixture 200 180°. In FIG. 7 the guide tubes 402 are shown with letters a thru p. The letters a thru p correspond to detectors that will be inserted into the guides to make measurement. FIGS. 8A, 8B, and 8C are schematic representations of three possible arrangements of detectors and the associated apparatus used in collecting and analyzing the data obtained by using this invention. FIG. 8A is a schematic 800A with neutron detectors 802. The signal from detectors 802 feeds to a preamp 804 then to a multi-channel analyzer 806, then to a computer 808 for data output. FIG. 8B is a schematic 800B with gamma detectors 810. The signal from detectors 810 feed to pre amp 804 then to the multi channel analyzer 806 and on to the computer 808 for data output. FIG. 8C is a schematic 800C with both neutron 802 and gamma 810 detectors. The signal again goes to the pre amp 804 and to analyzer 806 and to computer 808 for data output. The PDET tool uses the neutron (typically thermal neutrons), gamma, and ratio of gamma-to-neutron (typically thermal) signals from the guide tubes and normalizes them to the maximum among each set to form unique signatures that are primarily dependent on the geometric arrangement of the measurement locations (guide tubes) compared to other parameters. Spent fuel pools contain about 2000 ppm of boron dissolved in the water and assemblies are typically separated by borated aluminum slats in the storage racks for criticality safety reasons. Studies have shown that the boron content in the pool tilts the signature to some extent while still maintaining the unique geometry dependent shape. Contributions from neighboring assemblies depending on burnup gradients between them and the assembly where the measurement is being made (test assembly) also tend to tilt the base signature without altering the base shape of the signature. A set of base shapes for the normalized neutron, gamma, and gamma-to-neutron ratios are shown in FIG. 9 and constitute signatures that would be obtained by PDET when an assembly is intact. Removal of fuel pins from an assembly constitutes a partial defect and detection of partial defects is an important aspect of International Safeguards Criteria to prevent clandestine removal of fissile material. Studies on the PDET methodology indicate that as low as 10% missing fuel from an assembly can be visually detected by examining the distortion of the base signature. This far exceeds the detection threshold of 50% or more set by the IAEA Safeguards Criteria. The methodology also does not require any operator provided information and does not require the movement of fuel from its storage location, i.e. measurements are made in an in-situ condition. The base signature is maintained in an intact assembly with very little sensitivity to burnup, cooling time or initial enrichment. The concept envisions use of standard detectors already in use at the IAEA and a design that is portable for field applications. These features of PDET are all unique and overcome limitation and disadvantages presented by currently used devices such as the Fork detector or the Cerenkov Viewing Device. Thus, this device that would fill an important need in the safeguards area for partial defect detection, a technology that the IAEA has been seeking for the past few decades. Normalized Gamma-to-Thermal Neutron Ratio The normalized signature presented by the ratio of the gamma-to-thermal neutron, the primary signature of interest, was examined extensively in various simulation studies. The signature itself is established by plotting the normalized values of the ratio starting with an inner guide tube location and moving in a counter clockwise manner e.g. ‘c’, ‘d’, ‘a’, ‘b’, etc. This produces a smoothly varying pattern with peaks at the peripheral locations and valleys at the central locations. This is because the gamma and neutron populations are the lowest at the periphery though owing to the greater variation between peak and valley in the neutron signal, the ratio is a mirror image of the individual normalized signals (See FIG. 9). Multi-assembly configurations with both uniform burnup assemblies as well as assemblies with large intra-assembly burnup profiles were simulated. Additional sensitivity studies with variable boron concentrations, cooling times and initial enrichments were also examined. As a result of these studies it was possible to establish an upper and lower bound of the base signature of the ratio. This variation was about ±0.1 about the mean with the constraint that the peak can be a maximum of 1.0. FIG. 10 shows the base signature for the ratio with the upper and lower bounds (dotted lines). FIG. 11 shows a 3×3 array of assemblies with the central assembly among the nine being the test assembly, i.e. where the measurements would be made. These assemblies have a significant burnup gradient from one corner to the diagonally opposite one-27 MWd/kg to 37 MWD/kg. FIGS. 12 A & 12B show two scenarios of the test assembly where 22 pins have been removed. This represents 12% of the total number of fuel pins in the assembly. In one of the cases the cluster of missing pins is in the high burnup region and in the other it is in the low burnup region. FIG. 13 shows the variation of the ratio signature as a result of these two diversion scenarios clearly showing the deviation from the base case of an intact assembly. FIG. 13 is a graph showing missing pin signatures in the test assembly. The gamma signal drops in the vicinity of the missing pins mainly due to the loss of the local source of gammas. The drop in the gamma signal contributes to the drop in the relative ratio to a large extent. There is an increase in the thermal neutron population at the locations in the low burnup region in the vicinity of missing pins because of the migration of neutrons from the high burnup regions that are intact as well as the lack of fuel to absorb the thermal neutrons. For the case of missing pins in the high burnup region, there is a slight increase in the locations surrounded by a few missing pins (e.g., location ‘a’ with six pins missing) for reasons just discussed. The impact on the neutron population is smaller in the high burnup regions with missing pins unless a very large number of the high burnup pins is missing. The deviation of the perturbed signature from the base signature is attributable in addition to the drop in the gamma signal, also to an increase in the neutron signal in the vicinity of the missing pins. The surrounding assemblies have a larger influence on the change in the magnitude of the neutron signal than they do on the more localized gamma signal. A combination of the drop in the gamma signal combined with increases in the thermal neutron signal makes the relative ratio in the signature drop, leading to an overall change of shape in the signature that can be visually detected. The simulations were benchmarked against measurements at the Korea Atomic Energy Institute's spent fuel pool that contained assemblies with missing fuel. One of the assemblies where measurements were made was one which had missing fuel dispersed all over the assembly. FIG. 14 shows the missing fuel in this assembly that had a burnup gradient of 20 MWd/kg in one corner compared to 40 MWd/kg at the diagonally opposite corner. FIG. 14 is an illustration showing an assembly with dispersed missing fuel. The normalized ratio signature of this assembly is shown in FIG. 15. FIG. 15 is a graph showing signatures from the case with dispersed missing fuel. This figure includes the base signature as well as the measured and simulated signatures for this assembly with partial defects. The guide tube designations in this plot correspond to that shown in FIG. 14, i.e. J5, 13 etc. The deviation of the signature from the base case is clearly seen in FIG. 15. The measurement and simulation agree well as can been seen in FIG. 15. This case represented 25 missing pins (22 empty slots and 3 slots with dummy pins represented by stars in FIG. 14) amounting to about 14% diversion from the assembly. Normalized Neutron Signature In addition to the ratio, the normalized neutron signature can also indicate pin diversion. The normalized neutron signature for the case with missing fuel in the low burnup region as shown in FIG. 12 is shown in FIG. 16. FIG. 16 is a graph showing the neutron signature for 12% missing fuel. FIG. 16 clearly shows that the thermal neutron flux has increased in the vicinity of the cluster of empty slots (see FIGS. 12 A & 12B) as a result of increased thermalization in the water. FIG. 16 also shows a tilt in the base signature caused by the effect of the surrounding assemblies that feed neutrons into the low burnup regions thus increasing the population. The neutron signature for the case corresponding to FIG. 14 is shown in FIG. 17. FIG. 17 is a graph showing for dispersed missing fuel neutron signature. Once again deviations are seen in the signature when compared to the base case. Normalized Gamma Signatures Since the signatures are symmetric in nature primarily based on the layout of the guide tubes, the issue of symmetric pin diversion needs to be examined. The gamma signature based on the more localized nature of the gamma signal is particularly useful for symmetric pin diversion detection. FIG. 18 shows the case of an isolated assembly with a burnup distribution that varies from 27-37 MWd/kg with a symmetric set of 28 pins removed around the center. This represents about 16% pin diversion from this assembly. FIG. 19 shows the normalized gamma signature compared to the base signature. FIG. 19 is a graph showing a comparison of signatures with 28 missing pins in a symmetric pattern. The gammas signal is localized since they get absorbed in the high Z, high density fuel and do not travel very far. However, given the pattern of missing fuel in FIG. 18, relatively larger numbers are able to reach the periphery increasing the gamma signals there. Thus, the normally depressed relative signal at the peripheral locations is now higher and that at the central locations is depressed leading to the signature being almost a mirror image of the base signature. The diversion can be seen clearly from examining the perturbed signature compared to the base (dotted line). The neutron signature in this case did not exhibit any change in shape of the signature from the base since the pattern produces a symmetric thermalizing medium for the migrating neutrons keeping the relative population unchanged from the base. Symmetric pin diversion in a uniform burnup assembly (32 MWd/kg) surrounded by a checkerboard pattern of high (37 MWD/kg) and low (27 MWd/kg) burnups is shown in FIG. 20. FIG. 20 shows twenty missing pins in a 5×5 array. The diversion here consists of 20 missing pins (11% of total) in clusters of five symmetric about each set of 4 guide tubes. FIG. 21 shows the gamma signature from this symmetric case and it is clear that the perturbed signature is very different. FIG. 21 is a graph showing gamma signature for 20 missing pins. In this case, the corner and center locations in each cluster of four guide tubes have three adjacent pins missing. This makes the gamma signal lower than usual at the center locations. For the corner locations that normally see the lowest gamma signals the effect is magnified. Once again because of the symmetric diversion, the neutron signal does not show any deviation from the base signature. The ratio plot is similar to the gamma plot. The present invention provides methods and apparatus for detecting diversion of spent fuel from Pressurized Water Reactors (PWR). One embodiment provides a method of determining possible diversion of pins in a PWR spent fuel assembly having guide tube holes, including the steps of providing a detector cluster containing gamma ray detectors, inserting the detector cluster containing the gamma ray detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring gamma ray radiation responses of the gamma ray detectors in the guide tube holes, processing the gamma ray radiation responses at the guide tube locations by normalizing them to the maximum value among them and producing a signature based on these normalized values, and producing an output that consists of this signature that can indicate possible diversion of the pins from the spent fuel assembly. Another embodiment provides a method of determining possible diversion of pins in a PWR spent fuel assembly having guide tube holes, including the steps of providing a detector cluster containing neutron detectors, inserting the detector cluster containing the neutron detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring neutron radiation responses of the neutron detectors in the guide tube holes, processing the neutron radiation responses at the guide tube locations by normalizing them to the maximum value among them and producing a signature based on these normalized values, and producing an output that consists of this signature that can indicate possible diversion of the pins from the spent fuel assembly. Another embodiment provides a method of determining possible diversion of pins in a PWR spent fuel assembly having guide tube holes, including the steps of providing a detector cluster containing neutron detectors and gamma ray detectors, inserting the detector duster containing the neutron detectors and the gamma ray detectors into the spent fuel assembly through the guide tube holes in the spent fuel assembly, measuring neutron radiation responses and gamma ray radiation responses of the neutron detectors and the gamma ray detectors, processing the neutron radiation responses and the gamma ray radiation responses of the neutron detectors and the gamma ray detectors and determining whether pins are missing or have been replaced with dummy or fresh pins, and providing an output indicating possible diversion of the pins in the spent fuel assembly. Another embodiment provides an apparatus for determining whether some fuel pins within PWR spent fuel assemblies are missing or replaced with dummy fuel rods, wherein the spent fuel assemblies have guide tube holes, including a cluster which contains neutron detectors or gamma detectors or neutron detectors and gamma detectors, the cluster inserted into the spent fuel assemblies through the guide tube holes in the spent fuel assemblies; and a measuring and analyzing device, the measuring and analyzing device measuring radiation responses of the detectors simultaneously at a location or multiple locations within the guide tube holes and processing the radiation responses and determining whether pins are missing or have been replaced with dummy or fresh pins. Additional details of the present invention are described in the articles identified below. The article, “Monte Carlo Characterization of Pressurized Water Reactor Spent Fuel Assembly for the Development of a New Instrument for Pin Diversion”, by Y. S. Ham, et al, Jun. 19, 2006, INMM06, Nashville, Tenn., Jul. 16, 2006 through Jul. 20, 2006. The article, “Development of a Safeguards Verification Method and Instrument to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies”, by Y. S. Ham, Oct. 16, 2006, Symposium on International Safeguards, Vienna, Austria, Oct. 16, 2006 through Oct. 20, 2006. The article, “Characterization of a Safeguards Verification Methodology to Detect Pin Diversion from Pressurized Water Reactor (PWR) Spent Fuel Assemblies using Monte Carlo Techniques,” by S. Sitaraman and Y. S. Ham, 48th Annual Meeting of the Institute of Nuclear Materials Management, Tucson, Ariz., July 2007. The article, “Sensitivity Studies for an In-situ Partial Defect Detector (PDET) in Spent Fuel Using Monte Carlo Techniques”, by S. Sitaraman and Y. S. Ham, INMM-49th Annual Meeting, Nashville, Tenn., Jul. 13, 2008 through Jul. 17, 2008. The article, “Symmetric Pin Diversion Detection using a Partial Defect Detector (PDET), by Shivakumar Sitaraman and Young S. Ham, INMM-50th Annual Meeting, Tucson, Ariz., Jul. 12, 2009 through Jul. 16, 2009. The article, “Y. S. Ham, S. Sitaraman, H. Shin, S. Eom, and H. Kim, “Experimental Validation of the Methodology for Partial Defect Verification in Pressurized Water Reactor Spent Fuel Assemblies”, 50th Annual Meeting of the Institute of Nuclear Materials Management, Tucson, Ariz., July 2009. The articles identified above are incorporated in this patent application in their entirety for all purposes by this reference. While the invention may be susceptible to various modifications and alternative forms, specific embodiments have been shown by way of example in the drawings and have been described in detail herein. However, it should be understood that the invention is not intended to be limited to the particular forms disclosed. Rather, the invention is to cover all modifications, equivalents, and alternatives falling within the spirit and scope of the invention as defined by the following appended claims. |
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055966152 | abstract | A fuel assembly avoiding the generation of irradiation damage, a Zr alloy used for the same, and a manufacturing method thereof. According to one embodiment, a super-saturated solid-solution Zr alloy powder having a crystal grain size in the range of 1000 nm or less and containing Fe, Ni and Cr is prepared by mechanical alloying, and the alloy powder is subjected to HIP, hot-working, cold-working and final heat-treatment. |
claims | 1. An electron beam column for automated inspection of manufactured substrates, the electron beam column comprising:an electron gun including a source for emitting electrons and a gun lens for focusing the electrons into an electron beam;an objective lens for focusing the electron beam onto a beam spot on a surface of a target substrate;a continuously-variable aperture configured to select a beam current;a ground plate below the objective lens;a charge-control plate above the surface of the target substrate; anda pre-charge control plate between the ground plate and the charge-control plate. 2. The electron beam column of claim 1, wherein the continuously-variable aperture is formed by overlapping blades. 3. The electron beam column of claim 2, wherein the overlapping blades form a square aperture. 4. The electron beam column of claim 3, wherein V-cut edges of two blades form the square aperture. 5. The electron beam column of claim 3, wherein edges of four blades form the square aperture. 6. The electron beam column of claim 1, further comprising:a gate valve configured between the electron gun and the continuously-variable aperture. 7. A method of using an electron beam column for automated inspection of manufactured substrates, the method comprising:emitting electrons from a source;forming an electron beam using a gun lens to focus the electrons;selecting a beam current using a continuously-variable aperture;focusing the electron beam onto a beam spot on a surface of a target substrate using an objective lens;applying a voltage to a pre-charge control plate between a ground plate below the objective lens and a charge control plate above the target substrate;scanning the beam spot over the surface of the target substrate; anddetecting secondary electrons from the beam spot using a detector. 8. The method of claim 7, further comprising:adjusting an aperture size of the continuously-variable aperture by changing an amount of overlap between overlapping blades. 9. The method of claim 8, wherein the overlapping blades form a square aperture. |
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047708433 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to a method and apparatus for accurately determining on-line, and controlling, the stability of the individual fuel assemblies in a boiling water reactor. More particularly, it relates to selecting the fuel assemblies most likely to exhibit instability and performing a stability analysis based on physical solution of the non-linear conservation equations which takes into account nuclear feedback as well as hydraulic effects on the individual fuel assemblies with the effects of cross-coupling included when appropriate. The invention further relates to determining control action required to return a core with unstable fuel assemblies to stable operation, thus providing an effective on-line expert system in real time. 2. Background Information Boiling flow instabilities must be considered in the design and analysis of many devices used in energy production. In particular, such instabilities should be avoided in most apparatus of interest. Sufficiently large excursions and/or oscillations from the steady state can affect the efficiency of the process, erode thermal margins, and may cause physical damage to mechanical components. Flow instabilities are of particular concern in boiling water reactor (BWR) cores. BWR plant operators are under strict Nuclear Regulatory Commission guidelines to be alert for, and to suppress, any flow/nuclear instability. Monitoring is typically done by observing neutron flux signals from local power range monitors (LPRMs), and simultaneously adjusting power/flow conditions to remain below a prespecified core stability limit. In the event that an oscillation is noticed in the LPRM signals, the operator inserts control rods or increases total core flow to attempt to suppress instabilities. Additional control action is then required to bring the plant into a desired configuration. The LPRMs in a BWR are arranged in strings distributed across the reactor core, and a typical BWR may have 160 such detectors. While this means that the operator must monitor a large numer of readings, there are still many fuel assemblies which are not adjacent to an LRPM. This can lead to a safety problem in the form of fuel damage, and release of fission products could also occur, should a fuel assembly some distance away from an LPRM string undergo sustained regional instabilities. Such instabilities may not be detected by the LPRM signals, and hence go unnoticed. Thus, the practice of controlling instabilities through monitoring LPRM signals is cumbersome at best, and could lead to increased downtime and hence economic loss. U.S. Pat. No. 4,319,959 suggests a system for supervising stability in a BWR in which flux readings from the LPRMs, and signals regarding such operating conditions as rod position, the flow quantity of recirculation water and the thermal power of the core determined by the heat balance of the plant, are used to determine values for the coolant flow quantity and thermal power in each assembly. Channel stability of each assembly is then determined using an equation which takes into account hydrodynamic factors such as inlet flow velocity, inlet subcooling, heat flux, and mean pressure. However, axial distribution of power, cross flow between channels of subchannels, and nuclear feedback effects are not considered. In one embodiment, a correlation of signals from LPRMs spaced along a channel is used to determine flow rate. Indications of channel stability can be used as a guide for adjusting control rod position to improve stability. Such an approach uses lump parameters for assemblies adjacent LPRMs and can not provide meaningful indications of stability in assemblies remote from an LPRM. It also requires calculation of stability for all of the assemblies to locate any that might be unstable. It is a primary object of the present invention to provide a method and apparatus for controlling the stability of the fuel assemblies in a boiling water reactor which provides an accurate quantitative evaluation of stability based upon physical laws in real-time for on-line implementation. It is a another object of the invention to achieve the first object by selecting for the accurate quantitative evaluation of stability a limited number of fuel assemblies most susceptible to instability based upon current observed operating conditions of the core. It is yet another object of the invention to provide such a method and apparatus which generates and uses, in said selection and evaluation, distributed values of pertinent core parameters. It is still another object of the invention to provide a method and apparatus which evaluates stability taking into account the nuclear feedback as well as hydrodynamic effects on the stability of the individual fuel assemblies. Finally it is an object of the invention to provide an expert system environment to the plant operator which suggests ways to obviate unstable operation and also to safely recover from unstable operating modes. SUMMARY OF THE INVENTION These and other objects are realized by the invention in which the stability of the fuel assemblies is controlled by measuring on an on-line basis reactor parameters including flow, temperature, control rod position and pressure. A digital computer utilizes these measurements to generate nodal distributions of selected reactor parameters with at least one node per fuel assembly in the radial plane and a plurality of nodes in the axial direction. The computer then calculates from these nodal distributions a stability index for selected fuel assemblies taking into account core physical parameters which are a measure of power level, axial power distribution, flow, enthalpy, void drift, detailed fuel rod dynamics, nuclear reactivity feedback, and where appropriate, flow cross-coupling in the radial and axial directions. The stability indexes, which in the preferred form of the invention are decay ratios, are reported to the operator. The stability of the least stable fuel assembly is, in addition, compared to a prdetermined stability index. If this stability index limit is exceeded by the least stable fuel assembly, the computer iteratively assumes incremental changes in either control rod position or coolant flow, as selected by the operator, and recalculates the stability index until the index is within the prescribed limit. While the calculations are detailed in taking into account the nuclear feedback as well as hydrodynamic effects on stability, they are carried out rapidly to provide recommendations for real-time adjustments in flow or control rod position. The cumulative adjustment needed to return the least stable assembly to stable operation is reported to the operator, who may make the recommended change in flow or rod position manually at his discretion. Alternatively, the recommended change in flow or rod position can be provided as a control signal to a controller which automatically makes the required parameter change. As another aspect of the invention, the number of fuel assemblies for which stability calculations need to be generated by the digital computer is reduced to those most susceptible to instability based upon current operating conditions. To this end, the power generated by each assembly, and the axial distribution of power, including the axial location of the peak power in each assembly, are determined. Only those assemblies generating power above a certain level, preferably above the average level of power generated in all the assemblies, and having their peak axial power occur at a location below the average location for all fuel assemblies are selected for stability calculations. This reduces the number of assemblies for which the stability calculations must be made from several hundred to about a dozen or less. With such a manageable number of fuel assemblies, the detailed calculations can be carried out in time for real-time control of fuel assembly stability. The invention relates to both the above method and apparatus for controlling stability in a boiling water reactor. |
description | The present application is a divisional of and claims the benefit of U.S. patent application Ser. No. 11/692,952, filed Mar. 29, 2007, entitled METHOD OF APPLYING A BURNABLE POISON ONTO THE EXTERIOR OF NUCLEAR FUEL ROD CLADDING, the disclosure of which is incorporated herein by reference. 1. Field of the Invention This invention relates generally to nuclear fuel rods. More specifically, this invention relates to a method of applying a burnable poison onto the exterior of a nuclear fuel rod, which burnable poison will adhere and be effective even after contact with coolant water. 2. Description of the Related Art Burnable poisons, which are materials that have a high neutron absorption cross-section that gradually burn-up under neutron irradiation, are typically utilized in nuclear reactors to control excess reactivity in the nuclear fuel without having to employ one or more control rods. Burnable poisons are currently incorporated into the fuel of a nuclear reactor. Moreover, due to the burn-up of the burnable poison, the negative reactivity of the burnable poison decreases over core life. Examples of patents in this area include U.S. Pat. Nos. 3,520,958; 4,774,051, 5,075,075, and 5,337,337 (Versteeg et al.; Peeks et al.; Kopel and Aoyama et al.; respectively) and U.S. Patent Publication No. U.S. 2006/0109946 A1 (Lahoda et al.). Various needs are met by various embodiments of this invention which provide an application device positioned adjacent to a surface of a nuclear fuel rod. The application device is used to spray the nuclear fuel rod with a variety of materials such as, for example, an abrasive material, a burnable poison, and/or a finishing coat. In accordance with one embodiment of the invention, a method for applying a burnable poison to a nuclear fuel rod that comprises providing a nuclear fuel rod having an axis and an outer surface having or not having a number of oxides and surface deposits, as well as providing at least one application device adjacent the surface of the nuclear fuel rod. The method also comprises rotating the nuclear fuel rod about its axis, or moving the at least one application device and holding the rod still, and removing a portion of any oxides and surface deposits on the outer surface of the nuclear fuel rod by spraying an abrasive material onto the nuclear fuel rod via the at least one application device while adjusting the position of the application device in relation to the nuclear fuel rod, or optionally not removing oxides or other surface deposits. The method also comprises applying a burnable poison onto the surface of the nuclear fuel rod by spraying the burnable poison onto the nuclear fuel rod via the application device while the position of the at least one application device is adjusted in relation to the nuclear fuel rod. In accordance with another embodiment of the invention, a method for applying a burnable poison to a nuclear fuel rod that comprises providing a nuclear fuel rod that has an axis and an outer surface which a number of oxides and surface deposits. The method also comprises providing an application device adjacent the surface of the nuclear fuel rod. The application device includes a channel that extends therethrough. The channel is in communication with a pressurized gas source as well as a particle source. The method further comprises providing an image capture device adjacent the nuclear fuel rod. The image capture device is adapted to transmit an image to a remote viewing station. The method also comprises rotating the nuclear fuel rod about its axis and removing a portion of the oxides and surface deposits on the outer surface of the nuclear fuel rod by spraying the nuclear fuel rod with an abrasive material via the application device as the position of the application device is adjusted in relation to the nuclear fuel rod. Specifically, an abrasive particle is introduced into the channel of the application device via the particle source and pressurized gas is expelled from the pressurized gas source through the channel of the application device thereby spraying the abrasive particle onto the surface of the nuclear fuel rod. The method further comprises applying a burnable poison onto the surface of the nuclear fuel rod by spraying the burnable poison onto the nuclear fuel rod via the application device as the position of the application device is adjusted in relation to the nuclear fuel rod. Specifically, a burnable poison particle is introduced into the channel of the application device via the particle source and pressurized gas is expelled from the pressurized gas source through the channel of the application device thereby spraying the burnable poison onto the surface of the nuclear fuel rod. In accordance with yet another embodiment of the invention, a method for applying a burnable poison to a nuclear fuel rod that comprises (a) providing a nuclear fuel rod having an axis and an outer surface which has a number of oxides and surface deposits as well as (b) providing an application device adjacent the surface of the nuclear fuel rod. The method also comprises (c) rotating the nuclear fuel rod about its axis and (d) stopping the rotation of the nuclear fuel rod about its axis. The method further comprises (e) removing a portion of the oxides and surface deposits on the outer surface of the nuclear fuel rod by spraying an abrasive material onto the nuclear fuel rod via the application device while adjusting the position of the application device is adjusted in relation to the nuclear fuel rod as well as (f) applying a burnable poison onto the surface of the nuclear fuel rod by spraying the burnable poison onto the nuclear fuel rod via the application device while adjusting the position of the application device is adjusted in relation to the nuclear fuel rod. The pressure of the gas, the particle size and the distance from the rod cladding is adjusted to be effective so that the particles impact at a velocity high enough to cause surface activity, such as, in the case of a metal cladding, to melt a surface layer of the metal of the fuel rod cladding. Subsequent water contact with the impacted, layered particles will form a protective oxide coating on the outside of the exterior particle layer, such that it keeps the other interior burnable poison layers from dissolving in the reactor coolant. This is an unexpected result, since the burnable poisons are generally soluble with aqueous reactor coolant. A major advantage of having the burnable poison outside the fuel rod cladding is that, for example, if boron is used as part of the burnable poison, resulting helium gas generated during nuclear reaction will not over pressurize the inside of the fuel rod. As employed herein, the term “number” means one or an integer greater than one (i.e., a plurality). As employed herein, the term “burnable poison” refers broadly to a material that captures neutrons without giving out neutrons but has the capacity for absorbing neutrons reduced over time (burnable)—as it absorbs neutrons, for example, elemental boron or boron containing compounds such as ZrB2 or HfB2; rare earths such as elemental Hf, Gd or Er; and rare earth oxides, preferably Gd2O3 and Er2O3, and their mixtures. The most preferred material is ZrB2. When referring to any numerical range of values, such ranges are understood to include each and every number and/or fraction between the stated range minimum and maximum. Directional phrases used herein, such as, for example, upper, lower, left, right, vertical, horizontal, top, bottom, above, beneath, clockwise, counterclockwise and derivatives thereof, relate to the orientation of the elements shown in the drawings and are not limiting upon the claims unless expressly recited therein. Referring to FIG. 1, a nuclear fuel rod 2, containing fuel pellets or particles (not shown), having an axis 4 and an outer surface 6, such as metal cladding (for example zirconium based metal) or ceramic cladding (for example SiC) is provided. At least one application device 8 is positioned adjacent the outer cladding surface 6 of the nuclear fuel rod 2. The nuclear fuel rod 2 is connected to an apparatus (not shown) that is adapted to rotate the nuclear fuel rod 2 about its axis 4. In one embodiment, the rotation of the nuclear fuel rod 2 about its axis 4 is continuous. In another embodiment, however, the rotation of the nuclear fuel rod 2 is incremental (i.e., the nuclear fuel rod 2 is rotated then the rotation of the nuclear fuel rod 2 is stopped). If the nuclear rod 2 is continuously rotated about its axis 4, then in one embodiment, the rate of rotation is adjusted such that the a deposit of burnable poison particles of the required thickness is obtained. This thickness can involve 2 or more layers and the total thickness can range from 0.001 mil to 10 mils (1 mil=0.001 inch). Under 0.001 mil. (0.025 micrometers). There is too little neutron absorption and over 10 mils (254 micrometers) there is too much neutron absorption by the burnable poison, so that it becomes difficult to start the reactor. The application device 8 includes a channel 10 that extends therethrough. The channel 10 is in communication with a pressurized gas source 14 via a gas inlet tube 16 and a particle source 18 via a particle inlet tube 20. Pressurized gas that is expelled from the gas source 14 travels through the gas inlet tube 16 and the channel 10 of the application device 8 thereby propelling particles 22, which are injected into the channel 10 via the particle inlet tube 20. The particles 22 from the application device 8 spray the nuclear fuel rod 2 with the particles 22. As will be discussed in greater detail below, the particles 22 that are sprayed onto the nuclear fuel rod 2 can include, for example, initially and optionally, abrasive particles, and then burnable poison particles having a particle size of from 1 micrometer to 250 micrometers, and finally, optionally, already protective particles that can be used to coat the burnable poison. These latter outer already protective coating particles include Zr metal particles. Any particles applied are non-electrostatic, that is they are not charged electrically and caused to adhere merely by electrostatic means, as that would not provide sufficient intimate, very adherent layers. A very important feature of the invention is spraying the cladding with burnable poison particles 33 at such a velocity, for example 1,500 ft./second to 2,500 ft./second (457 meters/second to 762 meters/second), to initiate a surface phase change, generally shown as 30, at the exterior of the cladding, especially for metal cladding, so that some molecular surface melting occurs and the impacting particles adhere, forming a base particle layer, 42 in FIG. 2, that improves further layer adhesion. The larger the particle size the less velocity needed. By “molecular surface melting” is meant anything from interatom bonding between metal cladding and the burnable poison particles, to forming melt craters. The top burnable poison layer, 46 in FIG. 2, of, for example ZrB2 burnable poison particles is subject, during operation in a nuclear environment 50 on FIG. 2, such as a nuclear reactor, illustrated in FIG. 2, and described further on in the application, to oxidation by the passing cooling water thus forming a protective ZrO2 product that inhibits further reaction of the ZrB2 layers beneath 42 and 44 in FIG. 2, by the reaction. B2O3+3 H2O→2 H3BO3 (Soluble) It had been assumed that all ZrO2 would be removed/dissolved and carried away by the reactor coolant. Optionally, additional inherently already protective material such as Zr metal can be applied to the outer surface. The application device 8 is mounted onto a translocation apparatus (not shown), such as a robotic arm, which is adapted to adjust the position of the application device 8 in relation to the nuclear fuel rod 2. For example, the robotic arm can translocate (move) the application device 8 along the length of the nuclear fuel rod 2. Additionally, the translocation apparatus can also adjust the angle application device 8 in relation to the nuclear fuel rod 2. For instance, the robotic arm can position the application device 8 such that the application device 8 is positioned co-axial to the nuclear fuel rod 2. An image capture device 24 may also be positioned adjacent the nuclear fuel rod 2. The image capture device 24 is adapted to transmit an image to a remote viewing station 26 so that the process of coating the nuclear fuel rod 2 may be monitored. In one embodiment the image capture device 24 is also connected to the robotic arm to which the application device 8 is mounted. In another embodiment, however, the image capture device 24 is mounted to another translocation apparatus (not shown) so that the image capture device 24 may be translocated independent of the application device 8. In one embodiment of the process, the nuclear fuel rod 2 is rotated continuously about its axis 4. As the nuclear fuel rod 2 is rotated, the application device 8 is positioned adjacent to the outer surface 6 of the nuclear fuel rod 2. Once the application device is in the proper position, an abrasive material, such as particles of aluminum oxide or boron nitride, is sprayed onto the outer surface 6 of the nuclear fuel rod 2 via the application device 8 in order to remove one or more oxides and/or surface deposits that are disposed on the outer surface 6 of the nuclear fuel rod 2. The process of removing the oxides and/or surface deposits from the nuclear fuel rod 2 is typically done in order to provide a substantially clean surface onto which a burnable poison can be applied onto the nuclear fuel rod 2 in one or more subsequent steps. As the abrasive material is expelled from the application device 2, the position and/or angle of the application device 8 is adjusted in relation to the rotating nuclear fuel rod 2. Note that this step may not be necessary as the oxides or other surface contaminants may have been removed previously or may be low enough to provide acceptable adhesion of the burnable poison to the exterior surface of the rod. After a portion of the oxides and/or surface deposits are optionally removed from the nuclear fuel rod 2, the application device 8 is used to apply a burnable poison onto the nuclear fuel rod 2 as the nuclear fuel rod 2 continues to rotate. Specifically, elemental particles of a burnable poison such as, without limitation, elemental boron, elemental gadolinium, elemental hafnium or elemental erbium, Er2O3, Gd2O3, HfB2, or ZrB2 is expelled from the application device 8 and sprayed onto the nuclear fuel rod 2. These particles are shown specifically as particles 33. The burnable poison particles 33 would always be applied after (not with) application of optional particles 22 such as abrasive cleaning particles (shown in the same FIG. 1 but particles 22 and 33 are sprayed at separate times and never together). As the burnable poison particles are sprayed onto the nuclear fuel rod 2, the position and/or angle of the application device 8 is adjusted in relation to the nuclear fuel rod 2 such that a layer of burnable poison begins to accumulate on the outer surface 6 of the nuclear fuel rod 2. The velocity at which the burnable poison is expelled from the application device 8 is dependent upon that rate at which the pressurized gas is expelled from the pressured gas source 14 as well as the mass of the particles that are introduced into the channel 10 of the application device 8. In one embodiment, a plurality of substantially uniform layers of burnable poison is deposited onto the nuclear fuel rod 2 which total thickness layers can range from about 0.001 mil to 10 mils (coating thickness). In another embodiment, the layers are substantially non-uniform. That is, for example an initial layer, to example 42, may have particle size of 300 micrometers to cause good adhesive impact, followed by 2 to 10 layers of 50 micrometer particles. After application, the rod can be discharged immediately from the application device. After the process of applying the burnable poison onto the nuclear fuel rod 2 is complete, an inherently already protective finishing coat may optionally be applied onto the deposited burnable poison layers. The finishing coat may be a metal such as, without limitation, zirconium, hafnium, titanium or similar material that provides a protective barrier coating for the burnable poison that is already sprayed and deposited. As with previous steps, the position and/or angle of the application device 8 is typically adjusted in relation to the nuclear fuel rod 2 during the process of depositing the finishing coat onto the burnable poison. In other embodiments, a nuclear fuel rod 2 having a finishing coat may optionally be subjected to various downstream processes, such as mechanical processing, in order to impart a desired surface finish on the coated nuclear fuel rod 2. In another embodiment of the invention, the nuclear fuel rod 2 is not continuously rotated about its axis 4. Rather, the nuclear fuel rod 2 is incrementally rotated about its axis 4. In this particular embodiment, the nuclear fuel rod 2 is rotated incrementally so that the burnable poison may be deposited onto the outer surface 6 of the nuclear fuel rod 2 in a strip-like manner. Accordingly, one would appreciate that in this particular embodiment the nuclear fuel rod 2 will have a number of strips of burnable poison that extend along the length of the nuclear fuel rod 2 substantially parallel to the axis 4 of the nuclear fuel rod 2. In another embodiment, the rod 2 can be stationary and one or more application devices 8 can be rotated transversely about the rod axis. Also shown in FIG. 1 is arrow 28 representing the velocity of the powder particles, especially the burnable poison particles 33; the arrow 29 showing one direction of the sprayer motion; and arrow 31 showing one embodiment of the rod rotation. The initial layer of burnable poison particles 33 impinge on the metal or ceramic surface 6 of cladding 35, which is shown as a tube, with a velocity effective to cause a surface “action” 30, such as a slight molecular surface melting, inner atom diffusion between the burnable poison and metal or ceramic cladding, or atom bonding due to crystal formation; which causes adherence of the particles 33, to provide one or more particle layers 32 in FIG. 1. These burnable particle layers can each be all the same size (diameter) particles or a mixture of particle sizes. Referring now to FIG. 2, a fuel rod 2, in a nuclear environment 50, having cladding 35 and a cladding outer surface 6 is shown in section. The cladding, shown here, contains interior fuel pellets 40, which may be inserted within the fuel rod before or after spraying. A first applied burnable particle layer is shown as 42, second and third applied burnable particle layers are shown as 44 and 46 respectively. Cooling water 48 is shown passing near outer/contacting burnable layer outer 46, which aqueous coolant material will react, in a heated/hot environment, usually about 200° C. to about 360° C. with the burnable particles to form a protective oxide on or in layer 46 (shown hatched) which will protect the interior burnable layers 42 and 44. While specific embodiments of the invention have been described in detail above, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular arrangements disclosed are meant to be illustrative only and not limiting as to the scope of the disclosed and claimed concept which is to be given the full breadth of the claims appended and any and all equivalents thereof. For instance, as mentioned above, moving the application device and not the rod. |
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claims | 1. An X-ray CT apparatus comprising:a scanning table configured to support a subject thereon and to move the subject within the X-ray CT apparatus;a scanning gantry comprising:an X-ray generator;an X-ray detector configured to detect X-rays generated by the X-ray generator, the X-ray detector positioned in opposition to the X-ray generator; anda rotation device configured to rotate the X-ray generator and the X-ray detector, the X-ray generator configured to expose the X-rays to the subject moved by the scanning table while the X-ray generator and the X-ray detector are rotated about the subject, the scanning gantry configured to perform a scan including detecting the X-rays transmitted through the subject at the X-ray detector to acquire X-ray projection data;a scanning condition setting device configured to set parameters for controlling a movement of the scanning table along a moving direction during the scan, the parameters including an acceleration and a deceleration of the movement of the scanning table;a predicting device configured to predict a plurality of positions of the scanning table along the moving direction for each view of the scan by calculating the plurality of positions of the scanning table using the parameters set by the scanning condition setting device; andan image reconstructing device configured to reconstruct a plurality of tomographic images within a range scanned during the acceleration and the deceleration of the movement of the scanning table along the moving direction by reconstructing the X-ray projection data, wherein the X-ray projection data is correlated to the plurality of predicted positions. 2. The X-ray CT apparatus according to claim 1, wherein the parameters comprise at least one of a scanning table acceleration, a scanning table deceleration, a scanning table constant velocity, a scanning table initial position, a scanning table stop position, a scanning table acceleration end position, and a scanning table deceleration start position. 3. The X-ray CT apparatus according to claim 1, wherein the predicting device is further configured to add the parameters to the X-ray projection data as a part of header information of the X-ray projection data. 4. The X-ray CT apparatus according to claim 1, wherein the predicting device is further configured to record the parameters to a file associated with the X-ray projection data. 5. The X-ray CT apparatus according to claim 1, wherein the scanning condition setting device is further configured to set as one of the parameters a tilt parameter for controlling a tilt angle of the scanning gantry during the scan, and wherein the predicting device is configured to calculate the plurality of positions by using the parameters and the tilt parameter. 6. The X-ray CT apparatus according to claim 1, wherein the scan is a helical shuttle scan. 7. The X-ray CT apparatus according to claim 1, wherein the image reconstructing device is configured to perform a three-dimensional image reconstruction. 8. A method for producing an X-ray CT image by reconstructing projection data, said method comprising:obtaining the projection data during a scan including an acceleration and a deceleration of a movement of a scanning table using an X-ray CT apparatus, wherein the X-ray CT apparatus comprises the scanning table configured to support a subject thereon and to move the subject within the X-ray CT apparatus, and a scanning gantry comprising an X-ray generator, an X-ray detector configured to detect X-rays generated by the X-ray generator, the X-ray detector positioned in opposition to the X-ray generator, and a rotation device configured to rotate the X-ray generator and the X-ray detector, the X-ray generator configured to expose the X-rays to the subject moved by the scanning table while the X-ray generator and the X-ray detector are rotated about the subject, the scanning gantry configured to perform the scan including detecting the X-rays transmitted through the subject at the X-ray detector to acquire the projection data;predicting a plurality of positions of the scanning table along a moving direction for each view of the scan by calculating the plurality of positions of the scanning table using parameters for controlling movement of the scanning table along the moving direction; andreconstructing a plurality of tomographic images within a range scanned during the acceleration and the deceleration of the movement of the scanning table by reconstructing the projection data, wherein the projection data are correlated to the plurality of predicted positions. 9. The method for producing X-ray CT image according to claim 8, wherein predicting a plurality of positions further comprises calculating the plurality of positions of the scanning table using parameters comprising at least one of a scanning table acceleration, a scanning table deceleration, a scanning table constant velocity, a scanning table initial position, a scanning table stop position, a scanning table acceleration end position, and a scanning table deceleration start position. 10. The method for producing X-ray CT image according to claim 8 further comprising adding the parameters to the projection data as a part of header information of the projection data. 11. The method for producing X-ray CT image according to claim 8 further comprising recording the parameters to a file associated with the projection data. 12. The method for producing X-ray CT image according to claim 8, wherein predicting a plurality of positions of the scanning table further comprises using the parameters and a tilt parameter for controlling a tilt angle of the scanning gantry during, the scan to predict the plurality of positions. 13. The method for producing X-ray CT image according to claim 8, wherein obtaining the projection data during a scan further comprises performing a helical shuttle scan. 14. The method for producing X-ray CT image according to claim 8, wherein reconstructing a plurality of tomographic images further comprises performing a three-dimensional image reconstruction. |
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claims | 1. A scintillation crystal comprising NaX:Tl, La, wherein:X represents a halogen;each of Tl and La has a dopant concentration of at least 1×10−5 mol %; andthe scintillation crystal has an energy resolution less than 6.4% when measured at 662 keV, 22° C., and an integration time of 1 microsecond. 2. The scintillation crystal of claim 1, wherein La has a dopant concentration of at least 5×10−4 mol %. 3. The scintillation crystal of claim 1, wherein La has a concentration no greater than 0.9 mol %. 4. The scintillation crystal of claim 1, wherein X is iodine. 5. A method comprising;providing the scintillation crystal of claim 1;capturing radiation within the scintillation crystal;determining a pulse decay time and an actual light yield of the radiation captured;determining an estimated light yield corresponding to the pulse decay time; andcalculating an adjusted light yield that is a product of the actual light yield times the light yield of NaX:Tl divided by the estimated light yield. 6. A scintillation crystal comprising NaX:Tl, Sr, wherein:X represents a halogen;each of Tl and Sr has a concentration of at least 1×10−5 mol %; andthe scintillation crystal has an energy resolution less than 6.0% when measured at 662 keV, 22° C., and an integration time of 1 microsecond. 7. The scintillation crystal of claim 6, wherein Sr has a concentration no greater than 5 mol %. 8. The scintillation crystal of claim 6, wherein the scintillation crystal has a greater light yield as compared to a NaI:Tl crystal when the scintillation crystal and the NaI:Tl crystal are measured at 22° C. 9. The scintillation crystal of claim 6, wherein X is I. 10. The scintillation crystal of claim 6, wherein the scintillation crystal has a pulse decay time that is at least 5% less than a pulse decay time a NaI:Tl crystal when the scintillation crystal and the NaI:Tl crystal are measured at 22° C. and exposed to gamma radiation having an energy of 662 keV. 11. The scintillation crystal of claim 6, wherein at energies in the range of 32 keV to 81 keV, the scintillation crystal has an average relative light yield as normalized to a light yield at 2615 keV of no greater than 1.15. 12. The scintillation crystal of claim 6, wherein at energies in the range of 122 keV to 511 keV, the scintillation crystal has an average relative light yield as normalized to a light yield at 2615 keV no greater than 1.07. 13. The scintillation crystal of claim 6, wherein Tl has a concentration in a range of 1×10−4 mol % to 0.2 mol %. 14. A scintillation crystal comprising NaX:Tl, Ca, wherein:X represents a halogen;each of Tl and Ca have a concentration of at least 1×10−5 mol %; andthe scintillation crystal has a lower energy resolution as compared to a NaI:Tl crystal when the scintillation crystal and the NaI:Tl crystal are measured at 662 keV, 22° C., and an integration time of 1 microsecond. 15. The scintillation crystal of claim 14, wherein Ca has a concentration no greater than 5 mol %. 16. The scintillation crystal of claim 14, wherein X is I. 17. The scintillation crystal of claim 14, wherein the scintillation crystal has a pulse decay time that is at least 5% less than a pulse decay time a NaI:Tl crystal when the scintillation crystal and the NaI:Tl crystal are measured at 22° C. and exposed to gamma radiation having an energy of 662 keV. 18. The scintillation crystal of claim 14, wherein at energies in the range of 32 keV to 81 keV, the scintillation crystal has an average relative light yield as normalized to a light yield at 2615 keV of no greater than 1.15. 19. The scintillation crystal of claim 14, wherein at energies in the range of 122 keV to 511 keV, the scintillation crystal has an average relative light yield as normalized to a light yield at 2615 keV no greater than 1.07. 20. The scintillation crystal of claim 14, wherein Tl has a concentration in a range of 1×10−4 mol % to 0.2 mol %. |
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059563800 | summary | BACKGROUND OF THE INVENTION FIELD OF THE INVENTION The invention relates to a method and an apparatus for determining the neutron flux density of a neutron-emitting source, in particular a reactor core in a nuclear power facility, having a plurality of fuel assemblies. An apparatus for determining neutron flux density, a so-called neutron measurement system, is often used in a nuclear power facility to monitor start-up and shut-down procedures. The neutron measurement system has neutron detectors which, in particular, supply a signal proportional to the neutron flux density. The neutron flux density, when a nuclear power facility is in the shut-down, sub-critical state, differs by several orders of magnitude from that when the nuclear power facility is producing power. A book entitled "Strahlung und Strahlungsme.beta.technik in Kernkraftwerken" Radiation and Radiation Metrology in Nuclear Power Facilities!, published by Elmar Schrufer, Elitera Verlag, Berlin, 1974 deals comprehensively with the construction and method of operation of neutron detectors, particularly in Sections 3.4, 6.1 and 6.2 thereof. As described in Section 6.1, neutron flux measurement systems with neutron detectors are used to determine the reactor power of a nuclear power facility. In that case a neutron detector may be disposed both outside and inside the reactor core, between adjacent fuel assemblies or elements. The neutron detector may be configured in such a way that it can be moved along a major axis. It can thus be removed from the reactor core during regular power operation of the nuclear power facility. With regard to the method of operation and measurement accuracy, a distinction is drawn between three different systems for neutron detectors. So-called pulsed systems are preferably used in a boiling water nuclear power facility, where the intention is to achieve high sensitivity in strong gamma radiation. Since in that case low pulse rates can be measured much more easily than very small currents, pulsed systems can also be used to measure lower neutron flux densities. The low gamma pulse level allows discrimination between neutrons and the gamma radiation, for example by using threshold value discriminators. The operating range of a pulsed system covers neutron flux densities in a range from about 10.sup.-1 neutrons/(cm.sup.2 .multidot.s) to about 10.sup.5 neutrons/(cm.sup.2 .multidot.s). That corresponds to a reactor power level of up to about 10.sup.-3 %. So-called direct-current systems are preferably suitable for medium and high neutron flux levels in a range from about 10.sup.2 neutrons/(cm.sup.2 .multidot.s) to about 10.sup.9 neutrons (cm.sup.2 .multidot.s). Discrimination between neutrons and the gamma radiation is preferably carried out using so-called gamma-compensated ionization chambers. At low neutron flux levels, the use of a direct-current system is generally limited by the influence of the gamma radiation. A neutron detector in a direct-current system is preferably based on a fission chamber and/or boron meter, as are described, for example, in Section 3.4 of the above-mentioned book. In the case of the so-called alternating-current system, the alternating current which is produced in a fission chamber, ionization chamber or boron meter is superimposed on the direct current and is used to form information. Due to the high ionization rate of the gas in such a chamber, it is no longer possible to resolve the individual pulses from the ionized gas separately. In consequence, a direct current is produced as a mean value, and an alternating current is superimposed thereon. The mean square value of the alternating-current signal is directly proportional to the neutron flux density, as is the direct current as well. The ratio of the signal from the detected neutrons to the signal caused by gamma radiation in an alternating-current system may be greater by a factor of 1000 than that in a direct-current system. An alternating-current system is thus preferably suitable for the medium and high ranges of reactor power levels, with neutron flux densities between 10.sup.6 neutrons/(cm.sup.2 .multidot.s) and 10.sup.14 neutrons/(cm.sup.2 .multidot.s) . An alternating-current system is thus also suitable for the power range of a nuclear power facility. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a method and an apparatus for determining neutron flux density, in particular in a nuclear power facility, which overcome the hereinafore-mentioned disadvantages of the heretofore-known methods and apparatuses of this general type, in which the method uses different measurement signals in a simple and reliable manner and over a wide range to determine the neutron flux density, in particular from a shut-down to a normal power range of a nuclear power facility, and in which the apparatus determines the neutron flux density of a source that emits neutrons, in particular in a nuclear power facility. With the foregoing and other objects in view there is provided, in accordance with the invention, a method for determining the neutron flux density of a neutron-emitting source, which comprises forming a wide-range signal W depending uniquely on a neutron flux density from a first measurement signal S.sub.1 and at least one second measurement signal S.sub.2 differing from one another and each dependent on the neutron flux density; selecting the first measurement signal S.sub.1 as a monotonal function of the neutron flux density for values of the neutron flux density less than a first limit flux density; equating the wide-range signal W to the first measurement signal S.sub.1 in accordance with the relationship W=f.sub.1 (S.sub.1)=S.sub.1 for values of the first measurement signal S.sub.1 corresponding to a neutron flux density less than a lower limit value which is less than the first limit flux density and to which a first signal value N.sub.1 of the first measurement signal S.sub.1 is allocated; forming the wide-range signal W as a function f.sub.2 of the measurement signals S.sub.1, S.sub.2 in accordance with the relationship W=f.sub.2 (S.sub.1, S.sub.2) for values of the first measurement signal S.sub.1 corresponding to a neutron flux density greater than the lower limit value; making the wide-range signal W continuous at the lower limit value in accordance with the relationship f.sub.1 (N.sub.1)=f.sub.2 (N.sub.1, S.sub.2); and using the wide-range signal W as a basis for determining the neutron flux density. Due to the unique dependency of the wide-range measurement signal on the neutron flux density, it is possible, in the case of a nuclear power facility, to determine the power level of the reactor of that nuclear power facility. This applies in particular to a nuclear power facility having a boiling water reactor or a pressurized water reactor. In this case, the lower limit value and the first limit flux density of the neutron flux density are each uniquely allocated a value of the first measurement signal. Therefore, it is also always possible to use the associated value of the first measurement signal, instead of the value of the neutron flux density. The first signal value (allocated to the lower limit value) of the first measurement signal can thus be used as the limit value when carrying out the method until the wide-range signal is formed from the first measurement signal. The wide-range signal is formed as a function of the first measurement signal and of the second measurement signal for values of the first measurement signal which are greater than the first signal value. This function thus depends on the first measurement signal and the second measurement signal such that, if the value of the first measurement signal is equal to that of the first signal value, it just corresponds to the first signal value. This ensures that, irrespective of the choice of the first signal value, that is to say of the lower limit value of the neutron flux density, the wide-range signal is continuous in all cases. In accordance with another mode of the invention, the second measurement signal is a monotonal function of the neutron flux density for values of the neutron flux density which are greater than a second limit flux density, and the second limit flux density is less than the first limit flux density. An overlapping region is used and is defined between the lower limit value and an upper limit value, with the upper limit value of the neutron flux density being greater than the second limit flux density. This ensures that, in the overlapping region, both the first measurement signal and the second measurement signal are respective monotonal functions of the neutron flux density in an interval defined by the first limit flux density and the second limit flux density. Both measurement signals preferably rise or fall monotonally. This provides a simple way of ensuring the unique dependency of the wide-range signal on the neutron flux density, even in the overlapping region. Due to the monotonal nature of the first measurement signal for neutron flux densities below the first limit flux density, and to the monotonal nature of the second measurement signal for neutron flux densities above the second limit flux density, the overlapping region is also uniquely defined by the first signal value of the first measurement signal and the second signal value of the second measurement signal. Thus, without any knowledge of the neutron flux density, it is possible to use just the values of the first measurement signal and of the second measurement signal to define the regions in which the wide-range signal is equated to the first measurement signal or is formed as a function of the first measurement signal and of the second measurement signal. This is irrespective of whether the first measurement signal and the second measurement signal rise or fall monotonally. It is self-evident that, in an analogous manner to the determination of the overlapping region, a monotonal wide-range signal allows a further overlapping region to be determined as a function of the first measurement signal and of the second measurement signal. In such a further overlapping region, which may be adjacent the first-mentioned overlapping region for smaller or larger values of the neutron flux density, the wide-range signal may be composed of a third measurement signal and the first or the second measurement signal, to provide a continuous transition for the wide-range signal at the region boundaries of the further overlapping region, in an analogous manner to the first-mentioned overlapping region. The method is particularly advantageous above all in nuclear power facilities in which the wide-range signal is also fed into the safety system of the nuclear power facility, since the safety system may include a rapid or unexpected change in the neutron flux density or the reactor power as a triggering criterion for scramming the nuclear power facility, and both of these parameters can be identified by the wide-range signal. Any discontinuity in the wide-range signal could, under some circumstances, lead to unjustified scramming of the nuclear power facility. In such a case, that is avoided from the start by the method as a consequence of the specified determination of the wide-range signal. In accordance with a further mode of the invention, the wide-range signal is equated to the second measurement signal for values of the second measurement signal to which a neutron flux density greater than the upper limit value is allocated. The wide-range signal is thus a function which is defined region-by-region and is defined differently in three respectively directly mutually adjacent regions. It is only in the overlapping region that there is a function of both the first measurement signal and the second measurement signal, with a continuous transition being ensured at the region boundaries of the overlapping region. As already mentioned, the wide-range signal may also be defined differently in four or more regions, although it is intended that the transition of the wide-range signal be continuous at each region boundary and that the wide-range signal should be monotonal within each region. These continuous transitions are preferably ensured by a function which, as a result of the fact that it is composed of the first measurement signal and the second measurement signal, coincides with the first measurement signal at the lower limit value, and coincides with the second measurement signal at the upper limit value. At the region boundaries and outside the overlapping region, the function is thus always dependent on only a single measurement signal, and precisely on that measurement signal which is a monotonal function of the neutron flux density outside the overlapping region, up to the respective limit value. Within the overlapping region, the function is preferably formed as a sum of the product of the first measurement signal and a factor dependent on the second measurement signal, and the product of the second measurement signal and a factor dependent on the first measurement signal. The factor which is dependent on the second measurement signal is formed in such a way that it assumes the value zero when the second measurement signal corresponds to the second signal value, that is to say at the upper limit value of the neutron flux density of the overlapping region. The factor that is dependent on the first measurement signal is defined in an analogous manner, so that it becomes zero when the first measurement signal assumes the first signal value, that is to say at the lower limit value of the overlapping region. The factor which is dependent on the second measurement signal is preferably the equation of the difference between the second signal value and the second measurement signal and the difference between the second signal value and the first signal value. The factor which is dependent on the first measuring signal is, correspondingly, the equation of the difference between the first measurement signal and the first signal value and the difference between the second signal value and the first signal value. In accordance with an added mode of the invention, the wide-range signal is a monotonally rising function of the neutron flux density. In accordance with an additional mode of the invention, this is the situation, for example, where the measurement signals are those from a neutron measurement signal which has an ionization chamber, a fission chamber, a boron meter, a so-called "self powered neutron" detector (SPN detector) or a counting tube, without using any recalculation or conversion. The first measurement signal and the second measurement signal are preferably produced in an ionization chamber, a fission chamber, a boron meter, an SPN detector or a counting tube, and it is always possible to produce both measurement signals in the same chamber or in the same counting tube. The construction and the method of operation of a neutron detector for producing a measurement signal which is, in particular, proportional to the neutron flux density is described, for example, in German Utility Model G 93 05 956.6 and in the book entitled "Radiation, Detection and Measurement" by Glenn F. Knoll, John Wiley & Sons, New York, 2.sub.nd Edition, 1985 in particular Chapters 5 and 14. That relates in particular to a neutron detector having a fission chamber or a boron meter, which are of such compact construction that they are suitable for use within a reactor, that is to say for so-called in-core instrumentation. In accordance with yet another mode of the invention, the first measurement signal is a so-called pulsed signal from an ionization chamber, and the signal is produced by gas atoms or gas molecules ionized by neutrons. A pulsed signal is distinguished by the fact that, particularly in the case of low neutron flux densities of less than 10.sup.5 neutrons/(cm.sup.2 .multidot.s), it is a particularly good measurement signal which can be clearly distinguished from other types of radioactive radiation, in particular gamma radiation. At low neutron flux densities, the pulsed signal is proportional to the neutron flux density. The second measurement signal is preferably a direct-current or alternating-current signal from the same ionization chamber. At high neutron flux densities, at which levels individual pulses from the ionized gas atoms or gas molecules can no longer be clearly separated from one another, the current signal is a measurement signal proportional to the neutron flux density. The current signal allows neutron flux densities from about 10.sup.3 neutrons/(cm.sup.2 .multidot.s) to about 10.sup.10 neutrons/(cm.sup.2 .multidot.s) to be detected reliably. The current produced in an ionization chamber by the ionized gas atoms or gas molecules contains an alternating-current element, due to the fluctuation in the number of pulses. This alternating-current element is squared and then averaged to obtain an alternating-current signal which is proportional to the neutron flux density. The alternating-current signal is preferably determined by using the so-called cross-correlation method, in which a signal is picked off both from an internal electrode and from an external electrode of the ionization chamber. These signals are of opposite polarity, but are otherwise identical. They are used for a cross-correlation function, which is proportional to the mean square of the alternating-current element, and thus to the neutron flux density. The cross-correlation method makes it possible to eliminate scatter and statistical disturbances (noise) which are in each case contained in only one of the signals. The second measurement signal that is formed in this way is directly proportional to the neutron flux density up to a lower neutron flux density of about 10.sup.3 neutrons/(cm.sup.2 .multidot.s). The maximum neutron flux density which can be measured for the second measurement signal is about 10.sup.10 neutrons/(cm.sup.2 .multidot.s). In this way, particularly in a pressurized water nuclear power facility, the neutron flux density can be reliably detected in a range from 10.sup.-5 % to 100% of the reactor power. What has been said above applies analogously to a fission chamber and to a counting tube. In accordance with yet a further mode of the invention, the method for determining the neutron flux density is used in a nuclear power facility. In this case, the source which emits the neutrons is the reactor core of the nuclear power facility, having a plurality of fuel assemblies. The neutron flux density is determined outside the reactor core and, through the use of the so-called in-core instrumentation, between the fuel assemblies within the reactor core. It is possible to determine the neutron flux density during a starting-up process, a shutting-down process or during normal operation of the nuclear power facility, over the entire power range, by producing the wide-range signal, which is uniquely dependent on the neutron flux density. Defining the wide-range signal region-by-region in such a manner that, by definition, the wide-range signal changes over continuously at the region boundaries, ensures that the neutron flux density can be detected reliably even in the event of changes within the reactor core, for example in the event of the fuel elements or assemblies burning away, or in the event of changes in the neutron measurement systems, for example as a result of burning away or as a result of material or temperature fatigue. With the objects of the invention in view there is also provided an apparatus for determining the neutron flux density of a neutron-emitting source, comprising a measurement device for producing and transmitting a first measurement signal S.sub.1 and a second measurement signal S.sub.2 differing from one another and each dependent on a neutron flux density, the first measurement signal S.sub.1 being a monotonal function of the neutron flux density, in particular monotonally rising, for values of the neutron flux density less than a first limit flux density; and an evaluation device to be connected to the measurement device for forming a wide-range signal W depending uniquely on the neutron flux density and used to determine the neutron flux density, wherein the wide-range signal W corresponds to the first measurement signal S.sub.1 in accordance with the relationship W=f.sub.1 (S.sub.1)=S.sub.1 when the first measurement signal S.sub.1 assumes values corresponding to a neutron flux density of less than a lower limit value which is less than the first limit flux density and to which a first signal value N.sub.1 of the first measurement signal S.sub.1 is allocated; the wide-range signal W is a function f.sub.2 of the measurement signals S.sub.1 and S.sub.2 in accordance with the relationship W=f.sub.2 (S.sub.1, S.sub.2) when the first measurement signal S.sub.1 assumes values corresponding to a neutron flux density greater than the lower limit value; and the wide-range signal W is continuous at the lower limit value in accordance with the relationship f.sub.1 (N.sub.1)=f.sub.2 (N.sub.1, S.sub.2). The wide-range signal in this case is identical to the first measurement signal, as long as the first measurement signal is less than or greater than a predetermined first signal value, depending on whether the first measurement signal rises or falls monotonally as a function of the neutron flux density. If the first measurement signal is greater than the first signal value (monotonally rising profile), then the wide-range signal is formed from a function which depends both on the first measurement signal and on the second measurement signal. The function has the characteristic of being identical to the first measurement signal if the latter just assumes the first signal value. This results in a continuous transition of the wide-range signal at the first signal value which is, to be precise, irrespective of the value of the second measurement signal and irrespective of the profile of the second measurement signal as a function of the neutron flux density. A corresponding situation applies in the event of the first signal value being undershot for the case in which the first measurement signal falls monotonally as a function of the neutron flux density. A corresponding continuity also occurs at a transition at which the wide-range signal is formed by another function composed of one or more, but at least two, measurement signals. In order to determine the neutron flux density, the evaluation device may, for example, be a computer with a computation program, or an electronic circuit. In addition to a neutron detector, an ionization chamber, a fission chamber, a boron meter, an SPN detector or a counting tube, the measurement device may have corresponding electrical cables as well as operational amplifiers. 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 and an apparatus for determining neutron flux density, in particular in a nuclear power facility, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. |
claims | 1. A nuclear reactor lower internals structure having an axial and a circumferential dimension comprising:a tubular core barrel supported coaxially within the lower internals and having an inside vertically extending surface and an upper and lower end; anda reflector having a plurality of axial and circumferential segments and an outside curvature that substantially matches a curvature of the inside surface of the core barrel and contacts at least a substantial portion of the inside surface of the core barrel covered by the reflector, is supported substantially over at least a central axial length of the core barrel that extends substantially from or above the lower end toward the upper end and over at least a portion of the circumferential dimension covered by the reflector, the reflector further has an inside contour that matches an outside contour of an abutting array of a plurality of nuclear fuel assemblies that are designed to be supported within the core barrel to form a nuclear core and the axial and circumferential reflector segments are separately, fixedly connected to and respectively supported by the inside vertically extending surface of the core barrel at a plurality of axial and circumferential locations and wherein the reflector does not extend substantially continuously from the lower end to the upper end of the core barrel and is divided into axial sections that interface with each other at an acute angle with the inside surface of the core barrel. 2. The nuclear reactor lower internals structure of claim 1 wherein the circumferential sections have a stepped interface with each other formed from alternating angles wherein adjacent alternating angles are bent in opposite directions. 3. The nuclear reactor lower internals structure of claim 2 wherein the alternating angles are substantially right angles. 4. The nuclear reactor lower internals structure of claim 1 wherein at least two of the circumferential sections are spaced from each other and are separated by a baffle and former structure. 5. The nuclear reactor lower internals structure of claim 1 wherein the acute angle is approximately between 25° and 45°. 6. The nuclear reactor lower internals structure of claim 5 wherein the acute angle is approximately thirty degrees. 7. The nuclear reactor lower internals structure of claim 1 wherein the number of axial sections is either three or four. 8. The nuclear reactor lower internals structure of claim 1 wherein there is a space provided between interfacing axial sections. 9. The nuclear reactor lower internals structure of claim 1 wherein at least two of the circumferential sections are spaced from each other and are separated by a baffle and former structure, wherein a former plate is positioned to cover substantially each circumferential end of the interface of the axial sections. 10. The nuclear reactor lower internals structure of claim 1 wherein the circumferential sections are not contiguous. 11. The nuclear reactor lower internals structure of claim 1 including a plurality of axially extending coolant channels between the outside curvature of the reflector and the inside surface of the core barrel. 12. The nuclear reactor lower internals structure of claim 1 wherein the reflector does not extend substantially continuously from the lower end to the upper end of the core barrel and is divided into axial sections that interface with each other including axial coolant channels through the reflector that extend across the interface and tubular sleeves inserted and closely received within at least some of the coolant channels in the vicinity of the interface that span the interface. 13. The nuclear reactor lower internals structure of claim 1 wherein the core barrel is attached to a reactor vessel at a plurality of spaced circumferential locations by a plurality of attachment bracket arrangements that maintain a space between the reactor vessel and an outside of the core barrel, wherein at least two of the attachment bracket arrangements are positioned circumferentially around and affixed to an outside surface of the core barrel at substantially different axial elevations. 14. The nuclear reactor lower internals structure of claim 13 wherein adjacent attachment bracket arrangements are positioned at the substantially different axial elevations. 15. The nuclear reactor lower internals structure of claim 13 wherein the space between the reactor vessel and the outside of the core barrel forms a coolant path and the substantially different axial elevations are spaced far enough apart so that coolant blocked by one of the attachment bracket arrangements at an upper elevation is reformed directly below the one of the attachment bracket arrangement before the coolant reaches a lower axial elevation of the next attachment bracket arrangement at the substantially different axial elevation. 16. The nuclear reactor lower internals structure of claim 1 wherein the reflector includes a specimen basket for supporting a radiation specimen. 17. The nuclear reactor lower internals structure of claim 16 wherein the specimen basket is positioned within a hollowed out portion of the reflector. 18. The nuclear reactor lower internals structure of claim 17 wherein the hollowed out portion extends from a top surface of the reflector. |
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abstract | It is intended to provide a plasma processing method and apparatus capable of increasing the uniformity of amorphyzation processing. |
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062467417 | abstract | A fuel assembly for a pressurized water reactor having control rod guide thimbles (5) each having a dashpot (12) for protecting against flexural deformation which may impair insertability of a control rod The fuel assembly includes, a plurality of control rod guide thimbles (5) having bottom and top end portions fixedly secured to a lower nozzle (2) and an upper nozzle (4), respectively, disposed in opposition to each other. The dashpot (12) of each control rod guide thimble (5) includes a small diameter section (13b) having an outer diameter smaller than that of the control rod guide thimble (5) formed at an upper portion of the dashpot (12), and a large diameter section (13a) having an outer diameter substantially equal to that of the control rod guide thimble (5) formed at a lower portion of the dashpot (12). With the length of the control rod guide thimble (5) represented by L, the effective length (S) of the small diameter section (13b) is selected to lie within a range of from 0.03 L to 0.1 L. |
047042478 | abstract | A method and apparatus for withdrawing spent fuel rods from a nuclear fuel rod assembly into a different nuclear fuel rod container wherein the spent fuel rods have a higher fuel rod density, whereby a greater number of spent fuel rods can be stored in a water storage pool. The individual rods are drawn upwardly through a transition funnel from the fuel rod assembly into a fuel rod container. Individual wires extend through the fuel rod container, through the transition funnel and are secured to the top ends of the individual fuel rods within a fuel rod assembly. All of the fuel rods are withdrawn concurrently and are merged toward one another into a tighter array within the fuel rod container. |
045483477 | claims | 1. In an automated loading apparatus for nuclear reactor fuel pins: gravity conveyor means for rolling parallel fuel pin assemblies along an inclined path perpendicular to their lengths; transversely movable transport means interposed in said path; said transport means including transversely spaced fuel pin supports for coaxially positioning a fuel pin assembly along a preselected operational axis; transfer means movably mounted relative to both said conveyor means and said transport means for selectively engaging and moving individual fuel pin assemblies between said conveyor means and the supports of said transport means; powered means mounted on said transport means for selectively shifting the transport means relative to said conveyor means in a transverse direction that is parallel to said operational axis; fuel pin handling means at a location transversely adjacent to one side of the conveyor means and intersected by said operational axis for selectively receiving one end of a fuel pin assembly shifted thereto by operation of said powered means; and roller means mounted on said supports for selectively imparting rotational movement to a fuel pin assembly about said operational axis while the fuel pin assembly is positioned by said supports. stop means located at a position adjacent said transport means in the path of the fuel pin assemblies rolling along said conveyor means for selective engagement by an individual fuel pin assembly; and movable incline means for selectively shifting a fuel pin assembly from engagement with the stop means and for causing it to roll onto said supports of said transport means. means for inserting a charge of fuel pellets into a fuel pin assembly shifted thereto. means for cleaning one end of a fuel pin assembly shifted thereto. means for inserting a charge of fuel pellets into one end of a fuel pin assembly shifted thereto; means for closing and welding the end of a fuel pin assembly shifted thereto. a loading hopper for storing a plurality of partially completed fuel pin assemblies each including a length of cylindrical cladding having an enlarged funnel removably mounted at one open end of the cladding and an end cap welded at its remaining end; gravity feed conveyor means leading outward from the loading hopper for permitting controlled rolling motion of parallel fuel pin assemblies on inclined support rails arranged perpendicular to the lengths of said fuel assemblies; first and second transversely movable transport means interposed along said gravity feed conveyor means, each transport means including transversely spaced fuel pin supports for coaxially positioning a fuel pin assembly along a preselected operational axis; separate transfer means for selectively moving individual fuel pin assemblies between said gravity feed conveyor means and each of said first and second transport means; separate powered means operably mounted on each transport means selectively operable for shifting the transport means relative to said gravity feed conveyor means in a direction parallel to its operational axis; a fuel pin loading station transversely adjacent to said gravity feed conveyor means and interesected by said operational axis of said first transport means for selectively receiving said one open end of a fuel pin assembly shifted thereto by operation of said powered means while positioned on said supports of the first transport means; a fuel pin welding station transversely adjacent to said gravity feed conveyor means and intersected by said operational axis of said second transport means for selectively receiving said one open end of a fuel pin assembly shifted thereto by operation of said powered means while positioned on said supports of the second transport means; an enclosure containing means for inserting a charge of fuel pellets into each fuel pin assembly through the enlarged funnel mounted thereto while said one end of the fuel pin assembly is held at a loading position within the enclosure by said first transport means; and funnel removal means contained within the enclosure in the path of each fuel pin assembly as it is shifted parallel to the operational axis of the first transport means following its receipt of a charge of fuel pellets for engaging and stripping the enlarged funnel from said one end of the fuel pin assembly. an enclosure containing means for inserting a charge of fuel pellets into each fuel pin assembly through the enlarged funnel mounted thereto while said one end of the fuel pin assembly is held at a loading position within the enclosure by said first transport means; and funnel removal means contained within the enclosure in the path of each fuel pin assembly as it is shifted parallel to the operational axis of the first trolley following its receipt of a charge of fuel pellets for engaging and stripping the enlarged funnel from said one end of the fuel assembly; and means contained within the enclosure for wiping any contaminants from the surfaces of said one end of each fuel fin assembly after stripping of the enlarged funnel mounted thereto. an enclosure containing means for inserting a charge of fuel pellets into each fuel pin assembly through the enlarged funnel mounted thereto while said one end of the fuel pin assembly is held at a loading position within the enclosure by said first transport means; and sealing means in one wall of the enclosure traversing the operational axis of said first trolley means for preventing contaminants from escaping about a fuel pin assembly extending through the sealing means while selectively permitting translational and/or rotational motion of the fuel pin assembly relative to the operational axis of said first transport means. an enclosure containing means for inserting a charge of fuel pellets into each fuel pin assembly through the enlarged funnel mounted thereto while said one end of the fuel pin assembly is held at a loading position within the enclosure by said first transport means; and funnel removal means contained within the enclosure in the path of each fuel pin assembly as it is shifted parallel to the operational axis of the first trolley following its receipt of a charge of fuel pellets for engaging and stripping the enlarged funnel from said one end of the fuel assembly; means contained within the enclosure for wiping any contaminants from the surfaces of said one end of each fuel fin assembly after stripping of the enlarged funnel mounted thereto; and means for applying a cap to said one end of each fuel pin assembly after wiping thereof. 2. The automated loading apparatus of claim 1, wherein said transfer means comprises: 3. The automated loading apparatus of claim 1, wherein the fuel pin handling means comprises: 4. The automated loading apparatus of claim 1, wherein the fuel pin handling means comprises: 5. The automated loading apparatus of claim 1, wherein the fuel pin handling means comprises: 6. In an automated loading apparatus for nuclear reactor fuel pins: 7. The automated loading apparatus of claim 6 wherein the fuel pin loading station comprises: 8. The automated loading apparatus of claim 6 wherein the fuel pin loading station comprises: 9. The automated loading apparatus of claim 6 wherein the fuel pin loading station comprises: |
summary | ||
description | This invention was developed under Contract DE-AC04-94AL85000 between Sandia Corporation and the U.S. Department of Energy. The U.S. Government has certain rights in this invention. Embodiments of the present invention are in the field of radioactive nuclei capture and immobilization, and, more particularly, relate to methods of capturing and immobilizing radioactive nuclei with metal fluorite-based inorganic materials. One of the great concerns in the nuclear power field throughout the world is the safe disposal and isolation of used fuels from reactors or waste streams from reprocessing plants. In particular, entrapment of highly mobile radionuclides such as iodine (129I) and technetium (99Tc) produced from a fission process and subsequent capturing and immobilization of these radionuclides in an appropriate waste form is a great technical challenge because of the high mobility of these radionuclides and the difficulty in incorporating them into any existing waste forms such as glass, ceramics, and grout. Iodine (129I) and technetium (99Tc) both have long half-lives and can exist as gaseous or anionic species that are highly soluble and poorly sorbed (absorbed or adsorbed) by natural materials. Waste forms are probably the only engineered barrier to limit their release into a human-accessible environment after their disposal. Furthermore, a majority (>99%) of 129I will enter into the dissolver off-gas stream during fuel reprocessing. It is thus highly desirable to develop a material that can effectively entrap gaseous iodine during the off-gas treatment. Thus, further advancements are needed in the area of radioactive nuclei capture and immobilization. Embodiments include methods of capturing and immobilizing radioactive nuclei with metal fluorite-based inorganic materials. In an embodiment, a method of capturing and immobilizing radioactive nuclei includes flowing a gas stream through an exhaust apparatus. The exhaust apparatus includes a metal fluorite-based inorganic material. The gas stream includes a radioactive species. The radioactive species is removed from the gas stream by adsorbing the radioactive species to the metal fluorite-based inorganic material of the exhaust apparatus. In another embodiment, a method of synthesizing a metal fluorite-based inorganic material includes charging a reaction vessel with a solution having an aluminum precursor and a fluorine precursor. A precipitating agent is added to the solution having the aluminum precursor and the fluorine precursor. An aluminum (Al)-rich fluorite precipitate is then isolated. In another embodiment, a nanoporous material is composed of a metal fluorite-based inorganic compound having a surface area greater than approximately 1 square meter per gram (m2/g). Methods of capturing and immobilizing radioactive nuclei with metal fluorite-based inorganic materials are described herein. In the following description, numerous specific details are set forth, such as reagents suitable for aluminum fluorite synthesis, in order to provide a thorough understanding of embodiments of the present invention. It will be apparent to one skilled in the art that embodiments of the present invention may be practiced without these specific details. In other instances, well-known processing operations, such as sample preparation, are not described in detail in order to not unnecessarily obscure embodiments of the present invention. Furthermore, it is to be understood that the various embodiments shown in the figures are illustrative representations and are not necessarily drawn to scale. Embodiments of the present invention may relate to methods and compositions useful in capture and possible disposal of radionuclides, particularly highly volatile or mobile radionuclides, as well as other hazardous materials. One or more embodiments provide a new set of getter materials and possible waste forms for entrapping or capturing and immobilizing radionuclides, especially 129I and 99Tc, as well as other hazardous materials. High performance inorganic solid adsorbents for capturing gaseous radionuclides may be required in multiple process operations of an advanced fuel cycle. For example, a vented fuel pellet or getter concept has been proposed to extend fuel burn-ups. Such a concept may be intentionally designed to enhance fission gas release from the fuel pellet and to sequester the gaseous fission products in the fuel's plenum using an adsorbent material. In used fuel reprocessing, efforts have been made to develop getter materials for capturing fission gases released from fuel dissolution and oxidation processes. In an embodiment, such materials are also useful for radionuclide containment in an event of nuclear plant accident. Furthermore, in an embodiment, at the backend of the fuel cycle, radionuclide getter materials are also utilized for either long-term geologic waste disposal or interim used fuel storage. In accordance with an embodiment of the present invention, fluorite-based inorganic solid adsorbents, either natural or synthetic, are used for off-gas treatment. Such adsorbents may perform far better than conventional oxide-based materials with respect to gaseous radionuclide sorption. In one such embodiment, the sorption affinity of fluorite-based inorganic solid adsorbents for gaseous iodine is 2 to 3 orders of magnitude stronger than the conventional oxide-based adsorbents. Described herein are chemical precipitation and thermal decomposition routes for the synthesis of high surface area nanoporous metal fluorite materials. In an embodiment, the synthesized fluorite materials exhibit excellent iodine sorption capabilities. Furthermore, these materials may have a wide range of applications in advanced nuclear fuel cycles as well as chemical industries. Given their high chemical stability, the adsorbent materials described in embodiments herein may be ideal for use in extreme chemical and physical environments, for example, in acidic elevated temperature environments. The sorption capability of an adsorbent material may, to a large extent, be determined by the interaction of an adsorbate with the surface terminating atoms of the material. Conventional solid inorganic adsorbents for radionuclide capture are exclusively based on oxide compounds, in which the surface terminating atoms are overwhelmingly oxygen atoms. Using a Grand Canonical Monte Carlo (GCMC) simulation, in accordance with an embodiment of the present invention, we herein describe that replacing (at least conceptually) surface oxygen atoms with fluorine on γ-aluminum oxide significantly improves the capability of the material for iodine sorption (see, e.g., description of FIG. 1B below). Such improvement may be due to the additional polarization of iodine molecules induced by fluorine atoms. The simulation results are consistent with actual experimental data presented below. In an embodiment, based both the simulated and the experimental data, we describe herein that non-oxide materials can be used as adsorbents for gaseous radionuclides, with sorption capabilities significantly improved over conventional oxide-based adsorbents. FIGS. 1A and 1B depict a Grand Canonical Monte Carlo (GCMC) simulation of iodine gas adsorption on a modified γ-alumina (Al2O3) surface, in accordance with an embodiment of the present invention. Specifically, FIG. 1A illustrates a plane view of a 60% fluorinated alumina surface model 100 for use in a GCMC simulation of iodine gas adsorption. The balls with the lightest shading represent fluorine (F) atoms, the balls with the darkest shading represent oxygen (O) atoms, and the balls with the intermediate shading represent aluminum (Al) atoms. FIG. 1B is a plot 102 of iodine adsorption onto unmodified (line 104) or fluorinated (line 106) alumina surfaces predicted from the GCMC simulation. As mentioned briefly above, and in accordance with one or more embodiments of the present invention, non-oxide materials (e.g., materials with fluorinated surfaces) show marked improvement versus conventional oxide-based materials with respect to radionuclide sorption. In one such embodiment, the sorption capability of a fluorite material for gaseous iodine (as normalized to surface area) is approximately 1000 times greater than that for conventional oxide materials. As an example, FIG. 1C is a plot 110 showing the measurements of surface normalized iodine sorption capability of metal fluorite (e.g., CaF) or hybrid fluorine/oxygen surfaces (e.g., Al—O—F) in comparison with conventional oxide-based adsorbents (e.g., Al—O), in accordance with an embodiment of the present invention. In consideration of plot 110, in an embodiment, one or both of two phenomena dictate the marked improvement of adsorption: (a) an increased surface area (although surface area is normalized in plot 110) and (b) binding energy (e.g., increased affinity using F-terminating atoms in place of O-terminating atoms). In an embodiment materials suitable for improved iodine (or other radionuclide) adsorption are synthesized in consideration of one or both of two key factors: (1) the demonstration described herein of high sorption affinity of metal fluorites for iodine-129 (e.g., as shown in plot 110 of FIG. 1C), and (2) techniques described herein for synthesizing high surface area nanoporous materials. In one such embodiment, the combination of both high sorption affinity and high specific surface area renders the resulting nanoporous metal fluorite materials as excellent radionuclide adsorbents. In an aspect of the present invention, radioactive nuclei are captured and immobilized with a fluorite-based material. For example, FIG. 2 depicts a flowchart 200 representing an exemplary series of operations in a method of capturing and immobilizing radioactive nuclei with a metal fluorite-based inorganic material, in accordance with an embodiment of the present invention. FIGS. 3A-3C illustrate schematic representations of various operations of the flowchart of FIG. 2, also in accordance with an embodiment of the present invention. Referring to operation 202 of flowchart 200, and to corresponding FIG. 3A, a method of capturing and immobilizing radioactive nuclei with a metal fluorite-based inorganic material includes flowing a gas stream through an exhaust apparatus including or charged with a metal fluorite-based inorganic material (either natural or synthetic), the gas stream carrying a radioactive species. In a specific example, a gas stream 302 carrying a radioactive species 304 is flowed through an exhaust apparatus 306 charged with a metal fluorite-based inorganic material 308. In an embodiment, flowing the gas stream 302 through the exhaust apparatus 306 includes flowing through a nanoporous metal fluorite-based inorganic material (e.g., a nanoporous example of material 308). In an embodiment, flowing the gas stream 302 through the exhaust apparatus 306 includes flowing through a natural or a synthetic aluminum (Al)-rich fluorite (e.g., examples of material 308). In one such embodiment, the flowing is performed through a synthetic aluminum (Al)-rich fluorite with a surface area greater than approximately 1 square meter per gram (m2/g). In an embodiment, a natural fluorite material, such as naturally occurring CaF2, is used. Referring to operation 204 of flowchart 200, and to corresponding FIG. 3B, the method also includes removing, from the gas stream, the radioactive species by adsorbing the radioactive species to the metal fluorite-based inorganic material of the exhaust apparatus. In a specific example, the gas stream 302′ exits the exhaust apparatus 306 with the radioactive species 304 adsorbed to the metal fluorite-based inorganic material 308. In an embodiment, removing the radioactive species 304 from the gas stream 302 includes removing a nuclei such as, but not limited to, 129I. Referring to FIG. 3C, the gas stream 302/302′ is terminated with the radioactive species 304 still adsorbed to the metal fluorite-based inorganic material 308 in the exhaust apparatus 306. In one such embodiment, referring to optional operation 206 of flowchart 200, the method further includes disposing of the metal fluorite-based inorganic material 308 having the radioactive species 304 adsorbed thereon. In another such embodiment, referring to optional operation 208 of flowchart 200, the method further includes releasing the radioactive species 304 from the metal fluorite-based inorganic material 308. For example, in a specific embodiment, the radioactive species 304 is released from the metal fluorite-based inorganic material 308 into a medium such as, but not limited to, a second gas stream, a liquid extraction medium, or a solid extraction medium for chemical separation or concentration. In an aspect of the present invention, a nanoporous-structured material may be engineered through thermal decomposition or direct precipitation (e.g., by sol-gel methods). For example, FIG. 4 depicts a flowchart 400 representing an exemplary series of operations in a method of synthesizing a metal fluorite-based inorganic material, in accordance with an embodiment of the present invention. FIGS. 5A-5D illustrate schematic representations of various operations of the flowchart of FIG. 4, also in accordance with an embodiment of the present invention. Referring to operation 402 of flowchart 400, and to corresponding FIG. 5A, a method of synthesizing a metal fluorite-based inorganic material includes charging a reaction vessel with a solution including an aluminum precursor and a fluorine precursor. Specifically, a reaction vessel 500 is charged with an aluminum precursor 502 and a fluorine precursor 504, e.g., in the form of a solution 506. In an embodiment, charging the reaction vessel 500 with the solution 506 including the aluminum precursor 502 and the fluorine precursor 504 includes charging the reaction vessel 500 with aluminum trichloride (AlCl3) and a trifluoracetate. In one such embodiment, the trifluoracetate is a neutralized product of trifluoracetic acid. In one embodiment, charging the reaction vessel 500 with AlCl3 and the trifluoracetate further includes using a water/ethanol solution 506. In an embodiment, charging the reaction vessel 500 with the solution including the aluminum precursor and the fluorine precursor includes using an aluminum:fluorine molar ratio of approximately 1:1 or 1:3, examples of which are described in more detail below. Referring to operation 404 of flowchart 400, and to corresponding FIG. 5B, the method also includes adding a precipitating agent to the solution including the aluminum precursor and the fluorine precursor. Specifically, the reaction vessel 500 charged with the aluminum precursor 502 and the fluorine precursor 504 has a precipitating agent 508 added thereto. In an embodiment, adding the precipitating agent 508 includes adding polyethylene oxide. Upon adding the precipitating agent 508, referring to FIG. 5C, the reaction vessel 500 has an aluminum (Al)-rich fluorite 510 precipitated therein. Referring to operation 406 of flowchart 400, and to corresponding FIG. 5D, the method also includes isolating an aluminum (Al)-rich fluorite precipitate. Specifically, an isolated aluminum (Al)-rich fluorite precipitate 512 may be collected on a filter 514. In an embodiment, isolating the Al-rich fluorite precipitate 512 includes collecting the Al-rich fluorite precipitate 510 of FIG. 5C on the filter 514, subsequently washing the Al-rich fluorite precipitate with deionized water, and subsequently heating the Al-rich fluorite precipitate. In one such embodiment, heating the Al-rich fluorite precipitate 510 of FIG. 5C dehydrates the Al-rich fluorite precipitate to provide the isolated aluminum (Al)-rich fluorite precipitate 512. In an embodiment, referring to optional operation 408 of flowchart 400, the method may also include, prior to adding the precipitating agent 508, adding a block co-polymer to the reaction vessel 500. In one such embodiment, isolating the Al-rich fluorite precipitate 512 includes thermalizing the precipitate 510 of FIG. 5C (which, in this embodiment, would include the block co-polymer encapsulated therein) to remove the block co-polymer. In an embodiment, the block co-polymer is used to fabricate porosity (and hence increased surface area) into a synthetic metal fluorite-based inorganic material. In a specific embodiment, the block co-polymer is a material such as, but not limited to poly(ethylene glycol)-block-poly(propylene glycol)-block-poly(ethylene glycol), known commercially as P123. As mentioned briefly above, thermal decomposition may be used as part of the fabrication or synthesis of a nanoporous aluminum fluorite material. In an embodiment, such a process is based on the initial precipitation of an Al—OOCCF3 precursor (e.g., may be the precipitate 510 of FIG. 5C) disposed around an organic polymer framework, e.g., such as a block co-polymer framework. In one such embodiment, the precursor is decomposed at approximately 350° C. Subsequently, the polymer is removed via calcination at approximately 550° C. The end result may be to leave behind a fluoride rich, high surface area Al-based precipitate. In specific set of experimental embodiments, solutions for Products 1-5 below were all prepared on a hot plate set at approximately 60° C. with constant stirring. Approximately 12 grams of a block co-polymer was dissolved in approximately 72 milliliters of ethanol (EtOH) (Solution 1) to provide a nanoporous template. Approximately 34 grams of aluminum trichloride (AlCl3) was dissolved in approximately 45 milliliters of 1:1 EtOH:H2O solution (Solution 2). Solution 2 was added to Solution 1 to provide a mixed Solution 3. Approximately 40 grams of polyethylene oxide (PO) was added to Solution 3. The precipitate was aged in an approximately 70-90° C. oven overnight, or longer as necessary for solvent removal. To the above basic operations, trifluoroacetic acid (3FAc) was added to provide a stoichiometric ratio of Al:F of approximately 1:3, and was added at various stages and in various forms to provide Products 1-5. Product 1: un-neutralized 3FAc was added to Solution 1. Product 2: neutralized 3FAc (with a pH of approximately 5) was added to Solution 1. Product 3: neutralized 3FAc was added to Solution 3. Product 4: neutralized 3FAc was added to Solution 2 (with an immediate precipitation operation and no other operations following). Product 5: identical to Product 2, but the PO was not added, and the precipitation proceeded via solvent evaporation in an oven. Once synthesized, all Products 1-5 were subjected to heating in a muffle furnace for approximately 2 hours at approximately 350° C., then for approximately 2 hours at approximately 600° C. As also mentioned briefly above, direct precipitation may be used as part of the fabrication or synthesis of a nanoporous aluminum fluorite material. In an embodiment, direct precipitation is based on making an Al—F precursor solution, followed by precipitation of an Al—F rich complex. The precipitation results from solvent insolubility with and without a polymer framework. In specific set of experimental embodiments, precursor solutions for Products 6-12 below were all prepared by dissolving approximately 50 grams of AlCl3 in an approximately 70:30 EtOH:H2O mixture (e.g., approximately 100 milliliters) and heated to near boiling. Either approximately 8 grams (Al:F=1:1), or approximately 35 grams (Al:F=1:3) of sodium fluoride (NaF) was dissolved in water (e.g., approximately 200 milliliters) and a sufficient amount of concentrated hydrochloric acid (HCl) to provide a pH less than approximately 2. In both cases, a slurry was created. The NaF slurry was then combined with the AlCl3 solution. The entire solution was then brought to a boil. Upon clarification of the solution, the solution was reduced in volume until precipitates began to form. A relatively small amount of water was added to re-dissolve the precipitates. The resulting parent solution was poured into various mixtures of EtOH and H2O. In all but one of the above mixtures, approximately 16 grams of block co-polymer had been added. The above process provided Products 6-12 listed in table 600 of FIG. 6A (table 600 is a material synthesis matrix for direct precipitation). All precipitates were then collected through solvent evaporation, dried in an approximately 90° C. oven overnight and subsequently calcined at approximately 600° C. for approximately 4 hours. FIG. 6B includes a table 602 of material characterization and sorption testing for Products 1-12 (synthesized metal fluorite materials) from above, in accordance with an embodiment of the present invention. Referring to table 602, all samples were characterized using BET surface area measurements (S. Brunauer, P. H. Emmett, and E. Teller, J. Am. Chem., 1938, 60, 309), which is a surface area technique based on monolayer adsorption of nitrogen gas, and gaseous iodine sorption studies. Results for a natural CaF2 mineral are also provided as a point of reference. These results tabulated in table 602 exemplify the importance of both the ordering of the operations of the synthesis as well as the solvent choice. In an embodiment, for the thermal decomposition materials (Products 1-5), there is a clear advantage to adding the neutralized 3FAc to the parent Al-polymer solution. Adding the neutralized 3FAc at other points provides significantly impacted results. In another embodiment, for the direct precipitation materials (Products 6-12), there is a clear advantage to using pure ethanol to precipitate the materials. For example, Products 6, 8, 9, and 10 all perform better than the CaF2 reference material, which is a natural CaF2, and all were prepared in ethanol. Other variables (such as, but not limited to, presence of polymer, stoichiometric ratios, temperature) play only a secondary role. It appears that the addition of fluorine in the various synthesis procedures increases the chemical affinity of iodine to the surface of the resulting synthetic compound. Despite the lower surface areas relative to the Al-O material, the fluorinated compounds still outperform existing oxide-based adsorbents in total iodine removal. As mentioned above, the high performance inorganic solid adsorbents described herein may be applied to capturing gaseous radionuclides in multiple process operations of an advanced fuel cycle from novel fuel concepts, to off-gas treatment of used fuel reprocessing, and to nuclear waste disposal. The applications may potentially be extended to the capture, separation, and immobilization of non-radioactive chemicals. Given their high chemical stability, the adsorbent materials described herein can, in an embodiment, be ideal for use in extreme chemical and physical environments, for example, in acidic elevated temperature environments. Although a variety of surface areas may be fabricated for an aluminum-rich fluorite species and may be suitable for radionuclide capture, investigations thus far indicate that the greater the surface area, the greater the immobilization ability. In an embodiment, a nanoporous material is composed of a metal fluorite-based inorganic compound having a surface area greater than approximately 1 square meter per gram (m2/g). In one such embodiment, the metal fluorite-based inorganic compound is an aluminum (Al)-rich fluorite species. In another such embodiment, the surface area is greater than approximately 5 square meters per gram (m2/g). In yet another such embodiment, the surface area is approximately in the range of 5-100 square meters per gram (m2/g). Embodiments of the present invention may be based on the capture and encapsulation of radionuclides, particularly highly mobile radionuclides, including radioactive isotopes of iodine and technecium. For example, in one embodiment, methods described herein are used to render hazardous materials less dangerous. Embodiments may also include generating getter materials and the resultant waste forms, and compositions of matter of the getter materials and the resultant waste forms. Overall, then, metal fluorite inorganic materials, such as nanoporous aluminum fluorite, represent a class of highly efficient materials for uptake of gaseous iodine (and other volatile radionuclides) even at an elevated temperature. Such materials may, in an embodiment, be converted to a durable waste form in which the sorbed iodine is effectively encapsulated. The technology of the invention may be applied to the treatment of other radionuclides and even non-radioactive hazardous materials. Other radionuclides for which the present invention may be useful include, but are not limited to, iodine (including 129I) (both in gaseous and anionic forms such as I− and IO3−), technetium (including 99Tc) (e.g., in anionic forms such as TcO4−), and radioactive isotopes of Pu, Am, U, Th, Np, Se, Cs, Sr, C, CI, H, Xe and Kr in the forms of molecular species, cations and anions. Non-radioactive hazardous materials that may be disposed of according to one or more embodiments of the invention include heavy metals such as Pb and Cd in the form of cations. It is to be understood that metal fluorite species suitable for capturing and immobilizing such ions may include metal fluorite species other than the aluminum rich versions described in details herein; and they may, in one or more embodiments, include the fluorite compounds of alkaline earth metals (e.g., Ca), transition metals (e.g., Mn, Zr, Re), or rare earth metals. Thus, methods of capturing and immobilizing radioactive nuclei with metal fluorite-based inorganic materials have been disclosed. In accordance with an embodiment of the present invention, a method of capturing and immobilizing radioactive nuclei includes flowing a gas stream through an exhaust apparatus. The exhaust apparatus includes a metal fluorite-based inorganic material. The gas stream includes a radioactive species. The radioactive species is removed from the gas stream by adsorbing the radioactive species to the metal fluorite-based inorganic material of the exhaust apparatus. In one embodiment, the method further includes disposing of the metal fluorite-based inorganic material having the radioactive species adsorbed thereon. In one embodiment, the method further includes releasing the radioactive species from the metal fluorite-based inorganic material into a medium such as, but not limited to, a second gas stream, a liquid extraction medium, or a solid extraction medium for chemical separation or concentration. |
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summary | ||
abstract | A scattered radiation grid of a CT detector is disclosed and includes a plurality of detector elements arranged in multiple cells in the phi direction and in the z direction of a CT system, having a plurality of free passage channels arranged to correspond to the detector elements, and walls fully enclosing the free passage channels at the longitudinal sides thereof. According to an embodiment of the invention, the walls of the scattered radiation grid are produced using a 3D screen-printing method. |
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048266540 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Hereinafter, some preferred embodiments of the invention will be described in detail, referring to the drawings. Embodiment 1 An embodiment of a fuel assembly for use in a boiling water reactor according to the present invention will be described referring to FIGS. 1 and 2. In FIG. 1, the fuel assembly 1 comprises a plurality of fuel rods fixed at both end portions by upper and lower tie plates 3, 4, water rods 6 disposed among the fuel rods 2, a plurality of fuel spacers 5 disposed in an axial direction to keep the fuel rods 2 and the water rods 6 in a predetermined laterally spaced relation, and a channel box 7, fitted to the upper tie plate 3, extending downwards so as to encompass the bundle of the fuel rods 2 and mounted on the lower tie plate 4. In FIG. 2, a space region 8 is formed at the center on the cross-sectional plane of the fuel assembly 1. The lower end of this space region 8 is positioned on the upper surface of the lower tie plate 4, and its upper end on the lower surface of the upper tie plate 3. The water rods 6 are disposed radially so as to extend outwards from the space region 8 every two of the water rods 6 (towards the four sides of the channel box 7). The space region 8 is encompassed by four water rods 6 and is represented by a square of dotted line in FIG. 2 for convenience' sake. In practice, however, any structure expressed by the square of the dotted line does not exist. When the fuel assemblies 1 are outside the reactor, the space region 8 is a mere space but when they are loaded in the reactor, it serves as a passage of cooling water. Four water rods 6 are directly adjacent to the space region 8, that is, the four water rods 6 face to the space region 8. The size of the space region 8 is such that one fuel rod 2 or one water rod 6 can be disposed. The fuel rods 2, the water rods 6 and the space region 8 are arranged in a grid form as shown in FIG. 2. Four kinds of fuel rods 21-24 are used as the fuel rods 2 as shown in Table 1. Though the fuel rods 21-23 do not contain gadolinium which is an burnable poison the fuel rod 24 does. The enrichment is 1.8 wt % for the fuel rod 21, 2.5 wt % for the fuel rod 22 and 3.3 wt % for the fuel rods 23 and 24. TABLE 1 ______________________________________ Fuel Rod No. 21 22 23 24 ______________________________________ enrichment (wt %) 1.8 2.5 3.3 3.3 gadolinium (wt %) -- -- -- 3.5 number of rods 4 8 52 8 ______________________________________ After the fuel assembly 1 of this embodiment 1 is loaded into the core of a boiling water reactor, the reactor is started to operate. During the operation of the boiling water reactor, cooling water flows into the fuel assembly 1 from the lower tie plate 4. This cooling water rises through the gaps formed among the fuel rods 2 and through the water rods 6 and the space region 8. During this elevation process, the cooling water is heated while it cools the fuel rods 2 and part of the cooling water changes to a vapor. Since the cooling water flows into the water rods 6 in the proximity of the upper surface of the lower tie plate 4, hardly any vapor is contained in the cooling water flowing inside the water rods 6. The vapor rises through the gaps among the fuel rods 2 and through the space region 8. In this embodiment, since the water rods 6 are adjacent to the space region 8, the void fraction of the space region 8 is lower than the mean void fraction of the fuel assembly 1 for the following reason. FIG. 3 shows the influences of the difference of relative position between the space region 8 and the water rod 6 upon the number of hydrogen atoms in the space region 8 by the use of the mean void fraction of the fuel assembly as a parameter. The reference value uses the number of hydrogen atoms in a unit cell for the water rod when the mean void fraction of the fuel assembly is 0%. The term "unit cell" hereby means one of square defined by orthogonally intersecting grid plates in a grid-like fuel spacer constituted by mutually crossing and different grid plates disposed in x and y directions, that is, the unit cell corresponds to a space 8 defined by the dotted line. A characteristic line 11 represent characteristics when the fuel rods 2 are disposed in all the unit cells adjacent to the space region 8 and a line 12 represent characteristics when a half of the unit cells adjacent to the space region 8, that is, 4, are filled with water rods 6 as shown in FIG. 2. The number of hydrogen atoms in the space region 8 at a low void fraction is greater than that in the water rod 6 represented by characteristics 13 by the number that corresponds to the absence of a hollow pipe of the water rod 6, irrespective of the relative position to the water rod. Under the high void fraction state where nonhomogeneity in both the axial and radial directions in the fuel assembly becomes the greatest, however, the number of hydrogen atoms in the space region 8 in which the fuel rods are disposed in all of the adjacent unit cells drops remarkably when compared with the number of hydrogen atoms of the water rod 6 (see the characteristics 11). On the other hand, in the space region 8 in which the water rods 6 are disposed in the half of the adjacent unit cells (see FIG. 2), the number of hydrogen atoms at a void fraction of up to 40% is greater than that of the water rod 6, and even at a void fraction higher than 40% the number of hydrogen atoms does not drastically drop when compared with that of the water rod 6. This results from the phenomenon that in the fuel assembly 1 having the water rods 6 disposed therein, the distribution of the void fractions inside the channel box 7 changes greatly depending upon the position. In other words, when the fuel rods 2 are disposed in all the unit cells adjacent to the space region 8, the void fraction of the space region 8 increases to a level substantially equal to the mean void fraction of the fuel assembly, and when the water rods 6 are disposed in the unit cells adjacent to the space region 8, the void fraction of the space region 8 becomes by far smaller than the mean void fraction of the fuel assembly 1 in the same way as in the region inside the water rod 6. In FIG. 3, when the mean void fraction of the fuel assembly 1 is 70%, the number of hydrogen atoms of the characteristics 12 increases by about 70% in comparison with that of the characteristics 11. This is a decrease of only about 10% in comparison with the case of the water rod 6. Furthermore, when the reducing effect of the void fraction in the water rods 6 adjacent to the space region 8 due to disposition of the space region 8 is taken into consideration, the space region 8 of the characteristics 12 has the number of hydrogen atoms substantially equal to that of the water rod 6. FIG. 4 shows the relation between the number of hydrogen atoms and reactivity in order to represent the nuclear effect of the characteristics 11 and 12. For example, the reactivity at a mean void fraction of 70% of the assembly rises by about 0.04% .DELTA.K.infin. owing to the change from the point A to the point B in FIG. 3. In other words, it can be understood that when the water rods are disposed in the unit cells adjacent to the space region 8, the space region 8 exhibits substantially the same nuclear function as the water rods 6 at a high void fraction. At a low void fraction, the reactivity of the space region 8 increases by about 0.05% .DELTA.K.infin. (per space region for one unit cell) in comparison with the water rod 6 due to the increase of the homogenization effect in the radial direction of the fuel assembly resulting from the increase in the number of hydrogen atoms and due to the decrease of the neutron absorption quantity resulting from the absence of the hollow pipe. In this manner, since the water rods 6 and the space region 8 are adjacent to one another in the fuel assembly 1 of this embodiment, the reactivity can be improved and hence, fuel cycle economy can be improved, too. On the other hand, the pressure drop reducing effect of the space region 8 is not dependent upon its relative position with the water rod 6 but is dependent upon the area of the space region 8. FIG. 5 shows a relation between the area of the space region 8 and the pressure drop. In this embodiment, as the cross-sectional area of the pipes disposed in the fuel assembly and the wetted area decrease when compared with the case where the water rods 6 are disposed inside the space region 8, too, the pressure drop can be reduced. Obviously, the fuel assembly 1 of this embodiment is superior to the fuel assembly using only the water rods 6 from the aspects of the pressure drop and economical efficiency of fuel. Though only reduction of the pressure drop has been described as the effect of increasing the flow path area by the space region 8, the following can also be accomplished by limiting the reduction of the pressure drop when there is a sufficient margin for the pressure drop: (i) promotion of uniformity of fuel assembly structure by the increase in the number of water rods; and (ii) extension of life of fuel assembly by increasing uranium loading quantity. The effect described above is not limited to the case where the water rods 6 are disposed in the half of the unit cells adjacent to the space region 8 as in the fuel assembly 1 shown in FIG. 2. FIG. 6 shows a relation between the number of water rods 6 occupying the unit cells adjacent to the space region 8 and the number of hydrogen atoms in the space region 8. It can be understood from FIG. 6 that if the water rod 6 is disposed in at least one of adjacent unit cells, the effect described above can be obtained. The enrichment distribution which shows the effect of this embodiment (shown in FIG. 2) in comparison with the fuel assembly 30 shown in FIG. 7(A), which is obtained by replacing the space region 8 of this embodiment by the water rod 6, is the same as that of this embodiment. The void fraction of the flow passage in the region A shown in FIG. 7(B) is compared with respect to this embodiment and the fuel assembly 30 in FIG. 7(A). The result at the upper end of the fuel assembly is shown in Table 2. In this embodiment, the void fraction drops not only in the region f which is the space region 8 but also in the regions d and e which are adjacent unit cells. As a result, the number of hydrogen atoms in the region A is substantially equal in the fuel assembly 1 of this embodiment and in the fuel assembly 30 in FIG. 7(A). TABLE 2 ______________________________________ Region Embodiment 1 FIG. 7(A) ______________________________________ a 67.4% 67.9% b 62.8 62.5 c 60.1 59.9 d 55.4 56.4 e 51.0 53.2 f 45.4 49.6 ______________________________________ Therefore, the neutron infinite multiplication factor is equal to that of the fuel assembly 30 at the upper part of the fuel assembly 1 and is higher by about 0.05 .DELTA.K.infin. at its lower part than that of the fuel assembly 30. On the other hand, the pressure drop in the fuel assembly 1 is lower by about 3% (about 0.021 Kg/cm.sub.2) than that of the fuel assembly 30. In the embodiment described above, the arrangement of the fuel rods and the like is 9 rows by 9 columns inside the fuel assembly by taking into consideration the fact that since a large number of water rods are disposed, the number of fuel rods decreases and the linear output density increases. However, the concept of this embodiment can of course be applied to fuel assemblies having the arrangement of 10 rows by 10 columns and 11 rows by 11 columns. In recent years, a fuel assembly capable of increasing the linear power density has been developed by bonding a Cu or Zr thin film on the inner wall of a cladding pipe to decrease the interaction between fuel pellets and the cladding pipe. In the case of such a fuel assembly, the arrangement of the fuel rods of 8 rows by 8 columns can be employed. A fuel assembly 35 shown in FIG. 8 is another application example of Embodiment 1. The fuel assembly 35 increases the mean enrichment of the fuel assembly 1 in Embodiment 1 by about 4 wt %. As fuel rods 3 in this fuel assembly 35, four kinds of fuel rods 31-34 as shown in Table 3 are used. TABLE 3 ______________________________________ Fuel rod number 31 32 33 34 ______________________________________ enrichment (wt %) 3.0 3.6 4.5 3.3 gadolinium (wt %) -- -- -- 4.5 number of rods 4 8 46 14 ______________________________________ If the enrichment is increased in a conventional fuel assembly, the moderation effect of neutrons drops so that the worth of the control rod decreases and the void coefficient (absolute value) increases. On the other hand, when a large number of water rods 6 are disposed to obtain a synergistic effect with the space region 8 as in this embodiment, hardly any increase in the pressure drop in the fuel assembly occurs and in addition, the moderation effect of neutrons can be increased. As a result, fuel cycle economy can be improved by about 7% in comparison with the conventional fuel assembly and a surplus linear power density at the stop of reactor can be limited to substantially the same level as the one in the case where the enrichment is about 3 wt %. Embodiment 2 FIG. 9(A) is a cross-sectional view of a fuel assembly 36 in another embodiment of the invention. In the fuel assembly 36 of this embodiment, the positions of the water rods 6 are different from those of the fuel assembly 1 in Embodiment 1. The fuel rods 2 are the same as in the Embodiment 1, using the fuel rods 21-24. In this embodiment, the water rods 6 are disposed in all the unit cells adjacent to the space region 8. Since the water rods 6 and the space region 8 are adjacent to one another in this embodiment, too, the same function as that of Embodiment 1 develops. The void fraction of the flow passage of the region B of FIG. 9(B) in this fuel assembly 36 is compared with fuel assembly obtained by replacing the space region 8 of this fuel assembly 36 by the water rods 6. The effect at the upper end of the fuel assembly 36 is shown in Table 4 and the effect is greater than in Embodiment 1. At the lower part of the fuel assembly 36, that is, in the low void region, however, the improvement in reactivity is less than at the lower part of the fuel assembly 1. On the other hand, the pressure drop in the fuel assembly 36 decreases by about 3% (about 0.021 Kg/cm.sup.2) in comparison with the fuel assembly 30 shown in FIG. 7(A). TABLE 4 ______________________________________ Fuel Assembly having water rods in Region Embodiment 2 place of space region of FIG. 9(A) ______________________________________ d 55.7% 56.7% e 50.5 53.1 f 42.5 48.1 ______________________________________ Embodiment 3 FIG. 10 is a cross-sectional view of a fuel assembly 37 in accordance with still another embodiment of the present invention. This embodiment has a space region 15 which is five times greater than the space region of Embodiment 1 in order to improve the reducing effect of the pressure drop in the fuel assembly. The water rods 6 are disposed in four of unit cells 16 adjacent to the space region 15. The void fraction of the flow passage in the region A shown in FIG. 7(B) is compared between the fuel assembly of this embodiment and that of FIG. 7(A) in the same way as in Embodiment 1. The result at the upper end of the fuel assembly is shown in Table 5. Though the effect is smaller than in Embodiment 1, the void fraction is lower. At a low void fraction, since the neutron absorption quantity by the hollow pipe decreases, the neutron infinite multiplication factor is improved by about 0.2% .DELTA.K.infin. than that of fuel assembly shown in FIG. 7(A). TABLE 5 ______________________________________ Region Embodiment 3 FIG. 7(A) ______________________________________ a 68.4% 67.9% b 62.5 62.5 c 59.2 59.9 d 54.9 56.4 e 51.4 53.2 f 48.3 49.6 ______________________________________ As to the pressure drop in the fuel assembly, on the other hand, it decreases by about 13% (about 0.084 Kg/cm.sup.2) in comparison with that in the fuel assembly shown in FIG. 7(A). Though each of the embodiments described above uses the water rod as the moderator rod, the same effect can be obtained when a moderator rod having sealed therein a solid moderator material having a high hydrogen density and a small neutron absorption cross-section is used in place of the water rod. When a burnable poison is mixed in the solid moderator rod, the void fraction distribution of the fuel assembly becomes further greater. In accordance with the present invention, since the space region and the moderator rods are disposed adjacent to each other, the reactivity of the fuel assembly can be improved, the fuel cycle economy can be improved and the sectional area of the coolant flow passage can be increased so that the pressure drop in the fuel assembly can be reduced. |
050892100 | claims | 1. In a boiling water reactor core having a plurality of vertically upstanding fuel bundles; each fuel bundle containing longitudinally extending sealed rods with fissile material therein; a surrounding channel for containing water flow within each said fuel bundle from water exterior of said fuel bundle; water moderator flowing in the confining channel from the bottom of said fuel bundle to the top of said fuel bundle for producing steam and moderating neutrons to a thermal energy state for producing continuing nuclear reaction in said fissile material; said water moderator exterior of said confining channels for defining a core bypass zone having relatively high concentrations of water for moderation of neutrons from high energy state neutrons to a thermal energy state for the continuation of said nuclear reaction; a plurality of said fuel rods including a component of fissile material including recovered plutonium and uranium distributed over an axial extent of said fuel assembly; at least some of said fuel rods containing a component of neutron absorbing material for controlling excess reactivity imparted by said recovered plutonium; the improvement in said distribution of fissile material and neutron absorbing material comprising: said fissile material including a mixture of uranium and recovered plutonium in rods of said fuel bundle at locations other than the corners of said fuel bundle; and, neutron absorbing material being located in rods of said fuel bundle at rod locations adjacent the corners of said fuel bundles whereby said neutron absorbing material has decreased shielding from said plutonium and maximum exposure to thermal neutrons for shaping said cold reactivity shutdown zone in said fuel bundle. at least some of said fuel rods containing a component of neutron absorbing material for controlling excess reactivity imparted by said recovered plutonium; the improvement in the distribution of fissile material and neutron absorbing material comprising; said fissile material including a mixture of uranium and recovered plutonium in rods of said fuel bundle at locations other than the corners of said fuel bundle; and, said neutron absorbing material being located in rods of said fuel bundle at rod locations in the corners of said fuel bundle; said neutron absorbing material having an axial distribution characterized by an enhancement in the axial zone of said fuel assembly, designated the cold shutdown zone, corresponding to at least a portion of the axial region of said core where the cold reactivity peaks, the aggregate amount of neutron absorbing material in said cold shutdown zone of said fuel assembly being greater than the aggregate amount of neutron absorbing material immediately above and below said cold shutdown zone whereby the cold shutdown reactivity is reduced relative to the cold shutdown reactivity in zones immediately above and below said cold shutdown control zone, said cold shutdown control zone having an axial extent measured from the bottom of the fuel assembly in the range between 60% and 90% of the height of said fissile material in said fuel assembly. 2. The invention of claim 1 and wherein said rods containing said neutron absorbing material do not contain fissile material including a mixture of uranium and recovered plutonium. 3. The invention of claim 1 and wherein said rods at the corners of said fuel bundle having a weighted axial distribution to define exclusively a "cold shutdown reactivity zone" of said fuel bundle. 4. The invention of claim 1 and wherein said fuel rods containing a component of fissile material including recovered plutonium and uranium include partial length rods. 5. In a boiling water reactor having a plurality of vertically upstanding fuel bundles; each fuel bundle containing longitudinally extending sealed rods with fissile material therein; a surrounding channel for containing water flow within each said fuel bundle from water exterior of said fuel bundle; water moderator flowing in the confined channel in the bottom of said fuel bundle to the top of said fuel bundle for producing steam and moderating neutrons to a thermal energy state for producing continuing nuclear reaction in said fissile material; said water moderator exterior of said confining channels for defining a core by-pass zone having relatively high concentration of water for moderation of neutrons from a high energy state to a thermal energy state for the continuation of said nuclear reaction; each rod including a component of fissile material distributed over an axial length of said fuel assembly; 6. The invention of claim 5 and wherein said fuel bundle includes an 8.times.8 array of fuel rods. 7. The invention of claim 5 and wherein said fuel bundle includes a 9.times.9 array of fuel rods. 8. The invention of claim 5 and wherein at least some of the rods at locations other than the corner of said fuel bundle are partial length rods. 9. The invention of claim 5 and including a fuel bundle having one corner rod placed in said bundle without neutron absorbing material being disposed in said corner rod. |
041475895 | abstract | A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilient members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod. |
description | Hereinafter, details of the embodiments of the present invention is explained referring to the FIGS. 1 to 5. The solidification facility of the present embodiment is capable of solidifying four kinds of radioactive waste, i.e. the miscellaneous solid waste, spent resin, dried powder of concentrated liquid waste, and ashes, into a solidifying container (details will be explained later), into a solidifying container 4. The solidification facility comprises; a solidifying agent injecting and kneading mechanism 50 (details will be explained later) for injecting and solidifying the miscellaneous solid waste (hereinafter, called only as injecting solidification), and for injecting the solidifying agent paste at kneading and solidifying the spent resin, the dried powder of concentrated liquid waste, and the ashes (hereinafter, called only as kneading solidification); a waste charging and kneading mechanism 60 (details will be explained later) for charging the radioactive waste and kneading at the kneading solidification; and a transferring mechanism 5 (detail will be explained later) for transferring the solidifying container 4 (details will be explained later) selectively to the solidifying agent injecting and kneading mechanism or the waste charging and kneading mechanism depending on whether the operation is for the injecting solidification or the kneading solidification. FIG. 2(a) and FIG. 2(b) are figures indicating the transferring route of the solidifying container 4 of the transferring mechanism 5 described above. In accordance with FIG. 2(a) and FIG. 2(b), the transferring mechanism 5 is composed of, for instance, a plurality of rollers 50 arranged in a transferring direction (so-called roller conveyer, refer to FIG. 4); respective of the rollers is central-controlled by control signals (not shown in the figure) from the controller 70 installed in a control room (not shown in the figure) of the solidification facility. The transferring mechanism 5 comprises; a main transferring route 5A; a sub-start up point route 5B, which merges with the main transferring route at a point of downstream side in the transferring direction near the start up point of the main transferring route 5A; and a sub-transferring route 5C, which branches at a point of downstream side from the merging point of the sub-start up point route 5B with the main transferring route 5A, and merges again with the main transferring route at a point of further downstream side from the branching point. Respective of the turn-tables 5a1-5a3 is provided at the respective of the merging point of the main transferring route 5A with the sub-start up point route 5B, and the merging point and the branching point of the main transferring route 5A with the sub-transferring route 5C. The turn-tables are controlled automatically by the controller 70 (or manual control by inputting operation signals from the control panel to the controller 70 can be used) to switch the transferring route of the solidifying container 4 (detail will be explained later). A sensor 80 (refer to FIG. 2) for detecting the solidifying container 4 when it is transferred is proved at the first position 5b on the main transferring route 5A (or a position before the first position 5b by a designated distance can be used). When detecting signals (refer to FIG. 2(a)) from the sensor 80 is transmitted to the controller 70, corresponding control signals (not shown in the figure) are output to the main transferring route 5A from the controller 70, and the solidifying container is stopped once at the first position 5b. The solidifying agent injecting and kneading mechanism 50 is provided at upper position of the first position 5b. A total outline of the compositions of the solidifying agent injecting and kneading mechanism 50 is indicated in FIG. 3(a). In accordance with FIG. 3(a), the solidifying agent injecting and kneading mechanism 50 comprises; a solidifying agent silo 11; a solidifying agent weighing apparatus 1 for weighing the solidifying agent, which is supplied from the solidifying agent silo 11 via the solidifying agent supply valve 16; an additive water supply line 12; an additive water weighing apparatus 2 for weighing the additive water, which is supplied from the additive water supply line 12 via the additive water supply valve 23; a kneader 3 for solidifying agent, which kneads the solidifying agent supplied from the solidifying agent weighing apparatus 1 through the solidifying agent supply valve 17 and the additive water supplied from the additive water weighing apparatus 2 via the additive water supply valve 18 to prepare the solidifying agent paste, and an injection valve 19 for injecting and filling the solidifying agent paste into the solidifying container 4. The kneader 3 for solidifying agent is out-drum type apparatus comprising a kneading vessel 3a, whereto the solidifying agent and the additive water are supplied, and a kneading blade (agitating blade) 3b driven by a motor for agitating inside of the kneading vessel 3a. Open-close operations of the waste supply valves 24a-24c and the waste supply valves 20a-20c are controlled by control signals from the controller 70 such as electric magnet valves, but details are omitted from the figures. The whole solidifying agent injecting and kneading mechanism 50 composed as described above is installed in an area separated from the area where the other radioactive handling apparatus and facilities such as the waste charging and kneading mechanism 60 and others by, for instance, separating walls. A sensor 81 (refer to FIG. 2) for detecting the solidifying container 4 when it is transferred there is provided at a second location 5c (at a position before a designated distance ) in the sub-transferring route 5B (that is, at downstream side in the transferring direction from the first location) as indicated in FIG. 2(a) and FIG. 2(b). When the detecting signals of the sensor 81 refer to FIG. 2) is transmitted to the controller 70, corresponding control signals are output from the controller 70 to the transferring mechanism sub-transferring route 5B, and the solidifying container 4 is stopped once at the second location 5c. The waste charging and kneading mechanism 60 is installed at the location upward the second location 5c. A whole schematic composition of the waste charging aid kneading mechanism 60 is indicated in FIG. 3(b). In accordance with FIG. 3(b), the waste charging and kneading mechanism 60 comprise; waste supply lines 13a-13c for supplying radioactive waste, waste weighing apparatus 6a-6c for weighing the radioactive waste supplied from the waste supply lines 13a-13c via the waste supply valves 24a-24c, a waste supply line 21 for supplying the radioactive waste, which is weighed at respective of the waste weighing apparatus 6a-6c, via the waste supply valves 20a-20c, a solidifying container elevator 10 for elevating the solidifying container 4, which is transferred to the second location 5c, upwards from the transferring line 5C (refer to FIG. 1, explained later), and a kneader 9 for kneading waste, which charges the radioactive waste supplied from the waste supply line 21 into the elevated solidifying container 4, and kneads the radioactive waste in the solidifying container 4. The kneader 9 for kneading waste is so-called in-drum type, which is provided with only kneading blade 9a (agitating blade) driven by motor for agitating inside of the solidifying container 4, and is composed so as to charge the radioactive waste supplied from the waste supply line 21 into the elevated solidifying container 4, and to knead the radioactive waste in the solidifying container 4 by dipping the kneading blade 9a therein (refer to FIG. 1, explained later). The solidifying container elevator 10 comprises a base 10a; an extendable arm mechanism 10b provided with, for instance, a hydraulic cylinder; and a solidifying container platform 10c located at the second location 5c in the sub-transferring route 5C; and the solidifying container platform 10c can be elevated or lowered depending on extending-shrinking motion of the extendable arm mechanism 10b corresponding to extending-shrinking motion of the hydraulic cylinder. The number of the waste supply lines 13a-13c, waste supply valves 24a-24c, waste weighing apparatus 6a-6c, and the waste supply valves 20a-20c to be provided are decided depending on the number of kinds of waste to be kneaded and solidified. For instance, spent resin is supplied through the waste supply line 13a, dried powder of concentrated liquid waste is supplied through the waste supply line 13b, and ashes is supplied through the waste supply line 13c. The solidifying agent supply valve 16, the solidifying agent supply valve 17, the additive water supply valve 23, and the additive water supply valve 18 are controlled their open-close motion 8 for instance electric magnet valve) by the control signals from the controller 70, but details are omitted in the figure. The whole waste charging and kneading mechanism 60 composed of the apparatus as explained above is installed in a radioactivity controlled area separated from the area where the other non-radioactive handling apparatus and facilities such as the solidifying agent injecting and kneading mechanism 50 and others by the separating walls 27 (refer to FIG. 3(a)). FIG. 4 is a schematic illustration indicating the composition of the turn-table 5a1 indicated in FIG. 2(a) and FIG. 2(b). The turn-table 5a is not operated for a special motion when the solidifying container 4 moves in a straight direction on the main transferring route 5A (that is, when it moves as route 5A1xe2x86x92turn-table 5a1xe2x86x92route 5A). However, when the moving direction must be changed (that is, when it moves as route 5A1xe2x86x92turn-table 5a1xe2x86x92route 5C), the turn-table is operated as follows. That is, a sensor 82 for detecting the solidifying container 4 when it is transferred is provided at a designated position on the turn-table 5a1 (or a position at a designated distance before the designated position on the route 5A1 can be used). When a detecting signal of the sensor 82 is transmitted to the controller 70, a stop controlling signal (not shown in the figure) corresponding to the detecting signal is output from the controller 70 to the driving roller 50 on the turn-table 5a1, and the solidifying container 4 is stopped once on the turn-table 5a1. Then, a control signal is output from the controller 70 to the driving device 83 for rotating the turn-table 5a1, and the turn-table 5a1 is rotated by 90 degrees in a direction indicated by an arrow A in FIG. 4. After completing the rotation, a driving control signal (not shown in the figure) is output from the controller 70 to the driving roller 50 on the turn-table 5a1, and the transfer is resumed by transferring the solidifying container 4 to the sub-transferring route 5C from the turn-table 5a1. Other two turn-tables 5a2, 5a3 are composed as same as above, but explanation is omitted. As explained above, the transferring mechanism 5 composes the transferring means for transferring the solidifying container as claimed in respective of the claims. A first kneading blade comprises the kneading blade 3b of the kneader 3 for solidifying agent, and the injecting means for injecting the solidifying agent paste in the kneader for solidifying agent comprises the injecting valve 19. The solidifying agent injecting and kneading means for injecting the solidifying agent paste into the solidifying container at the first location in the upstream side in the transferring direction of the transferring means comprises; solidifying agent silo 11, solidifying agent supply valve 16, solidifying agent weighing apparatus 1, solidifying agent supply valve 17, additive water supply valve 23, additive water weighing apparatus 2, additive water supply valve 18, kneader 3 for solidifying agent, and injecting valve 19. An elevating means for elevating the solidifying container, which is transferred to the second location by the transferring means, upwards from the transferring line of the transferring means comprises the solidifying container elevator 10, and the second kneading blade comprises the kneading blade 9a of the kneader 9 for waste. The waste charging and kneading means, which is capable of charging radioactive waste into the solidifying container at the second location in the downstream of the first location in the transferring direction of the transferring means and of kneading the radioactive waste in the solidifying container, comprises waste supply lines 13a-13c, waste supply valves 24a-24c, waste weighing apparatus 6a-6c, waste supply valves 20a-20c, waste supply lines 21, and kneader 9 for waste. The separating wall is composed of partition walls 27. Operation of the radioactive solidification facility composed as explained above of the present embodiment is explained hereinafter. In accordance with the solidification facility, an instruction for switching an injecting and solidifying mode for performing the injection and solidification, and a kneading and solidifying mode for performing kneading and solidification is input at the control room, and corresponding signal is output to the controller 70. Then, the controller 70 controls automatically respective of the apparatus so as to operate appropriately corresponding to respective of the wastes. (1) Injection and Solidification When the injecting and solidifying mode is selected in the control room, the solidifying container 4 is transferred by the route indicated in FIG. 2(a), and the solidification treatment is performed. The process is explained referring to FIG. 2(a) and FIG. 5. In accordance with FIG. 2(a), the solidifying container 4, wherein miscellaneous solid waste has been charged by operators previously, is loaded on the main transferring route 5A, and transferred to the first location 5b by the transferring mechanism 5 and stopped once. In a condition that the solidifying container 4 is stopped at the first location 5b, an adequate amount of solidifying agent for injection and solidification of the miscellaneous solid waste is weighed by the solidifying agent weighing apparatus 1, and injected into the kneader 3 for solidifying agent. Subsequently, an adequate amount of additive water for injection and solidification of the miscellaneous solid waste is weighed by the additive water weighing apparatus 2, and injected into the kneader 3 for solidifying agent. The solidifying agent and the additive water injected into the kneader 3 for solidifying agent is kneaded under a designated condition to be the solidifying agent paste, and injected into the solidifying container 4. The solidifying agent paste flows down through the intervals among the waste to fill the inside of the solidifying container 4. Accordingly, a solidified waste as same as the one obtained by the normal injection and solidification can be prepared. The solidified waste of the miscellaneous solid waste obtained as above is transferred as it is on the main transferring route 5A by the transferring mechanism 5, and stored in a storage place (not shown in the figure). (2) Kneading and Solidification When the kneading and solidifying mode is selected in the control room, the solidifying container 4 is transferred by the route indicated in FIG. 2(b), and the solidification treatment is performed. The process is explained referring to FIG. 2(b) and FIG. 1. In accordance with FIG. 2(b), the solidifying container 4 starts the sub-original point 5B as it is empty, which is different from the previous (1). The solidifying container 4 loaded on the sub-original point 5B is changed its transferring direction by the turn-table 5a1 to the main transferring route 5A, and transferred to the first location 5b by the transferring mechanism 5 and stopped once. In a condition that the solidifying container 4 is stopped at the first location 5b, an adequate amount of solidifying agent for treating the waste (hereinafter, called selected waste) selected at the control room from spent resin, dried powder of concentrated liquid waste, and ashes is weighed by the solidifying agent weighing apparatus 1, and injected into the kneader 3 for solidifying agent, as indicated in FIG. 1. Subsequently, an adequate amount of additive water for solidification of the selected waste is weighed by the additive water weighing apparatus 2, and injected into the kneader 3 for solidifying agent. The solidifying agent and the additive water injected into the kneader 3 for solidifying agent is kneaded under a designated condition to be the an 10 solidifying agent paste having an adequate water-cement ratio and weight for kneading and solidification of the selected waste, and injected into the solidifying container 4. The solidifying container 4 injected by the solidifying agent paste is transferred on the main transferring route 5A by the transferring mechanism 5, and changed its transferring direction by the turn-table 5a2 to the sub-transferring route 5C. The solidifying container 4 is transferred further to the second location 5c on the sub-transferring route 5C, and stopped there once. Under the condition that the solidifying container 4 is stopped at the second location 5c, the solidifying container 4 is elevated by the solidifying container elevator 10, which is installed under the kneader 9 for waste, until the upper periphery 4a (an opening portion) of the solidifying container 4 is touched with the lid portion 9b of the kneader 9 for waste. In this condition, under agitation and kneading by driving the kneading blade 9a, a designated amount of selected waste is charged into the solidifying container 4 from the waste weighing apparatus 6a-6c corresponding to the selected waste. Accordingly, a solidified waste, wherein the solidifying agent paste and the selected waste are mixed thoroughly, as same as the solidified waste (homogeneous solidified waste) obtained by the normal kneading and solidification can be prepared. At this time, the kneading is continued for a designated time after completion of charging the total amount of the selected waste. After completion of the kneading, the solidifying container 4 is lowered to the level of the transferring mechanism 5 again by the solidifying container elevator 10. During the above operation, the upper periphery 4a of the solidifying container 4 and the lower plane of the lid portion 9b of the kneader for waste is contacted tightly in order to prevent the kneaded material in the solidifying container 4 from splashing out. The homogeneous solidified waste obtained as described above is transferred on the sub-transferring route 5C again, changed its transferring direction by the turn-table 5a3 on the main transferring route 5A, and transferred on the main transferring route 5A to a storage place which is not shown in the figure. In accordance with the solidifying facility of the present embodiment, which is composed as described above, the injection and solidification of miscellaneous solid waste and kneading and solidification of a several kinds of waste can be performed selectively by a single facility. Furthermore, the solidifying agent injecting and kneading mechanism 50 can be made out-drum type, and the waste charging and kneading mechanism 60 can be made in-drum type. In any cases of the above injection-solidification and kneading-solidification, the solidifying agent injecting and kneading mechanism 50 is used only for kneading the non-radioactive solidifying agent and the additive water. Therefore, the solidifying agent injecting and kneading mechanism 50 can be installed in an area separated by a partition wall 27 from the area where the waste charging and kneading mechanism 60, which is radioactive apparatus, is installed, and generated washing liquid waste can be treated and disposed readily because the liquid waste is non-radioactive. On the other hand, the waste charging and kneading mechanism 60 is radioactive because the radioactive waste is charged when kneading and solidifying, and the generated washing liquid waste becomes radioactive secondary waste. However, the waste charging and kneading mechanism 60, which is a radioactive apparatus, can be made in-drum type, and washing object is only the kneading blade 9a. Therefore, the generating amount of the radioactive secondary waste can be made small. Generally, the solidifying agent paste prepared by kneading the solidifying agent and the additive water has a low viscosity, but the adding the waste and kneading increases the viscosity. In accordance with the out-drum type (kneading vessel+kneading blade) kneading and solidification, the increased viscosity of the kneading material generates a possibility to choke the outlet of the kneading vessel when the kneaded material is discharged. In accordance with the present embodiment, an advantage is realized that the choke described above can be prevented, because kneading the solidifying agent paste having a low viscosity is performed by the out-drum type solidifying agent injecting and kneading mechanism 50, and kneading the radioactive waste is performed by the in-drum type waste charging and kneading mechanism 60. In accordance with the present embodiment, changing the transferring direction of the solidifying container 4 is performed using the turn-table 5a by turning the turn-table 5a by 90 degrees under a condition that the solidifying container is loaded on the turn-table 5a, but methods of changing the transferring direction is not restricted to the above method, and other method is usable. That is, when the solidifying container 4 is stopped on the turn-table 5a, for instance, the solidifying container 4 is transferred to the sub-transferring route 5C by hanging the solidifying container 4 with a hanger which is provided particularly, by pushing the solidifying container 4 to the direction toward the sub-transferring route 5C with a pusher which is provided particularly, or other transferring mechanism, which has a function to transfer the solidifying container 4 to the sub-transferring route 5C, inserted beneath driving rollers 50 of the turn-table 5a to operate between the rollers 50. In accordance with the present invention, the injection and solidification operation and kneading and solidification operation can be performed selectively by a single facility, and generating amount of the radioactive secondary waste can be decreased. |
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055815880 | abstract | A method for mitigating crack initiation and propagation on the surface of metal components in a water-cooled nuclear reactor. An electrically insulating coating doped with a noble metal is applied on the surfaces of IGSCC-susceptible reactor components. The preferred electrically insulating material is yttria-stabilized zirconia doped with palladium or platinum. The presence of an electrically insulating coating on the surface of the metal components shifts the corrosion potential in the negative direction without the addition of hydrogen. Corrosion potentials .ltoreq.-0.5 V.sub.SHE are believed to be achievable even at high oxidant concentrations and in the absence of hydrogen, although the coatings are believed to be particularly suited to applications where a reductant, such as hydrogen, is present. |
abstract | Core debris generated during a molten reactor core in a reactor containment vessel penetrating the reactor containment vessel is configured to be caught by a core catcher located beneath the reactor containment vessel which has a main body having first stage cooling water channels and second stage surrounded by cooling fins extending radially. The number of the second stage cooling channels is larger than that of the first stage cooling channels. Cooling water is supplied from a cooling water injection opening and distributed to the first cooling water channels at a distributor. An intermediate header is formed between the first and the second cooling water channels, and the cooling water is distributed to the second cooling water channels uniformly. |
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062721976 | abstract | A fuel assembly for a nuclear reactor is described, the fuel assembly including: a plurality of fuel pins (12) extending substantially parallel to the axis of the assembly and to each other; at least two structural grids spaced apart from each other, the grids being in contact with said fuel pins (12) and maintaining said fuel pins substantially mutually parallel and preventing contact therebetween, wherein the fuel assembly further comprises at least one mixing grid (50) situated intermediate said at least two structural grids, the fuel assembly being characterized in that said mixing grid (50) is positioned and fixedly located out of substantial contact with said fuel pins (12), the mixing grid also having turbulence inducing means (61) to promote turbulence in a coolant (62) flowing through said fuel assembly in use and in that the mixing grid is formed from sheet metal wherein the plane of the metal sheet from which the mixing grid is formed lies in a plane which is transverse to the axis of the fuel pin assembly. |
048030430 | description | DETAILED DESCRIPTION OF THE INVENTION In the following description, like reference characters designate like or corresponding parts throughout the several views of the drawings. Also in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like are words of convenience and are not to be construed as limiting terms. In General Referring now to the drawings, and particularly to FIG. 1, there is shown an elevational view of a fuel assembly, represented in vertically foreshortened form and being generally designated by the numeral 10. The fuel assembly 10 is the type used in a pressurized water reactor (PWR) and basically includes a lower end structure or bottom nozzle 12 for supporting the assembly on the lower core plate (not shown) in the core region of a reactor (not shown), and a number of longitudinally extending guide tubes or thimbles 14 which project upwardly from the bottom nozzle 12. The assembly 10 further includes a plurality of transverse grids 16 constructed in accordance with the principles of the present invention, as will be described in detail below. The grids 16 are axially spaced along and supported by the guide thimbles 14. The assembly 10 also includes a plurality of elongated fuel rods 18 transversely spaced and supported in an organized array by the grids 16. Also, the assembly 10 has an instrumentation tube 20 located in the center thereof and an upper end structure or top nozzle 22 attached to the upper ends of the guide thimbles 14. With such arrangement of parts, the fuel assembly 10 forms an integral unit capable of being conveniently handled without damaging the assembly parts. As mentioned above, the fuel rods 18 in the array thereof in the assembly 10 are held in spaced relationship with one another by the grids 16 spaced along the fuel assembly length. Each fuel rod 18 includes nuclear fuel pellets 24 and the opposite ends of the rod are closed by upper and lower end plugs 26,28 to hermetically seal the rod. Commonly, a plenum spring 30 is disposed between the upper end plug 26 and the pellets 24 to maintain the pellets in a tight, stacked relationship within the rod 18. The fuel pellets 24 composed of fissile material are responsible for creating the reactive power of the PWR. A liquid moderator/coolant such as water, or water containing boron, is pumped upwardly through the fuel assemblies of the core in order to extract heat generated therein for the production of useful work. To control the fission process, a number of control rods 32 are reciprocally movable in the guide thimbles 14 located at predetermined positions in the fuel assembly 10. Specifically, the top nozzle 22 has associated therewith a rod cluster control mechanism 34 having an internally threaded cylindrical member 36 with a plurality of radially extending flukes or arms 38. Each arm 38 is interconnected to a control rod 32 such that the control mechanism 34 is operable to move the control rods 32 vertically in the guide thimbles 14 to thereby control the fission process in the fuel assembly 10, all in a well-known manner. Grids With Improved Sprinq Structure Referring now to FIGS. 2-11, there is shown the preferred embodiment of the transverse fuel rod grid 16 constructed in accordance with the principles of the present invention. Basically, the grid 16 includes a plurality of inner and outer straps 40,42 having slots 44 by which the straps are interleaved with one another in an egg-crate configuration to form a matrix of hollow cells 46 and a plurality of openings 48. At the intersections of the straps 40,42, they are suitably secured together, such as by welding. The hollow cells 46 of the grid 16 receive therethrough respective ones of the fuel rods 18, whereas the openings 48 of the grid 16 have sleeves 50 inserted therein and attached to the inner straps 40 by which the grid 16 is disposed along and attached to the guide thimbles 14. Each cell 46 receiving one fuel rod 18 is defined by pairs of opposing wall sections. The wall sections 52 compose the inner straps 40, whereas the wall sections 54 compose the outer straps 42. The inner strap wall sections 54 are shared with adjacent cells. As shown in FIGS. 3 to 5, the wall sections 54 of the outer strap 42 each has a pair of horizontally extending and vertically spaced fuel rod engaging dimples 56 integrally formed thereon in association with each cell 46. Similarly, the wall sections 52 of the inner straps 40 also each has a pair of horizontally extending and vertically spaced fuel rod engaging dimples 58 integrally formed thereon in association with each cell 46. One of the dimples 58 on each wall section 52 of the inner straps 40 is located above the spring structure 60 of the present invention, whereas the other dimple 58 is located below it. Each cell 46 formed along the periphery of the grid 16 by the inner and outer straps 40,42 has associated with it four dimples 56,58 and two spring structures 60, whereas each cell 46 formed in the grid 16 by inner straps 40 only has associated with it four dimples 56 and two spring structures 60. Thus, each fuel rod in each cell 46 is contacted at six circumferentially and axially displaced locations thereon. More particularly, as seen in FIGS. 6 to 18, each fuel rod spring engaging structure 60, in accordance with the principles of the present invention, is composed of resiliently yieldable flexible material of the inner straps 40, such as stainless steel metal. The components of the spring structure 60, which will be described next, are integrally formed, such as by a conventionally stamping operation, from and on each wall section 52 of the inner straps 40 in association with each cell 46 of the grid 16. Basically, each spring structure 60 includes a pair of laterally spaced elongated spring leg members 62 and 15 W.E. 53,856-I an elongated spring cross member 64. Each leg member 62 has a pair of opposite upper and lower ends 66,688 and is anchored by being integrally and rigidly connected to the upper and lower portions 70,72 of the respective wall section 52 at only the leg member upper and lower ends 66,688. Each cross member 64 has a pair of upper and lower opposite ends 74,76 and extends diagonally between and is integrally attached at such respective ends to the leg members 62 of a pair thereof such that the spring structure 60 formed by the leg and cross members 62,64 has an effective length greater than (for instance, two times) the actual length it occupies on each wall section 52 of the inner straps 40. More specifically, the cross member 64 at its upper end 74 is rigidly attached to one of the leg members 62 adjacent to the upper end 66 thereof and at its lower end 76 is rigidly attached to the other of the leg members 62 adjacent to the lower end 68 thereof. Also, each leg member 62 extends generally parallel to one another and in a direction generally parallel to the central longitudinal axis of the respective grid cell 46. Preferably the cross member 64 is disposed approximately forty-five degrees with respect to the leg members 62 and to the direction of coolant fluid flow through the grid and to the longitudinal axis of the respective grid cell 46. Still further, as seen in FIGS. 15 and 18, each leg member 62 is slightly bowed or arcuate-shaped in its configuration along a longitudinal section through the leg member. Given such curvature, the leg member 62 projects from the plane P of the wall section 52 into its associated one of the cells 46 toward the central longitudinal axis thereof. Each cross member 64 is arched or arcuate-shaped in configuration along a longitudinal section of the cross member. Given such configuration, the cross member 64 projects from the leg members 62 away from the wall section 52 a farther distance into the associated cell 46 toward its longitudinal axis than does the leg members 62. At such position, the cross member 64 at its middle point 78 engages the fuel rod 18 received through the cell 46. The cross member 64 is capable of resiliently deflecting or yielding in a direction generally orthogonal to and away from the longitudinal axis of the associated cell 46 and toward the wall section 52 upon engagement by a fuel rod when inserted in the cell 46. The leg members 62 project from the plane P of the wall section 52 when the cell 46 is unoccupied by a fuel rod 18. However, they are capable of resiliently deflecting back within the wall section plane P due to resilient deflection of the cross member 64 by its engagement with the fuel rod 18 in the cell 46. In such deflected positions, the leg members 62 do not block coolant flow through the grid 16. As seen in FIGS. 13 14, 16 and 17, the cross member 64 like the dimples 58 project from the wall section 52 so as to define an open space therebetween which permits unimpeded flow of coolant fluid therethrough and along the fuel rod received in the cell. In summary, the configuration of the spring structure 60 allows a very low profile, a reduced grid height, and a low spring constant. The spring cross member 64 is set forty-five degrees to the direction of coolant fluid flow, but in manufacture is stamped out parallel to the direction of flow. The spring structure 60 has the pair of integral flexible spring leg members 62 which allow the spring structure to be compliant without adding to flow blockage. While not forming part of the present invention, the grid 16 can have mixing vanes 80 formed along the top edge of the inner and outer straps 40,42 thereof. Grids With Improved Dimples Referring now to FIGS. 19-48, there is shown modified versions of dimples 78 on the inner straps 40 of the transverse fuel rod grid 16 of FIG. 2. Since all of the parts of the inner straps 40 illustrated in FIGS. 6-18 are identical to the parts of the inner straps 40 illustrated in FIGS. 19-48 except for the dimples 58 which are now replaced by modified dimples 78, all of these identical parts will not be identified again in FIGS. 19-48. Each dimple 78 is made of the same resiliently yieldable material of the straps 40 and integrally formed, such as by a stamping operation, on each wall section 52 thereof of each cell 46. As before, the dimples 78 as located spaced above and below the spring structure 60 on the wall section 52. Also, each dimple 78 extends or projects into the respective grid cell 46 and is arcuate-shaped along a longitudinal section therethrough. However, each dimple 78 is modified to extend in a diagonal orientation with respect to the central longitudinal axis of the respective grid cell 46. Preferably, the diagonal orientation of each dimple 78 is approximately forty-five degrees with respect to the cell longitudinal axis, the same as the angular orientation of the cross member 64 of each spring structure 60. In FIGS. 20, 21 and 28, the upper and lower dimples 78 are oriented generally parallel to one another. However, the dimples 78 are oriented generally orthogonal or perpendicular to the orientation of the spring structure cross members 64. In FIGS. 23, 24 and 35, the upper and lower dimples 78 are oriented again generally parallel to one another. This time the dimples 78 are also oriented generally parallel to the orientations of the spring structure cross members 64. Finally, in FIGS. 26, 27 and 47, the upper dimples 78 are oriented generally perpendicular to both the lower dimples 78 and the spring structure cross members 64. It should be readily apparent that the respective orientations of the upper and lower dimples 78 could be reversed with respect to one another and to the cross members 64. It is thought that the invention and many of its attendant advantages will be understood from the foregoing description and it will be apparent that various changes may be made in the form, construction and arrangement thereof without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinbefore described being merely a preferred or exemplary embodiment thereof. |
abstract | The invention comprises a method and apparatus for treating a tumor of a patient, in a beam treatment center comprising a floor, with positively charged particles, comprising: (1) a synchrotron mounted to an elevated floor section above the floor of the beam treatment center; (2) a beam transport system, comprising: at least three fixed-position beam transport lines, where none of the synchrotron and the beam transport system penetrate through the floor of the beam treatment center; (3) the positively charged particles transported from the synchrotron, through the beam transport system, to a position above a patient positioning system during use; and (4) an optional repositionable nozzle system connected to a first, second, and third fixed-position beam transport line at a first, second, and third time, respectively, where the nozzle track forms an arc of a circle and the repositionable nozzle system moves along the nozzle track. |
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description | The present application claims priority from Japanese Patent application serial no. 2013-185070, filed on Sep. 6, 2013, the content of which is hereby incorporated by reference into this application. 1. Technical Field The present invention relates to a method of chemical decontamination for carbon steel member of a nuclear power plant and more particularly to a method of chemical decontamination for carbon steel member of a nuclear power plant suitable for application to carbon steel member of a boiling water nuclear power plant. 2. Background Art For example, the boiling water nuclear power plant (hereinafter referred to as BWR plant) includes a reactor having a core disposed in a reactor pressure vessel (referred to as RPV). Reactor water (cooling water) supplied to the core by a recirculation pump (or an internal pump) is heated by heat generated due to nuclear fission of a nuclear fuel material in a fuel assembly loaded in the core and is partially turned to steam. The steam is introduced from the RPV to a turbine to rotate the turbine. The steam discharged from the turbine is condensed by a condenser to water. The water is supplied to the RPV as feed water. Metallic impurities are mainly removed from the feed water by a demineralizer installed in a water feed pipe so as to suppress generation of a radioactive corrosion product in the RPV. The reactor water is cooling water existing in the RPV. Further, a corrosion product which is a base of the radioactive corrosion product is generated on a surface of a structure member of a BWR plant such as an RPV and primary loop recirculation system piping (referred to as recirculation system pipe), the surface coming into contact with the reactor water, so that stainless steel and a nickel based alloy of less corrosion are used for the main primary-system structure members. Further, overlay welding of stainless steel exists on an inner surface of the RPV made of low alloy steel, thus the low alloy steel is prevented from direct contact with the reactor water. Furthermore, part of the reactor water is cleaned up by a demineralizer of a reactor water clean-up system, thus metallic impurities slightly existing in the reactor water is removed positively. However, even if such a corrosion countermeasure as mentioned above is taken, very little metallic impurities unavoidably exist in the reactor water, so some metallic impurities, as a metallic oxide, are adhered to the surface of each fuel rod included in a fuel assembly. The impurities (for example, a metallic element) deposited on the surface of each fuel rod cause a nuclear reaction by irradiation of neutrons discharged by nuclear fission of the nuclear fuel in each fuel rod and become radioactive nuclides such as cobalt 60, cobalt 58, chromium 51, and manganese 54. These radioactive nuclides are mostly kept to be adhered to the surface of each fuel rod in a form of an oxide. However, some radioactive nuclides are eluted as ions into the reactor water depending of the solubility of the taken-in oxide and are re-discharged into the reactor water as an insoluble called a crud. The radioactive material included in the reactor water is removed by the reactor water clean-up system communicated with the RPV. The radioactive material not removed by the reactor water clean-up system is accumulated on the surface of the structure member (for example, pipe) of the nuclear power plant which comes into contact with the reactor water while circulating in the re-circulation system together with the reactor water. As a result, a radiation is discharged from the surface of the structure member, causing radiation exposure to an operator during the periodic inspection operation. The exposure dose of the operator is controlled so as not to exceed the regulated value for each operator. The regulated value has been reduced in recent years and there is the need to decrease the exposure dose for each operator as much as possible. Therefore, when the exposure dose during the periodic inspection operation is expected to be high, the chemical decontamination for dissolving and removing the radioactive nuclide deposited on the pipe is executed. For example, Japanese Patent Laid-open No. 2000-105295 proposes a chemical decontamination method of executing reduction decontamination using an aqueous solution (a reduction decontaminating solution) including an oxalic acid and hydrazine, decomposition of the oxalic acid and hydrazine, and oxidation decontamination using an aqueous solution (an oxidation decontaminating solution) including a potassium permanganate. The chemical decontamination method is executed for the pipe and the like of the nuclear power plant. Japanese Patent Laid-open No. 2001-74887 describes a chemical decontamination method executed to a recirculation system pipe made of stainless steel which is connected to the RPV and a purification system pipe made of carbon steel member of the reactor water clean-up system which is connected to the recirculation system pipe. In the chemical decontamination method, a potassium permanganate aqueous solution is supplied into the recirculation pipe and the purification system pipe to execute the oxidation decontamination for the inner surfaces of those pipes. Thereafter, an aqueous solution including the oxalic acid and hydrazine is supplied to the recirculation system pipe and the purification system pipe to execute the reduction decontamination. After the reduction decontamination, the oxalic acid and hydrazine included in the aqueous solution are decomposed. Further, Japanese Patent Laid-open No. 2004-286471 and Japanese Patent Laid-open No. 2004-170278 describe a chemical decontamination method of storing the decontamination objects such as the equipment made of stainless steel and pipe which are removed from the nuclear power plant in a decontamination bath and executing the chemical decontamination. In the chemical decontamination method, a mixed aqueous solution including a formic acid of a concentration ratio of 0.9 and an oxalic acid of a concentration ratio of 0.1 is supplied into the decontamination bath to decontaminate the decontamination objects and the reduction decontamination of the decontamination objects is executed in the decontamination bath by using the mixed aqueous solution. After completion of the reduction decontamination, hydrogen peroxide (or ozone) is supplied into the mixed solution and the formic acid and oxalic acid included in the mixed aqueous solution are decomposed by the hydrogen peroxide (or ozone). Japanese Patent Laid-open No. 2002-333498 describes a chemical decontamination method. In the chemical decontamination method, the chemical decontamination, concretely, reduction decontamination of carbon steel member using an aqueous solution (a reduction decontamination aqueous solution) including an organic acid (for example, the formic acid) and hydrogen peroxide is executed. Furthermore, in the chemical decontamination method described in Japanese Patent Laid-open No. 2003-90897, the reduction decontamination for the carbon steel member is executed using the oxalic acid aqueous solution, and after the reduction decontamination, an acid aqueous solution (for example, a formic acid aqueous solution) is brought into contact with the carbon steel member. Therefore, at the time of the reduction decontamination using the oxalic acid aqueous solution, the ferrous oxalate generated on the surface of the carbon steel member is removed by action of the acid aqueous solution. Japanese Patent Laid-open No. 62-250189 describes a chemical decontamination method of executing the reduction decontamination for equipment made of stainless steel of a primary cooling system device by using a solution including a malonic acid, the oxalic acid, and hydrazine. [Patent Literature 1] Japanese Patent Laid-open No. 2000-105295 [Patent Literature 2] Japanese Patent Laid-open No. 2001-74887 [Patent Literature 3] Japanese Patent Laid-open No. 2004-286471 [Patent Literature 4] Japanese Patent Laid-open No. 2004-170278 [Patent Literature 5] Japanese Patent Laid-open No. 2002-333498 [Patent Literature 6] Japanese Patent Laid-open No. 2003-90897 [Patent Literature 7] Japanese Patent Laid-open No. 62(1987)-250189 In the reduction decontamination using the oxalic acid aqueous solution aiming at a stainless steel member, the iron concentration in the oxalic acid aqueous solution does not rise so as to deposit ferrous oxalate. However, as described in Japanese Patent Laid-open No. 2001-74887, when executing the reduction decontamination for the carbon steel member (for example, the purification system pipe of the reactor water clean-up system) using the oxalic acid aqueous solution, if the ratio of the carbon steel member to the oxalic acid aqueous solution rises, the iron concentration in the oxalic acid aqueous solution rises and ferrous ions eluted in the oxalic acid aqueous solution due to dissolution of magnetite which is a base metal of the carbon steel member and an oxide film, reacts the oxalic acid to form a complex and the complex, that is, ferrous oxalate is deposited on the surface of the carbon steel member in contact with the oxalic acid aqueous solution. The ferrous oxalate is low in solubility, so that it deposits on the surface of the carbon steel member which is a main generation source of ferrous ions. When the ferrous oxalate is deposited on the oxide film formed on the surface of the carbon steel member, the dissolution of the oxide film by the oxalic acid aqueous solution is hindered at the time of reduction decontamination. As a result, the dissolution of the radioactive nuclide included in the oxide film is suppressed and the efficiency of the chemical decontamination for the carbon steel member is reduced. In Japanese Patent Laid-open No. 2002-333498, an aqueous solution including an organic acid (for example, a formic acid) and hydrogen peroxide is used to improve the solubility of the oxide film formed on the surface of the carbon steel member. To remove the ferrous ions eluted in the aqueous solution by the dissolution of the oxide film and cations of the radioactive nuclide, the aqueous solution including the organic acid, hydrogen peroxide, and ferrous ions needs to be supplied to a cation exchange resin column filled with a cation exchange resin. However, the hydrogen peroxide deteriorates the cation exchange resin in the cation exchange resin column, so that the aqueous solution including the eluted ferrous ions, eluted cations of the radioactive nuclide, organic acid, and hydrogen peroxide cannot be supplied to the cation exchange resin column, and the concentrations of the ferrous ions and cations of the radioactive nuclide cannot be lowered. As a result, the chemical decontamination efficiency for the carbon steel member is reduced. In the chemical decontamination method described in Japanese Patent Laid-open No. 2003-90897, after the oxalic acid included in the oxalic acid aqueous solution is decomposed, the ferrous oxalate deposited on the surface of the carbon steel member in the reduction decontamination of the carbon steel member is dissolved by using the oxalic acid aqueous solution using the formic acid aqueous solution. However, since the ferrous oxalate is deposited on the oxide film on the surface of the carbon steel member while the reduction decontamination for the carbon steel member using the oxalic acid aqueous solution is executed, the dissolution of the oxide film due to the oxalic acid aqueous solution is suppressed. Further, the chemical decontamination method described in Japanese Patent Laid-open No. 2003-90897 executes the ferrous oxalate decomposition process using a formic acid aqueous solution after the reduction decontamination process for the carbon steel member using the oxalic acid aqueous solution. Thus, in the chemical decontamination method described in Japanese Patent Laid-open No. 2003-90897, the time required for the chemical decontamination for the carbon steel member becomes longer. An object of the present invention is to provide a chemical decontamination method for the carbon steel member of the nuclear power plant capable of further improving efficiency of reduction decontamination for the carbon steel member. A feature of the present invention for attaining the above object is a chemical decontamination method comprising steps of bringing a reduction decontaminating solution including a malonic acid and an oxalic acid within a range from 50 to 400 ppm into contact with a surface of a carbon steel member of a nuclear power plant; and executing reduction decontamination for the surface of the carbon steel member by the reduction decontaminating solution. The film of a ferrous oxide formed on the surface of the carbon steel member is dissolved by the oxalic acid, and the base metal of the carbon steel member is dissolved by the malonic acid. As a consequence, the ferrous oxide, and the radioactive nuclides included in the base metal of the carbon steel member are eluted into the reduction decontaminating solution. The oxalic acid concentration included in the reduction decontaminating solution is within the range from 0 ppm to 400 ppm, so that the deposition of the ferrous oxalate onto the ferrous oxide film formed on the surface of the carbon steel member is suppressed and the dissolution of the ferrous oxide film by the oxalic acid can be performed efficiently. Since the dissolution of the ferrous oxide film can be performed efficiently, the dissolution of the portion including the radioactive nuclide of the base metal of the carbon steel member also can be performed efficiently by the malonic acid. Therefore, the reduction decontamination efficiency for the carbon steel member can be further improved. The above object can be accomplished even by bringing a reduction decontaminating solution including the malonic acid and oxalic acid with oxygen gas injected into contact with the surface of the carbon steel member of the nuclear power plant and performing the reduction decontamination by the reduction decontaminating solution for the surface of the carbon steel member. According to the present invention, the reduction decontamination effects for the carbon steel member can be further improved. The inventors variously investigated a method of being able to furthermore improve efficiency of reduction decontamination for a carbon steel member and as a result, have come to recognize that the suppression of deposition of the ferrous oxalate and the continuous removal of the ferrous ions eluted into the reduction decontaminating solution by the reduction decontamination and cations of the radioactive nuclide need to be accomplished at the time of the reduction decontamination for the carbon steel member. And, the inventors found a method of chemical decontamination for the carbon steel member capable of accomplishing them. The investigation contents performed by the inventors and the obtained results will be explained below. The inventors, firstly, conducted a test of confirming the effects of the reduction decontamination which is a kind of chemical decontamination for test specimens made of carbon steel using an aqueous solution (reduction decontaminating solution) of chemical decontamination agent, concretely, the respective aqueous solutions of the oxalic acid, formic acid, and malonic acid. In this test, the oxalic acid aqueous solution, formic acid aqueous solution, and malonic acid aqueous solution were filled in different beakers and a test specimen made of carbon steel was separately immersed in the aqueous solution at 90° C. in each of the beakers for 6 hours. In this way, the reduction decontamination for each test specimen by each aqueous solution was performed. The results obtained by this test are shown in FIG. 4. FIG. 4 shows the changes in the dissolution thickness of the test specimens for the change in the pH of each of the aqueous solutions. The dissolution thickness of the test specimens made of carbon steel member depends on the aqueous solution with each test specimen immersed and the result of (the formic acid aqueous solution)>(the malonic acid aqueous solution)>(the oxalic acid aqueous solution) was obtained from the test results shown in FIG. 4. The dissolution thickness of the test specimens immersed in the formic acid aqueous solution was largest and the dissolution thickness of the test specimens immersed in the oxalic acid aqueous solution was smallest. The test specimens immersed in the oxalic acid aqueous solution were dissolved little. Further, yellow deposits seen as ferrous oxalate were adhered to the surface of each test specimen immersed in the oxalic acid aqueous solution. In the reduction decontamination for the test specimens using the malonic acid aqueous solution, when the pH of the aqueous solution was within the range from 1.7 (the malonic acid concentration of the malonic acid aqueous solution: 19000 ppm) to 2.0 (the malonic acid concentration: 5200 ppm), the test specimens made carbon steel was able to be dissolved. Furthermore, if the pH of the malonic acid aqueous solution becomes 1.8 (the malonic acid concentration: 12000 ppm) or lower, the dissolution of the test specimens made of carbon steel increases more quickly than a case of the pH of 1.9 (the malonic acid concentration: 7800 ppm) or higher. Furthermore, the test of confirming the solubility of the hematite (α-Fe2O3) and the magnetite (Fe3O4) which are ferrous oxides was conducted using the oxalic acid aqueous solution, the formic acid aqueous solution, and the malonic acid aqueous solution. In this test, the oxalic acid aqueous solution, the formic acid aqueous solution, and the malonic acid aqueous solution of 300 ml each were filled in different beakers and the temperature of each aqueous solution was kept at 90° C. The pH of each aqueous solution is 2.0. The hematite which is a ferrous oxide was immersed for 6 hours in the aqueous solution filled in each beaker and the solubility of the hematite by each aqueous solution was confirmed. And, the magnetite which is a different ferrous oxide was immersed in each aqueous solution filled in different beakers under the same condition as the hematite and the solubility of the magnetite by the respective aqueous solutions was confirmed. The results obtained by this test are shown in FIG. 5. FIG. 5 shows the solubility of the hematite and magnetite by the ferrous ion concentration in the oxalic acid aqueous solution, the formic acid aqueous solution, and the malonic acid aqueous solution which are a reduction decontaminating solution. It shows that the respective solubility of the hematite and magnetite increases as the ferrous ion concentration increases. The solubility of the hematite and magnetite became (the oxalic acid aqueous solution)>(the malonic acid aqueous solution)>(the formic acid aqueous solution) and the solubility of the hematite and magnetite by the oxalic acid aqueous solution became highest. Further, the formic acid aqueous solution could hardly dissolve the hematite. According to the above test results, it is found that the malonic acid is preferable for the dissolution of the carbon steel member and ferrous oxide. Further, if a very small quantity of oxalic acid is added to the malonic acid aqueous solution, the dissolution of the ferrous oxide which is an oxide film formed on the surface of the carbon steel member can be improved with the dissolution rate of the carbon steel member kept. The inventors conducted the test of confirming the dissolution of the carbon steel member by the aqueous solution including the malonic acid and oxalic acid which was generated by adding the oxalic acid to the malonic acid aqueous solution. The oxalic acid concentration was changed from 0 ppm to 1200 ppm in the malonic acid aqueous solution with a malonic acid concentration of 5200 ppm, and the malonic acid aqueous solutions with a different oxalic acid concentration were filled in different beakers in a predetermined volume, and the temperature of each malonic acid aqueous solution was held at 90° C. The test specimens made of carbon steel were immersed in the malonic acid aqueous solution with a different oxalic acid concentration in each beaker for 6 hours, and the reduction decontamination was performed for each test specimen. In this test, no oxygen gas was injected into the malonic acid aqueous solution in each beaker. The results obtained by this test are shown by ◯ marks (no oxygen is injected into the malonic acid aqueous solution) in FIG. 6. Further, in FIG. 6, the test results obtained by immersing the test specimens made of carbon steel in the malonic acid aqueous solutions of a different oxalic acid concentration with oxygen gas injected are also shown by ● marks. The conditions of the test using the malonic acid aqueous solutions of a different oxalic acid concentration with oxygen gas injected are the same as the conditions of the test using the malonic acid aqueous solutions of a different oxalic acid concentration with no oxygen gas injected. When the oxalic acid concentration of the malonic acid aqueous solution was within a range from 50 to 400 ppm, the dissolution thickness of each test specimen made of carbon steel became larger than the dissolution thickness of each test specimen made of carbon steel by the malonic acid aqueous solution with no oxalic acid added. On the other hand, if the oxalic acid concentration of the malonic acid aqueous solution became 500 ppm or higher, the dissolution thickness of each test specimen made of carbon steel became smaller than the dissolution thickness of the test specimen made of carbon steel by the malonic acid aqueous solution including no oxalic acid. Further, when oxygen gas was injected into the malonic acid aqueous solutions with a different oxalic acid concentration, the dissolution thickness of each test specimen made of carbon steel was increased than the case that no oxygen gas was injected into the malonic acid aqueous solution including the oxalic acid within the range of the oxalic acid concentration from 50 to 400 ppm. The inventors, furthermore, conducted the test of confirming the dissolution of the ferrous oxide using the malonic acid aqueous solution with the oxalic acid concentration changed within a range from 0 to 200 ppm. The malonic acid concentration of the malonic acid aqueous solution (reduction decontaminating solution) used in this test is 5200 ppm. The oxalic acid concentration in the malonic acid aqueous solution with a malonic acid concentration of 5200 ppm was changed at the four stages of 0 ppm, 50 ppm, 100 ppm, and 200 ppm within the range from 0 to 200 ppm. As mentioned above, four kinds of malonic acid aqueous solutions with a different oxalic acid concentration were filled in different beakers in volume of 300 ml each and the temperature of the malonic acid aqueous solution in each beaker was held at 90° C. The ferrous oxide (for example, the hematite or magnetite) was immersed in the malonic acid aqueous solution in each beaker for 6 hours. The obtained test results are shown in FIG. 7. Based on the test results shown in FIG. 7, it is found that the ferrous ion concentration increases, that is, the solubility of the ferrous oxide increases as the oxalic acid concentration of the malonic acid aqueous solution increases. The inventors conducted the test of confirming the change with time of the dissolution thickness of each test specimens made of carbon steel when the reduction decontamination was performed by the aqueous solution including the malonic acid and oxalic acid. The results obtained by this test are shown in FIG. 8. In FIG. 8, the changes in the ferrous ion concentration in the aqueous solution (reduction decontaminating solution) including the malonic acid and oxalic acid are also shown together with the change with time of the dissolution thickness of each test specimen. If the ferrous ion concentration in the reduction decontaminating solution enters the saturation state, the dissolution thickness of each test specimens made of carbon steel is apt to be saturated as well. A ferrous dissolution rate dM/dt from the carbon steel member which is a test specimen is expressed by Formula (1) based on an Fe ion concentration Cbulk in the bulk water, an Fe ion concentration Cs on the surface of the carbon steel member, and a ferrous dissolution rate k from the carbon steel member. Namely, if the Fe ion concentration Cbulk in the bulk water increases, the ferrous dissolution rate k from the carbon steel member is reduced.dM/dt=k×(Cbulk−Cs) (1) Therefore, the removal of ferrous ions from the reduction decontaminating solution is necessary to increase the solubility of the carbon steel member. The inventors conducted the test of investigating the effect on the dissolution of the carbon steel member by the temperature of the aqueous solution including the malonic acid and oxalic acid. In this test, the malonic acid aqueous solution (no oxalic acid is included) with a malonic acid concentration of 5200 ppm and the aqueous solution including the malonic acid of 5200 ppm and the oxalic acid of 100 ppm were filled separately in beakers, and the test specimens made of carbon steel were separately immersed in the aqueous solutions in the respective beakers. And, the temperature of each aqueous solution was changed within the range from 60° C. to 90° C. and the dissolution thickness of each test specimen immersed in each aqueous solution was measured under each temperature condition. Further, when a certain aqueous solution is boiled, the radioactive nuclide dissolved in the aqueous solution may be scattered in correspondence with the generated steam, so that the temperature of the aqueous solution is held at lower than the boiling point. The results obtained in this test are shown in FIG. 9. Based on the test results shown in FIG. 9, it is found that if the temperature of the aqueous solution including the malonic acid and oxalic acid is kept at 60° C. or higher, the carbon steel member can be dissolved. Particularly, if the temperature of the aqueous solution including the malonic acid and oxalic acid is increased to 80° C. or higher, the solubility of the carbon steel member is increased. Based on the above test results, a first proposal of realizing the suppression of deposition of the ferrous oxalate and the continuous removal of the ferrous ions and cations of the radioactive nuclide eluted into the reduction decontaminating solution by the reduction decontamination and furthermore improving efficiency of the reduction decontamination for the carbon steel member is to execute the reduction decontamination for the carbon steel member using the aqueous solution (reduction decontaminating solution) including the malonic acid and oxalic acid with an oxalic acid concentration existing within the range from 50 to 400 ppm. By performing the reduction decontamination for the carbon steel member using such a solution, it is possible to improve the solubility of the ferrous oxide formed on the surface of the carbon steel member in contact with the reduction decontaminating solution for the carbon steel member with the solubility of the carbon steel member by the malonic acid kept and also improve the efficiency of the reduction decontamination for the carbon steel member further. The malonic acid concentration of the reduction decontaminating solution including the malonic acid and oxalic acid with the oxalic acid concentration existing within the range from 50 to 400 ppm is desirably set within the range from 2100 to 19000 ppm. The malonic acid concentration of the aforementioned reduction decontaminating solution is desirably set within a range from 2100 to 7800 ppm from the viewpoint of suppressing damage of the equipment and pipes used in the nuclear power plant in common. On the other hand, in the aforementioned reduction decontaminating solution (the solution including the malonic acid and oxalic acid with the oxalic acid concentration existing within the range from 50 to 400 ppm) used in the reduction decontamination for the equipment and pipes (carbon steel member) made of carbon steel which are removed due to replace in the nuclear power plant and become wastes, the malonic acid concentration is desirably set within a range from 12300 to 19000 ppm. The temperature of the reduction decontaminating solution during the reduction decontamination is desirably set within a range from 60° C. to the temperature at the boiling point of the reduction decontaminating solution, preferably within a range from 80° C. to the temperature at the boiling point. A second proposal of realizing the suppression of deposition of the ferrous oxalate and the continuous removal of the ferrous ions and cations of the radioactive nuclide eluted into the reduction decontaminating solution by the reduction decontamination and furthermore improving the efficiency of the reduction decontamination for the carbon steel member is to execute the reduction decontamination for the carbon steel member using the aqueous solution including the malonic acid and oxalic acid with oxygen gas supplied. By performing the reduction decontamination for the carbon steel member using such an aqueous solution, it is possible to improve the solubility of the ferrous oxide formed on the surface of the carbon steel member in contact with the reduction decontaminating solution for the carbon steel member with the solubility of the carbon steel member by the malonic acid kept and also improve the efficiency of the reduction decontamination for the carbon steel member further. The embodiments of the present invention in which the aforementioned investigation results are reflected will be explained below. A method of chemical decontamination for a carbon steel member of a nuclear power plant according to embodiment 1 which is a preferred embodiment of the present invention will be explained by referring to FIGS. 1, 2, and 3. The method of chemical decontamination for a carbon steel member of a nuclear power plant according to the present embodiment is an example applied to a pipe (for example, the purification system pipe) of the boiling water nuclear power plant (hereinafter referred to as BWR plant), the pipe being made of carbon steel. This pipe is a carbon steel member. A general structure of the BWR plant to which the method of chemical decontamination for a carbon steel member of a nuclear power plant according to the present embodiment is applied will be explained by referring to FIG. 2. The BWR plant is provided with a reactor 1, a turbine 10, a condenser 12, a primary loop recirculation system, a reactor water clean-up system, and a water feed system. The reactor 1 installed in a reactor primary containment vessel 7 includes a reactor pressure vessel (hereinafter referred to as RPV) 2 having a core 3 disposed in the RPV 2. Jet pumps 6 are installed in the RPV 2. A plurality of fuel assemblies (not shown) are loaded in the core 3. Each fuel assembly includes a plurality of fuel rods filled with a plurality of fuel pellets manufactured with a nuclear fuel material. Several primary loop recirculation systems include a recirculation pump 5 and a primary loop recirculation system piping (referred to as recirculation system pipe) 4 made of stainless steel, respectively and the recirculation pump 5 is installed on the recirculation system pipe 4. In the recirculation system pipe 4, a valve 9 is installed on the upstream side of the recirculation pump 5 and a valve 8 is installed on the downstream side of the recirculation pump 5. Particularly, the valve 9 is installed on the upstream side of a connection point of the recirculation system pipe 4 to a purification system pipe 21. The water feed system has a structure that a condensate pump 14, a condensate purification apparatus 15, a low-pressure feed water heater 16, a water feed pump 17, and a high-pressure feed water heater 18 are installed on a water feed pipe 13 connecting the condenser 12 to the RPV 2 in this order from the condenser 12 toward the RPV 2. A hydrogen injection apparatus 20 is connected to the water feed pipe 13 on the upstream side of the condensate pump 14. The reactor water clean-up system is structured so that a purification system pump 22, a regeneration heat exchanger 23, a non-regeneration heat exchanger 24, and a reactor water purification apparatus 25 are installed on the purification system pipe 21 connecting the recirculation system pipe 4 and the water feed pipe 13 in this order from the upstream side toward the downstream side. The purification system pipe 21 is connected to the recirculation system pipe 4 on the upstream side of the recirculation pump 5. Cooling water (hereinafter referred to as reactor water) in the RPV 2 is pressurized by the recirculation pump 5 and is jetted into a bell mouth (not shown) of the jet pump 6 from a nozzle (not shown) of the jet pump 6 through the recirculation system pipe 4. The reactor water existing around the nozzle is sucked into the bell mouth by the action of the jetted water jetted from the nozzle. The reactor water discharged from the jet pump 6 is supplied to the core 3 and is heated by heat generated due to nuclear fission of a nuclear fuel material in the fuel rods. Part of the heated reactor water is turned steam. The steam is discharged into a main steam pipe 11 from the RPV 2, is introduced to the turbine 10 through the main steam pipe 11, and rotates the turbine 10. A generator (not shown) connected to the turbine 10 is also rotated and generates power. The steam discharged from the turbine 10 is condensed to water by the condenser 12. This water is supplied into the RPV 2 through the water feed pipe 13 as feed water. The feed water flowing through the water feed pipe 13 is pressurized by the condensate pump 14, and impurities including in the feed water are moved by the condensate purification apparatus 15. The feed water is further pressurized by the water feed pump 17 and is heated by the low-pressure feed water heater 16 and the high-pressure feed water heater 18. The extraction steam extracted from the main steam pipe 11 and the turbine 10 by the extraction pipe 19 is supplied to the low-pressure feed water heater 16 and the high-pressure feed water heater 18 as a heating source for the feed water flowing through the water feed pipe 13. The reactor water in the RPV 2 is subjected to irradiation of a radiation generated in correspondence to nuclear fission of a nuclear fuel material included in each fuel assembly loaded in the core 3, thereby causes radiolysis, and generates an oxidizing agent such as hydrogen peroxide and oxygen. The electrochemical corrosion potential of the structure member of the BWR plant which makes contact with the reactor water rises by the oxidizing agent. Therefore, in the BWR plant, hydrogen is injected into the feed water flowing in the water feed pipe 13 from the hydrogen injection apparatus 20. The hydrogen included in the feed water is injected into the reactor water in the RPV 2. The hydrogen and the oxidizing agent such as the hydrogen peroxide and oxygen included in the reactor water are reacted on each other, thus the oxidizing agent concentration of the reactor water is reduced and the electrochemical corrosion potential of the structure member of the BWR plant is lowered. In the BWR plant mentioned above, since the BWR plant is shut down in order to exchange the fuel assemblies loaded in the core 3, the chemical decontamination for the purification system pipe 21 which is a carbon steel member is executed after the operation of the BWR plant is stopped. The chemical decontamination is performed in the state that one end portion of a circulation pipe 29 of a chemical decontamination apparatus 28 is connected to a valve 26 installed on the purification system pipe 21 and the other end portion of the circulation pipe 29 is connected to a valve 27 installed on the purification system pipe 21. A recirculation system pipe 4 side of the valve 26 is closed by a closed plug (not shown) so as to prevent the chemical decontaminating solution from flowing, and a regeneration heat exchanger 23 side of the valve 27 is also closed by another closed plug (not shown). The detailed structure of the chemical decontamination apparatus 28 will be explained by referring to FIG. 3. The chemical decontamination apparatus 28 is provided with the circulation pipe (the chemical decontaminating solution pipe) 29, a cooling apparatus 30, a surge tank 31, a malonic acid injection apparatus 32, an oxalic acid injection apparatus 37, a cation exchange resin column 42, a mix bed ion exchange resin column 43, a decomposition apparatus 44, an oxidation agent supply apparatus 45, and circulation pumps 82 and 83. An open/close valve 48, the circulation pump 82, the cooler 30, valves 49 and 50, the surge tank 31, the circulation pump 83, and an open/close valve 51 are installed on the circulation pipe 29 in this order from the upstream side. A valve 53, the cation exchange resin column 42 with the cation exchange resin filled, and a valve 54 are installed on a pipe 52 with both ends connected to a circulation pipe 29 for bypassing the valve 49. A heater 61 is installed in the surge tank 31. A valve 56, the mix bed ion exchange resin column 43 with the cation exchange resin and anion exchange resin filled, and a valve 57 are installed on a pipe 55 with both ends connected to the pipe 52 for bypassing the valve 53, the cation exchange resin column 42, and the valve 54. A valve 59, the decomposition apparatus 44, and a valve 60 are installed on a pipe 58 for bypassing the valve 50 and both ends of the pipe 58 is connected to the circulation pipe 29. The decomposition apparatus 44 is internally filled with, for example, ruthenium catalyst supported on an activated carbon surface. The oxidation agent supply apparatus 45 includes a chemical tank 46 filled with an oxidation agent (for example, hydrogen peroxide), a feed pump 47, and an oxidation agent feed pipe 48. The chemical tank 46 is connected to the pipe 58 between the valve 59 and the decomposition apparatus 44 by the oxidation agent feed pipe 48 on which the feed pump 47 is installed. The malonic acid injection apparatus 32 and the oxalic acid injection apparatus 37 are connected to the circulation pipe 29 between the valve 50 and the surge tank 31. The malonic acid injection apparatus 32 includes a chemical tank 33, an injection pump 34, and an injection pipe 36. The chemical tank 33 is connected to the circulation pipe 29 by the injection pipe 36 having the injection pump 34 and a valve 35. The chemical tank 45 is filled with the malonic acid aqueous solution. The oxalic acid injection apparatus 37 includes a chemical tank 38, an injection pump 39, and an injection pipe 41. The chemical tank 38 is connected to the circulation pipe 29 by the injection pipe 41 having the injection pump 39 and a valve 40. The chemical tank 38 is filled with an oxalic acid aqueous solution. The method of chemical decontamination for carbon steel member of a nuclear power plant according to the present embodiment using the chemical decontamination apparatus 28 will be explained based on the procedure shown in FIG. 1. The chemical decontamination apparatus is connected to a piping of executing the chemical decontamination in the BWR plant (step S1). In the state that the operation of the BWR plant is stopped, as mentioned above, one end of the circulation pipe 29 of the chemical decontamination apparatus 28 is connected to the valve 26 installed on the purification system pipe 21 and another end of the circulation pipe 29 is connected to the valve 27 installed on the purification system pipe 21. In the state that the chemical decontamination apparatus 28 is connected to the purification system pipe 21, a closed loop including the circulation pipe 29 and the purification system pipe 21 is formed. A closed plug (not shown) is installed on the valve 26 on the side of the recirculation system pipe 4 so as to prevent the reduction decontaminating solution from flowing into the recirculation pipe 4. Furthermore, a closed plug (not shown) is installed on the side of the regeneration heat exchanger 23 so as to prevent the reduction decontaminating solution from flowing into the regeneration heat exchanger 23. The temperature adjustment of circulation water is performed (step S2). The valves 35 and 40 are set in the closed state, and the open/close valves 48 and 51 and the valves 49, 50, 53 to 57, 59, and 60 are opened. Ion exchange water is supplied into the purification system pipe 21 between the valve 26 and the valve 27, the circulation pipe 29, the pipes 52, 55, and 58, the surge tank 31, the cation exchange resin column 42, the mix bed ion exchange resin column 43, the decomposition apparatus 44, and the circulation pumps 82 and 83 through the water feed pipe (not shown) connected to the circulation pipe 29 and those units are filled with the ion exchange water. The open/close valves 48 and 51 and the valves 49 and 50 are kept opened, and the valves 53 to 57, 59, and 60 are closed, and the circulation pumps 82 and 83 are driven. The ion exchange water existing in the circulation pipe 29 and the surge tank 31 circulates in the closed loop including the circulation pipe 29 and the purification system pipe 21. An electric current is passed through the heater 61 and the ion exchange water in the surge tank 31 is heated by the heater 61. When the temperature of the water circulating in the closed loop rises to a preset temperature (for example, 90° C.) by heating by the heater 61, the heating of the circulating water by the heater 61 is stopped. The temperature of the ion exchange water circulating in the circulation pipe 29 and the purification system pipe 21 is adjusted to 90° C. which is a preset temperature by the heater 61. The malonic acid is injected (step S3). The malonic acid aqueous solution is injected from the malonic acid injection apparatus 32 into the circulation pipe 29. Namely, the valve 35 is opened, and the injection pump 34 is driven. The malonic acid aqueous solution in the chemical tank 33 is injected into the ion exchange water flowing in the circulation pipe 29 through the injection pipe 36. The oxalic acid is injected (step S4). The oxalic acid aqueous solution is injected from the oxalic acid injection apparatus 37 into the circulation pipe 29. Namely, the valve 40 is opened, and the injection pump 39 is driven. The oxalic acid aqueous solution in the chemical tank 38 is injected into the ion exchange water flowing in the circulation pipe 29 through the injection pipe 41. When the malonic acid aqueous solution injected from the malonic acid injection apparatus 32 reaches a connection point of the injection pipe 41 and the circulation pipe 29, the injection of the oxalic acid aqueous solution is performed. An aqueous solution including the malonic acid and the oxalic acid is generated in the circulation pipe 29. The respective concentrations of the malonic acid and oxalic acid in the aqueous solution in the surge tank 31 are suitably measured by an ion chromatograph. When the oxalic acid concentration measured in the aqueous solution in the surge tank 31 becomes 400 ppm, the injection pump 39 is stopped and the valve 40 is closed. By doing this, the injection of the oxalic acid aqueous solution into the circulation pipe 29 is stopped. Also while the oxalic acid aqueous solution is injected, the malonic acid aqueous solution is injected. Though when the malonic acid concentration measured in the aqueous solution in the surge tank 31 becomes 5200 ppm, the injection pump 34 is stopped and the valve 35 is closed. By doing this, the injection of the malonic acid aqueous solution into the circulation pipe 29 is stopped. In the injection of the malonic acid aqueous solution and oxalic acid aqueous solution into the circulation pipe 29, the malonic acid aqueous solution may be injected after the injection of the oxalic acid aqueous solution in place of the injection of the oxalic acid aqueous solution after the injection of the malonic acid aqueous solution. In this case, it is desirable to connect the oxalic acid injection apparatus 37 to the circulation pipe 29 so that it is positioned on the upstream side of the malonic acid injection apparatus 32. By the injection of the malonic acid aqueous solution and oxalic acid aqueous solution into the ion exchange water flowing in the circulation pipe 29, a aqueous solution (reduction decontaminating solution) including the malonic acid with a concentration of 5200 ppm and the oxalic acid with a concentration of, for example, 400 ppm at 90° C. is generated in the surge tank 31. The reduction decontamination is executed (step S5). The aqueous solution including the malonic acid of 5200 ppm and the oxalic acid of 400 ppm at 90° C., by driving the circulation pumps 82 and 83, is supplied into the purification system pipe 21 which is a carbon steel member of the BWR plant through the circulation pipe 29. When flowing in the purification system pipe 21, the aqueous solution including the malonic acid and oxalic acid makes contact with the inner surface of the purification system pipe 21. The oxide film formed on the inner surface of the purification system pipe 21 is dissolved more by the action of the oxalic acid included in the aqueous solution and part of the carbon steel member which is a base metal of the purification system pipe 21 is dissolved by the action of the malonic acid. Therefore, the radioactive nuclide included in the oxide film and the radioactive nuclide included in the base metal in the neighborhood of the inner surface of the purification system pipe 21 are eluted in the aqueous solution including the malonic acid and oxalic acid. The aqueous solution including the malonic acid and oxalic acid includes the ferrous ions and cations of the radioactive nuclide eluted from the oxide film and the base metal of the purification system pipe 21 and is discharged from the purification system pipe 21 into the circulation pipe 29. When the reduction decontamination is started (or when the malonic acid aqueous solution is injected) at step S5, the valves 53 and 54 are opened, and degree of opening of the valve 49 is reduced by adjusting the degree of the opening thereof. Part of the aqueous solution discharged from the purification system pipe 21 into the circulation pipe 29 is introduced to the cation exchange resin column 42. The ferrous ions and cations of the radioactive nuclide which are included in the aqueous solution including the malonic acid and oxalic acid are adsorbed to the cation exchange resin and removed in the cation exchange resin column 42. A radiation detector (not shown) is installed in the neighborhood of a decontamination target area of the purification system pipe 21 wherein the reduction decontamination is executed, and the radiation discharged from the decontamination target area of the purification system pipe 21 is measured by the radiation detector. The dose rate in the reduction execution target area is obtained based on a radiation detection signal outputted from the radiation detector. While the aqueous solution including the malonic acid of 5200 ppm and the oxalic acid of 400 ppm at 90° C. circulates in the circulation pipe 29 and the purification system pipe 21, the reduction decontamination for the inner surface of the purification system pipe 21 is executed until the obtained dose rate reaches a preset dose rate (for example, 0.1 mSv/h) or lower and the ferrous ions eluted in the solution and cations of the radioactive nuclide are removed by the cation exchange resin column 42. When the dose rate of the purification system pipe 21 in the decontamination target area becomes the preset dose rate (for example, 0.1 mSv/h) or lower or when a preset period of time (for example, 6 to 12 hours) elapses from the start of the reduction decontamination for the purification system pipe 21, the reduction decontamination for the purification system pipe 21 finishes. The reduction decontamination agent is decomposed (step S6). When the reduction decontamination finishes, the valves 59 and 60 are opened, and agree of opening of the valve 50 is reduced. Part of the aqueous solution including the malonic acid and oxalic acid which is discharged from the purification system pipe 21 into the circulation pipe 29 is supplied to the decomposition apparatus 44. The malonic acid and oxalic acid are a reduction decontamination agent. By driving the feed pump 47, the hydrogen peroxide is supplied to the decomposition apparatus 44 from the medical fluid tank 46 through the oxidation agent feed pipe 48. The malonic acid and oxalic acid which are included in the aqueous solution are decomposed by the action of the hydrogen peroxide and activated carbon catalyst in the decomposition apparatus 44. The malonic acid (C3H4O4) is decomposed to carbon dioxide and water due to the reaction to the hydrogen peroxide shown in Formula (2). Further, the oxalic acid (C2H2O4) is also decomposed to carbon dioxide and water due to the reaction to the hydrogen peroxide shown in Formula (3).C3H4O4+4H2O2=3CO2+4H2O (2)C2H2O4+H2O2=2CO2+2H2O (3) Thus, when the malonic acid concentration is CMA and the oxalic acid concentration is COA, the reaction equivalent CHP of the hydrogen peroxide can be calculated based on Formula (4).CHP=(4·CMA/104+COA/90)×34 (4) Therefore, when the malonic acid concentration in the aforementioned aqueous solution including the malonic acid and oxalic acid is approx. 5200 ppm and the oxalic acid concentration is 400 ppm, the reaction equivalent of the hydrogen peroxide in the aqueous solution which is introduced into the decomposition apparatus 44, the reaction equivalent being calculated by Formula (4), becomes 6950 ppm. It is desirable to inject the hydrogen peroxide into the aqueous solution in the decomposition apparatus 44 so as to obtain a concentration about 1 to 2 times the reaction equivalent. Thus, when the malonic acid concentration in the aqueous solution including the malonic acid and oxalic acid which is introduced to the decomposition apparatus 44 is approx. 5200 ppm and the oxalic acid concentration is 400 ppm, hydrogen peroxide water is injected so as to control the hydrogen peroxide concentration in the aqueous solution to 6950 to 13900 ppm. The decomposition process of the malonic acid and oxalic acid is continuously executed until the respective concentrations of the malonic acid and oxalic acid in the aqueous solution in the surge tank 31 which are measured by the ion chromatograph become their respective detection limit values (about 10 ppm). When the respective concentrations are reduced to the respective detection limits, the drive of the feed pump 47 is stopped, and the supply of the hydrogen peroxide to the decomposition apparatus 44 is stopped, and the valve 50 is opened fully, and the valves 59 and 60 are closed. The reaction equivalent CHP of the hydrogen peroxide is obtained based on the respective measured values of the malonic acid concentration and oxalic acid concentration in the aqueous solution including the malonic acid and oxalic acid and the injection concentration of the hydrogen peroxide supplied to the decomposition apparatus 44 may be changed by the obtained reaction equivalent CHP. By applying such a method, the quantity of the hydrogen peroxide supplied to the decomposition apparatus 44 can be more reduced than in the case that the hydrogen peroxide concentration supplied to the decomposition apparatus 44 is held at a predetermined concentration. The purification process is executed (step S7). After completion of the decomposition process of the reduction decontamination agent (the malonic acid and oxalic acid), the applying power to the heater 61 installed in the surge tank 31 is stopped and then the cooling apparatus 30 is started. The valves 56 and 57 are opened, and the valves 53 and 54 are closed. In addition, the supply of the aqueous solution to the cation exchange resin column 42 is stopped. A cooling medium is supplied to the cooling apparatus 30 and the aqueous solution discharged from the purification system pipe 21 into the circulation pipe 29 is cooled by the cooling medium in the cooling apparatus 30. The solution is cooled by the cooling medium in the cooling apparatus 30 until it becomes a temperature (for example, room temperature) on a feedable level to the mix bed ion exchange resin column 43. The cooled solution is introduced to the mix bed ion exchange resin column 43. The anions included in the aqueous solution and the cations remaining without removed by the cation exchange resin column 42 are adsorbed to the anion exchange resin and cation exchange resin in the mix bed ion exchange resin column 43 and are removed. The aqueous solution is purified by the mix bed ion exchange resin column 43 while being cooled by the cooling apparatus 30 and circulating in the circulation pipe 29 and the purification system pipe 21. When the electric conductivity of the aqueous solution sampled from the surge tank 31 becomes 100 μS/m or lower, the valve 49 is opened and the valves 56 and 57 are closed. Furthermore, the circulation pumps 82 and 83 are stopped. The chemical decontamination apparatus is detached from the piping for which the chemical decontamination of the BWR plant has been executed (step S8). A valve (not shown) installed on a water discharge pipe (not shown) connected to the circulation pipe 29 is opened and the water existing in the purification system pipe 21 between the valves 26 and 27, the circulation pipe 29, the pipes 52, 55, and 58, the surge tank 31, the cation exchange resin column 42, the mix bed ion exchange resin column 43, the decomposition apparatus 44, and the circulation pumps 82 and 83 is discharged into a storage tank (not shown) through the water discharge pipe. After completion of the water discharge, one end of the circulation pipe 29 is detached from the valve 26 installed on the purification system pipe 21 and another end of the circulation pipe 29 is detached from the valve 27 installed on the purification system pipe 21. After the chemical decontamination apparatus 28 is removed from the purification system pipe 21 which is a chemical decontamination object of the BWR plant, the BWR plant is restarted. According to the present embodiment, the reduction decontamination for the inner surface of the purification system pipe 21 made of carbon steel is executed by using the aqueous solution (reduction decontaminating solution) including the malonic acid (for example, the concentration is 5200 ppm) and the oxalic acid of 400 ppm with a concentration within the range from 50 to 400 ppm, so that the oxide film formed on the inner surface of the purification system pipe 21 is dissolved furthermore by the action of the oxalic acid included in the aqueous solution and the carbon steel which is a base metal of the purification system pipe 21 is dissolved by the action of the malonic acid. The oxalic acid concentration included in the aqueous solution, that is, the reduction decontaminating solution including the malonic acid and oxalic acid is as low as 400 ppm, so that by performing the reduction decontamination for the inner surface of the purification system pipe 21 which is a carbon steel member by the reduction decontaminating solution, the deposition of the ferrous oxalate onto the oxide film formed on the inner surface of the purification system pipe 21 is suppressed and the dissolution of the oxide film by the oxalic acid can be performed efficiently. Furthermore, the carbon steel which is a base metal in the neighborhood of the inner surface of the purification system pipe 21 can be dissolved efficiently by the malonic acid. Therefore, the reduction decontamination efficiency for the inner surface of the purification system pipe 21 which is a carbon steel member can be improved, and the dose rate of the purification system pipe 21 can be reduced more. As a consequence, the exposure of an operator performing the maintenance inspection in the BWR plant can be reduced. In the present embodiment for performing the reduction decontamination for the carbon steel member using the aqueous solution including the malonic acid and the oxalic acid with a concentration within the range from 50 to 400 ppm, the time required for the reduction decontamination in the present embodiment can be shortened than the chemical decontamination method described in Japanese Patent Laid-open No. 2003-90897 because there is no need to decompose the ferrous oxalate deposited on the surface of the carbon steel member in a time period which the reduction decontamination is performed, the ferrous oxalate being decomposed by a formic acid aqueous solution, after the reduction decontamination for the carbon steel member is performed using the oxalic acid aqueous solution, while this was needed in the chemical decontamination method described in Japanese Patent Laid-open No. 2003-90897. A method of chemical decontamination for a carbon steel member of a nuclear power plant according to embodiment 2 which is another preferred embodiment of the present invention will be explained by referring to FIGS. 10, 11, and 12. The method of chemical decontamination for the carbon steel member of the nuclear power plant according to the present embodiment is an example applied to a pipe (for example, the purification system pipe) made of a carbon steel and another pipe (for example, the recirculation system pipe) made of a stainless steel in the BWR plant. The chemical decontamination executed in the present embodiment includes oxidation decontamination and reduction decontamination. A reduction decontamination apparatus 28A used in the method of chemical decontamination for a carbon steel member of a nuclear power plant according to the present embodiment will be explained by referring to FIG. 12. The reduction decontamination apparatus 28A has a structure in which an oxidation decontaminating solution injection apparatus 62 is added to the reduction decontamination apparatus 28 used in the method of chemical decontamination for the carbon steel member of the nuclear power plant according to embodiment 1. The oxidation decontaminating solution injection apparatus 62 includes a chemical tank 63, an injection pump 64, and an injection pipe 66. The chemical tank 63 is connected to the circulation pipe 29 by the injection pipe 66 having the injection pump 64 and a valve 65. The chemical tank 63 is filled with a potassium permanganate aqueous solution which is an oxidation decontaminating solution. A permanganate aqueous solution may be used as an oxidation decontaminating solution in place of the potassium permanganate aqueous solution. The method of chemical decontamination for the carbon steel member of the nuclear power plant according to the present embodiment using the chemical decontamination apparatus 28A will be explained on the basis of the procedure shown in FIG. 10. In the procedure of the method of chemical decontamination for the carbon steel member of the nuclear power plant according to the present embodiment, the processes of steps S9 to S11 are added to the processes of steps S1 to S8 executed in the method of chemical decontamination for the carbon steel member of the nuclear power plant according to embodiment 1. Firstly, the chemical decontamination apparatus is connected to a piping of executing the chemical decontamination in the BWR plant (step S1). In the state that the operation of the BWR plant is stopped, one end (the end on the side of the open/close valve 51) of the circulation pipe 29 of the chemical decontamination apparatus 28A is connected to the valve 8 installed on the recirculation system pipe 4 and another end (the end on the side of the open/close valve 48) of the circulation pipe 29 is connected to the valve 27 installed on the purification system pipe 21. In the state that the chemical decontamination apparatus 28A is connected to the recirculation system pipe 4 and the purification system pipe 21, a closed loop including the circulation pipe 29, the recirculation system pipe 4, and the purification system pipe 21 is formed. A closed plug (not shown) is installed on the valves 8 and 9 on the side of the RPV 2 so as to prevent the oxidation decontamination solution and reduction decontamination solution from flowing into the RPV 2. Furthermore, another closed plug (not shown) is installed on the side of the regeneration heat exchanger 23 so as to prevent the oxidation decontamination solution and reduction decontamination solution from flowing into the regeneration heat exchanger 23. Similarly to embodiment 1, the circulation water temperature adjustment is performed (step S2). In step S2, similarly to embodiment 1, the circulation pipe 29, the recirculation system pipe 4 between the valves 8 and 9, and the purification system pipe 21 between the recirculation system pipe 4 and the valve 26 are internally filled with ion exchange water. In the present embodiment, injection of the potassium permanganate (step S9), oxidation decontamination (step S10) and decomposition of oxidation decontamination agent (step S11) are executed before injecting the malonic acid (step S3) and injecting the oxalic acid (step S4). The oxidation decontamination agent is injected (step S9). In the present embodiment, the potassium permanganate is used as an oxidation decontamination agent. The potassium permanganate aqueous solution (the oxidation decontamination solution) is injected from the oxidation decontaminating solution injection apparatus 62 into the circulation pipe 29. Namely, when the valve 65 is opened and the injection pump 64 is driven, the potassium permanganate aqueous solution in the chemical tank 63 is injected into the ion exchange water flowing in the circulation pipe 29 through the injection pipe 66. The potassium permanganate aqueous solution injected into the ion exchange water is mixed with the ion exchange water in the surge tank 31 and becomes an oxidation decontamination solution. The mixed water of the potassium permanganate aqueous solution and the ion exchange water is referred to as the potassium permanganate aqueous solution (the oxidation decontamination solution) for the sake of convenience. The potassium permanganate aqueous solution is injected from the chemical tank 63 into the circulation pipe 29 so as to control the potassium permanganate concentration of the potassium permanganate aqueous solution which is generated by mixing with the ion exchange water, for example, to 300 ppm existing within a range from 200 to 500 ppm. It may be possible to use a permanganate as an oxidation decontamination agent and inject a permanganate aqueous solution from the chemical tank 63 into the circulation pipe 29. The oxidation decontamination is executed (step S9). The potassium permanganate aqueous solution including the potassium permanganate of 300 ppm at 90° C. is supplied into the recirculation system pipe 4 which is a stainless steel member of the BWR plant through the circulation pipe 29 by driving the circulation pumps 82 and 83. When flowing in the recirculation system pipe 4, the potassium permanganate aqueous solution makes contact with the inner surface of the recirculation system pipe 4. A chromium oxide film formed on the inner surface of the recirculation system pipe 4 is dissolved by the action of the potassium permanganate included in the solution. Therefore, chromate ions included in the chromium oxide film and cations of the radioactive nuclide included in the chromium oxide film are eluted into the potassium permanganate aqueous solution in the recirculation system pipe 4. The potassium permanganate aqueous solution in the recirculation system pipe 4 flows from the recirculation system pipe 4 into the purification system pipe 21 made of carbon steel and soon is discharged into the circulation pipe 29. A ferrous oxide film is formed on the inner surface of the purification system pipe 21 made of carbon steel, though no chromium oxide film is formed. Even if the potassium permanganate aqueous solution flows in the purification system pipe 21, the potassium permanganate does not dissolve the ferrous oxide film formed on the inner surface formed on the inner surface of the purification system pipe 21. The potassium permanganate aqueous solution performs no oxidation decontamination for the inner surface of the purification system pipe 21, and flows in the purification system pipe 21, and is discharged into the circulation pipe 29. The potassium permanganate aqueous solution executes the oxidation decontamination for the inner surface of the recirculation system pipe 4 while circulating in the circulation pipe 29, the recirculation system pipe 4, and the purification system pipe 21 for a predetermined period of time (for example, for 4 to 6 hours). The oxidation decontamination agent is decomposed (step S11). The oxalic acid aqueous solution, similarly to Step S4 of Example 1, is injected into the potassium permanganate aqueous solution flowing in the circulation pipe 29 from the chemical tank 38. The injection of the oxalic acid aqueous solution into the circulation pipe 29 is performed similarly to the injection of the oxalic acid aqueous solution into the circulation pipe 29 in embodiment 1. After the injection of the oxalic acid aqueous solution, the potassium permanganate (oxidation decontamination agent) included in the potassium permanganate aqueous solution is decomposed by the injected oxalic acid (oxidation decontamination agent decomposition process). The decomposition of the potassium permanganate can be confirmed by monitoring color of the aqueous solution in the surge tank 31 by a monitoring camera through a glass window installed on the surge tank 31. The color of the potassium permanganate aqueous solution is purple and when the purple becomes transparent by the injection of the oxalic acid aqueous solution, the potassium permanganate is judged to have been decomposed. When the potassium permanganate is decomposed, the injection of the oxalic acid aqueous solution into the circulation pipe 29 is stopped, and furthermore, the valves 53 and 54 are opened, and by the opening angle adjustment, degree of opening of the valve 49 is reduced by adjustment of degree of the opening. Part of the aqueous solution discharged from the purification system pipe 21 into the circulation pipe 29 is introduced to the cation exchange resin column 42. The processes at step S3 (injection of the malonic acid aqueous solution) and at step S4 (injection of the oxalic acid aqueous solution) are executed similarly to embodiment 1 and the reduction decontamination at step S5 is further executed. The reduction decontamination (step S5) is executed when the aqueous solution (reduction decontaminating solution) including the malonic acid of 5200 ppm and the oxalic acid of 100 ppm at 90° C. is supplied from the circulation pipe 29 into the recirculation system pipe 4 and furthermore, is introduced from the recirculation system pipe 4 to the purification system pipe 21. The reduction decontamination is performed for the respective inner surfaces of the recirculation system pipe 4 and the purification system pipe 21 in contact with the aqueous solution including the malonic acid of 5200 ppm and the oxalic acid of 100 ppm, by the act of the malonic acid and oxalic acid respectively, similarly to the reduction decontamination at step S5 of embodiment 1. It is possible to connect the oxalic acid injection apparatus 37 to the circulation pipe 29 so as to position the oxalic acid injection apparatus 37 on the upstream side of the malonic acid injection apparatus 32, continuously perform the injection of the oxalic acid aqueous solution from the oxalic acid injection apparatus 37 into the circulation pipe 29 even after the oxidation decontamination agent decomposition process finishes (oxalic acid injection at step S4), and perform the injection of the malonic acid at step S3. In the recirculation system pipe 4, the oxide film formed on the inner surface of the recirculation system pipe 4 is dissolved more by the action of the oxalic acid, and part of the stainless steel which is a base metal of the recirculation system pipe 41 is dissolved by the action of the malonic acid. Therefore, the radioactive nuclide included in the oxide film and the radioactive nuclide included in the base metal in the neighborhood of the inner surface of the recirculation system pipe 4 are eluted into the aqueous solution including the malonic acid and oxalic acid. Therefore, the aqueous solution including the malonic acid and oxalic acid flowing in the recirculation system pipe 4 includes the eluted ferrous ions and cations of the radioactive nuclide. Even in the purification system pipe 21, the ferrous ions and cations of the radioactive nuclide are eluted into the solution by the reduction decontamination by the malonic acid and oxalic acid, similarly to embodiment 1. The aqueous solution including the ferrous ions and cations of the radioactive nuclide and including the malonic acid and oxalic acid is discharged from the purification system pipe 21 into the circulation pipe 29 and is introduced to the cation exchange resin column 42. The ferrous ions and cations of the radioactive nuclide are adsorbed to the cation exchange resin in the cation exchange resin column 42 and are removed. While the aqueous solution including the malonic acid of 5200 ppm and the oxalic acid of 100 ppm is circulated in the closed loop including the circulation pipe 29, the recirculation system pipe 4, and the purification system pipe 21, the aqueous solution executes the reduction decontamination for the inner surfaces of the recirculation system pipe 4 and the purification system pipe 21. The ferrous ions and cations of the radioactive nuclide which are generated by the reduction decontamination are removed by the cation exchange resin column 42. When the dose rate in each decontamination object area of the recirculation system pipe 4 and the purification system pipe 21 becomes a preset dose rate (for example, 0.1 mSv/h) or lower or when a preset period of time (for example, for 6 to 12 hours) elapses after the reduction decontamination is started, the reduction decontamination for the recirculation system pipe 4 and the purification system pipe 21 finishes. Thereafter, the decomposition of the reduction decontamination agent (step S6), purification process (step S7), and the removal of the chemical decontamination apparatus (step S8) are executed successively, similarly to embodiment 1. After the chemical decontamination apparatus 28A is removed from the purification system pipe 21 which is a chemical decontamination target used in the BWR plant, the BWR plant is restarted. The present embodiment can obtain each effect generated in embodiment 1. Furthermore, according to the present embodiment, the chemical decontamination can be performed simultaneously for the recirculation system pipe 4 made of stainless steel and the purification system pipe 21 made of carbon steel, so the time required for the chemical decontamination can be shortened. When the chemical decontamination is performed separately for the recirculation system pipe 4 and the purification system pipe 21 using the chemical decontamination apparatus 28A, the operation of the connection and removal of both the chemical decontamination apparatus 28 for the purification system pipe 21 and the chemical decontamination apparatus 28A for the recirculation system pipe 4 needs to be performed and furthermore, the circulation water temperature adjustment at step S2 needs to be performed both for the chemical decontamination apparatus 28 and for the chemical decontamination apparatus 28A. In the present embodiment simultaneously performing the chemical decontamination for the recirculation system pipe 4 and the purification system pipe 21 using the chemical decontamination apparatus 28A, the overlapped operations of the connection and removal of the chemical decontamination apparatuses 28 and 28A which are generated when the chemical decontamination is performed separately for the recirculation system pipe 4 and the purification system pipe 21 can be integrated into one. Therefore, according to the present embodiment, the time required for the chemical decontamination can be shortened. It is possible to connect one end (the end on the side of the open/close valve 51) of the circulation pipe 29 of the chemical decontamination apparatus 28A to the valve 27 installed on the purification system pipe 21 and connect another end (the end on the side of the open/close valve 48) of the circulation pipe 29 to the valve 8 installed on the recirculation system pipe 4. In this case, in step S9 (oxidation decontamination), the potassium permanganate aqueous solution (oxidation decontaminating solution) is supplied from the circulation pipe 29 into the purification system pipe 21, is introduced from the purification system pipe 21 into the recirculation system pipe 4, and is discharged from the recirculation system pipe 4 into the circulation pipe 29. Further, in step S5 (reduction decontamination), the aqueous solution (reduction decontaminating solution) including the malonic acid of 5200 ppm and the oxalic acid of 100 ppm at 90° C. is also supplied from the circulation pipe 29 into the purification system pipe 21, is introduced from the purification system pipe 21 into the recirculation system pipe 4, and is discharged from the recirculation system pipe 4 into the circulation pipe 29. Even if the flowing direction of the oxidation decontaminating solution or reduction decontaminating solution is changed, the oxidation decontamination for the inner surface of the recirculation system pipe 4 or the reduction decontamination for the inner surfaces of the recirculation system pipe 4 and the purification system pipe 21 is performed. A method of chemical decontamination for a carbon steel member of a nuclear power plant according to embodiment 3 which is other preferable embodiment of the present invention will be explained by referring to FIGS. 13 and 14. The method of chemical decontamination for the carbon steel member of the nuclear power plant according to the present embodiment is an example applied to a carbon steel member detached from the BWR plant by the exchange or decommissioning action, for example, a pipe made of carbon steel. A chemical decontamination apparatus 28B used in the method of chemical decontamination for the carbon steel member of the nuclear power plant according to the present embodiment will be explained by referring to FIG. 13. The reduction decontamination apparatus 28B has a structure in which an oxygen gas feed apparatus 66 is added to the reduction decontamination apparatus 28 used in the method of chemical decontamination for the carbon steel member of the nuclear power plant according to embodiment 1; one end of the circulation pipe 29 is connected to the surge tank 31 in the reduction decontamination apparatus 28; and furthermore, the other end of the circulation pipe 29 is connected to the surge tank 31 to thereby form a closed loop including the circulation pipe 29 and the surge tank 31. The oxygen gas feed apparatus 66 includes an oxygen gas cylinder 67 and an oxygen gas feed pipe 68. One end portion of the oxygen gas feed pipe 68 is connected to the oxygen gas cylinder 67 and the other end of the oxygen gas feed pipe 68 is inserted into the surge tank 31. Many injection outlets (not shown) jetting oxygen gas are formed at the other end of the oxygen gas feed pipe 68 existing in the surge tank 31. An open/close valve 69 and a pressure reducing valve 70 are installed on the oxygen gas feed pipe 68 outside the surge tank 31. The other structure of the chemical decontamination apparatus 28B is the same as the chemical decontamination apparatus 28. Further, the chemical decontamination apparatus 28B has one circulation pump 82 installed on the circulation pipe 29 but no circulation pump 83. The method of chemical decontamination for the carbon steel member of the nuclear power plant according to the present embodiment using the chemical decontamination apparatus 28B will be explained based on the procedure shown in FIG. 13. In the method of chemical decontamination for the carbon steel member according to the present embodiment, each process at steps S12 and S14 is performed respectively in place of each process at steps S1 and S8 in the procedure of the method of chemical decontamination for the carbon steel member according to embodiment 1 and furthermore, the procedure with the process at step S13 added is executed. Each process at steps S2 to S4 and S5 to S7 which is executed by the method of chemical decontamination according to the present embodiment is the same as each process executed by the method of chemical decontamination according to embodiment 1. The decontamination target is put in the decontamination bath (step S12). The surge tank 31 also has a function of the decontamination bath. To exchange with a new pipe made of carbon steel, a pipe 84 which is a decontamination object detached from the BWR plant, the pipe being made of carbon steel, is transferred to the position of the surge tank 31 by transport equipment 71 and is put in the surge tank 31 with the upper end opened. The pipes made of carbon steel and the equipment made of carbon steel other than the pipe 84 removed from the BWR plant are put in the surge tank 31 by the transport equipment 71. After a plurality of decontamination objects are put in the surge tank 31, the surge tank 31 is attached with a cover and the surge tank 31 is sealed up. The circulation water temperature adjustment (step S2), the malonic acid injection (step S3), and the oxalic acid injection (step S4) are performed similarly to embodiment 1. Each process at steps S3 and S4 is executed, thus the aqueous solution including the malonic acid of 12300 ppm and the oxalic acid of 100 ppm at 90° C. is generated in the surge tank 31. In the present embodiment, it is desirable to remove the radioactive nuclide from the decontamination object, such as the pipe 84, put in the surge tank 31. Therefore, there is no need to consider damage of the equipment installed in the BWR plant as far as possible and as in embodiments 1 and 2, so that in the injection of the malonic acid aqueous solution into the circulation pipe 29 at step S3, it is desirable to control the malonic acid concentration generated in the surge tank 31 so as to reduce the pH of the solution to 1.8 or lower. Thus, the malonic acid aqueous solution is injected into the circulation pipe 29 from the malonic acid injection apparatus 32 so as to control the malonic acid concentration, for example, to 12300 ppm. When the malonic acid concentration of the aqueous solution becomes 12300 ppm, the injection of the malonic acid aqueous solution into the circulation pipe 29 is stopped. Further, when the oxalic acid concentration of the aqueous solution becomes 100 ppm, the injection of the oxalic acid aqueous solution into the circulation pipe 29 is stopped. Oxygen gas is injected (step S13). The oxygen gas in the oxygen gas cylinder 67 is introduced through the oxygen gas feed pipe 68 by opening the open/close valve 69 and is jetted into the aqueous solution including the malonic acid of 12300 ppm and the oxalic acid of 100 ppm at 90° C. in the surge tank 31 from the plurality of injection outlets formed at the end portion of the oxygen gas feed pipe 68 existing in the surge tank 31. Degree of opening of the pressure reducing valve 70 is adjusted so as to control the oxygen gas pressure jetted from each injection outlet of the oxygen gas feed pipe 68 to within the range from 0.1 to 1.0 MPa. In the present embodiment, the degree of opening of the pressure reducing valve 70 is adjusted so as to control the jet pressure of oxygen gas to, for example, 0.5 MPa. The injected oxygen gas is dissolved by the aqueous solution including the malonic acid and oxalic acid. In the reduction decontamination (step S5), the aqueous solution including the malonic acid of 12300 ppm, the oxalic acid of 100 ppm, and oxygen at 90° C. makes contact with each surface of the pipes 84 in the surge tank 31 and the reduction decontamination for the pipes 84 is performed. Since the circulation pump 82 is being driven, the aqueous solution in the surge tank 31 is discharged from the surge tank 31 into the circulation pipe 29, circulates once in the circulation pipe 29 forming the closed loop, and is returned into the surge tank 31. The valves 53 and 54 are opened and degree of opening of the valve 49 is reduced by adjustment of the degree of opening thereof. Part of the aqueous solution discharged from the surge tank 31 into the circulation pipe 29 is introduced to the cation exchange resin column 42. The ferrous oxide formed on the surface of the pipes 84 is dissolved by the reduction decontamination for the pipe 84 by the aqueous solution including the malonic acid of 12300 ppm, the oxalic acid of 100 ppm, and oxygen at 90° C., similarly to embodiment 1 and part of the carbon steel which is a base metal of each pipe 84 is dissolved. Similarly to embodiment 1, the ferrous ions and cations of the radioactive nuclide are eluted into the aqueous solution in the surge tank 31. The ferrous ions and cations of the radioactive nuclide included in the aqueous solution introduced to the cation exchange resin column 42 are adsorbed to the cation exchange resin in the cation exchange resin column 42 and are removed. The aqueous solution including the malonic acid of 12300 ppm, the oxalic acid of 100 ppm, and oxygen at 90° C. passes through the cation exchange resin column 42 while circulating in the surge tank 31 and the circulation pipe 29. The reduction decontamination for the pipes 84 in the surge tank 31 is performed by the circulating aqueous solution. The injection of oxygen gas into the aqueous solution including the malonic acid and oxalic acid in the surge tank 31 by the oxygen gas feeder 66 is performed continuously while the reduction decontamination for the pipes 84 is performed. When the dose rate of the pipe 84 obtained based on the radiation detection signal outputted from a radiation detector disposed in the neighborhood of the surge tank 31 becomes the preset dose rate (for example, 0.1 mSv/h) or lower or when a preset period of time (for example, for 6 to 12 hours) from the start of the reduction decontamination elapses, the reduction decontamination for the pipe 84 finishes. After completion of the reduction decontamination, the decomposition of the reduction decontamination agent (step S6) and the purification process (step S7) are performed, similarly to embodiment 1, while the aqueous solution is permitted to circulate in the surge tank 31 and the circulation pipe 29. After completion of the purification process, the decontamination object is taken out (step S14) from the decontamination bath. The surge tank 31 which is a decontamination bath is opened, and the pipes 84 with the reduction decontamination finished are taken out from the surge tank 31 using the transport equipment 71. After the pipes 84 with the reduction decontamination finished are taken out, the reduction decontamination for a new chemical decontamination objects are executed by repeating steps S12, S2 to S4, S13, S5 to S7, and S14. The present embodiment can obtain each effect generated in embodiment 1. Furthermore, the present embodiment can perform the reduction decontamination even for the carbon steel members taken out from the nuclear power plant. Further, in the present embodiment, it is possible to execute the reduction decontamination for the pipes 84 in the surge tank 31 without injecting oxygen gas into the aqueous solution including the malonic acid of 12300 ppm and the oxalic acid of 100 ppm at 90° C. and by permitting the solution with no oxygen gas injected to circulate in the surge tank 31 with the pipes 84 put and the circulation pipe 29. In cases where many carbon steel members (for example, the pipes 84) require reduction decontamination like decommissioning and remodeling of the BWR plant, the reduction decontamination for individual carbon steel members is performed as described below using the chemical decontamination apparatus 28B. In Step S12, the pipes 84 are put in the surge tank 31 and each process at steps S2 to S4, S13, and S5 is executed successively. When the reduction decontamination process at step S5 finishes, taking out the decontamination object from the surge tank 31 (step S14) is executed. A plurality of pipes 84 with the reduction decontamination finished are taken out from the surge tank 31 by the transport equipment 71 and is transferred to a washing apparatus 72 (refer to FIG. 15) installed separately from the chemical decontamination apparatus 28B. These pipes 84 are washed by the washing apparatus 72. A structure of the washing apparatus 72 will be explained below by referring to FIG. 15. The washing apparatus 72 includes a washing bath 73, a circulation pump 74, and a mix bed ion exchange resin column 75. One end portion of a circulation pipe 76 is connected to the washing bath 73 and another end portion of the circulation pipe 76 is also connected to the washing bath 73. A closed loop is formed by the washing bath 73 and the circulation pipe 76. The circulation pump 74 and the mix bed ion exchange resin column 75 are installed on the circulation pipe 76. The mix bed ion exchange resin column 75 is internally filled with the cation exchange resin and anion exchange resin. The pipes 84 taken out from the surge tank 31 and transferred by the transport equipment 71 are taken off the cap from upper end and are put in the washing bath 73, from an upper end of which a cap is taken off and which is filled with the washing water. After the plurality of pipes 84 reduction-decontaminated are put in the washing bath 73, the cap is attached to the upper end of the washing bath 73, and the washing bath 73 is sealed up. The circulation pump 74 is driven and the washing water in the washing bath 73 is circulated through the circulation pipe 76 and the mix bed ion exchange resin column 75. The pipes 84 in the washing bath 73 are washed by the circulating washing water. The radioactive nuclide adhered to the pipes 84 moves from the pipes 84 to the washing water, and is adsorbed to the ion exchange resin in the mix bed ion exchange resin column 75, and is removed from the washing water. When the dose rate of the pipes 84 in the washing bath 73 becomes the preset dose rate (for example, 0.1 mSv/h) or lower, the washing for the pipes 84 in the washing bath 73 finishes. The pipes 84 which have been washed and become the preset dose rate or lower are taken out from the washing apparatus 73. A plurality of pipes 84 to be newly reduction-decontaminated are put in the surge tank 31 of the chemical decontamination apparatus 28B from which the reduction-decontaminated pipes 84 have been taken out (step S12). Each process at Steps S12, S2 to S4, S13, and S5 is executed using the chemical decontamination apparatus 28B and the reduction decontamination is executed for the pipes 84 in the surge tank 31. After completion of the reduction decontamination at Step S5, as mentioned above, the plurality of pipes 84 for which the reduction decontamination has been executed are taken out from the surge tank 31. These pipes 84 are washed by the washing apparatus 72. A new plurality of pipes 84 to be reduction-decontaminated are put in the surge tank 31 and as mentioned above, the reduction decontamination is executed for these pipes 84. The reduction decontamination in the surge tank 31 is performed continuously until the pipes 84 which are a reduction decontamination object are exhausted. The aqueous solution (reduction decontaminating solution) including the malonic acid of 12300 ppm and the oxalic acid of 100 ppm which exists in the surge tank 31 and the circulation pipe 29 is reused when the reduction decontamination is performed for the new pipes 84 in the surge tank 31. After the reduction decontamination (step S5) for the last plurality of pipes 84 in the surge tank 31 finishes, the decomposition of the reduction decontaminating agent (step S6) and the purification process (step S7) are executed successively with those pipes 84 put in the surge tank 31 and furthermore, the take-out of the decontamination objects (step S14) are executed. The washing of the reduction-decontaminated pipes 84 taken out from the surge tank 31 is performed using the washing apparatus 72, so that when there are many decontamination objects subject to the reduction decontamination, there is no need to perform the decomposition of the reduction decontaminating agent (step S6) and the purification process (Step S7) whenever the reduction decontamination in the surge tank 31 finishes, enabling efficient reduction decontamination for the washing object. Therefore, the time required for the reduction decontamination when there are many decontamination objects subject to the reduction decontamination can be shortened. Further, the malonic acid and oxalic acid included in the reduction decontaminating solution are not decomposed whenever the reduction decontamination finishes, so that the reduction decontaminating solution including the malonic acid and oxalic acid can be reused. The oxygen gas feed apparatus 66 used in the reduction decontamination apparatus 28B may be changed to an oxygen gas feed apparatus 66A shown in FIG. 16. In the oxygen gas feed apparatus 66A, a circulation pump 79 and a micro-bubble generator 78 are installed on a pipe 80 with one end portion thereof connected to the bottom of the surge tank 31. Another end portion of the pipe 80 is inserted into the surge tank 31. A valve 81 is opened and the circulation pump 79 is driven. The aqueous solution including the malonic acid of 12300 ppm and the oxalic acid of 100 ppm at 90° C. in the surge tank 31 is supplied to the micro-bubble generator 78 through the pipe 80. The micro-bubble generator 78 supplies oxygen-included gas of a micron order (for example, air) to the aqueous solution. The aqueous solution including the oxygen-included gas of a micron order (micro bubbles) is injected into the aqueous solution in the surge tank 31 through the pipe 80. Therefore, the aqueous solution including the malonic acid of 12300 ppm, the oxalic acid of 100 ppm at 90° C., and the oxygen-included gas of a micron order makes contact with the pipes 84 in the surge tank 31. In the oxygen-included gas of a micron order, the contact solution area for the bubble volume is large, so that the oxygen included in the oxygen-included gas of a micron order is dissolved easily in the aqueous solution in the surge tank 31. Thus, the reduction decontamination effects can be improved by a small quantity of the oxygen-included gas. The oxygen gas feed apparatus 66 or 66A can be applied to the chemical decontamination apparatuses 28 and 28A. Therefore, even in embodiments 1 and 2, the reduction decontamination can be executed for the purification system pipe 21 and the recirculation system pipe 4 using the aqueous solution including the malonic acid, oxalic acid, and oxygen. In each of embodiments 1 to 3, oxygen gas or oxygen-included gas of a micron order is injected into the water which is a reduction decontaminating solution in the surge tank 31, so that the dissolution of oxygen into the aqueous solution is performed easily compared with the case that the oxygen gas or oxygen-included gas of a micron order is injected into the circulation pipe 29. 2: reactor pressure vessel, 4: primary loop recirculation system piping, 5: recirculation pump, 10: turbine, 13: water feed pipe, 21: purification system pipe, 28, 28A, 28B: chemical decontamination apparatus, 31: surge tank, 32: malonic acid injection apparatus, 33, 38, 46, 63: chemical tank, 34, 39, 64: injection pump, 37: oxalic acid injection apparatus, 42: cation exchange resin column, 43, 75: mix bed ion exchange resin column, 45: oxidation agent supply apparatus, 62: oxidation decontaminating solution injection apparatus, 66, 66A: oxygen gas feed apparatus, 78: micro-bubble generator. |
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abstract | An ion implantation system comprising an ion source that generates an ion beam along a beam path, a mass analyzer component downstream of the ion source that performs mass analysis and angle correction on the ion beam, a resolving aperture electrode comprising at least one electrode downstream of the mass analyzer component and along the beam path having a size and shape according to a selected mass resolution and a beam envelope, a deflection element downstream of the resolving aperture electrode that changes the path of the ion beam exiting the deflection element, a deceleration electrode downstream of the deflection element that decelerates the ion beam, a support platform within an end station for retaining and positioning a workpiece which is implanted with charged ions, and wherein the end station is mounted approximately eight degrees counterclockwise so that the deflected ion beam is perpendicular to the workpiece. |
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043274439 | abstract | Capillary liquid fuel elements, created by the method of confining a liquid fuel in horizontal capillary troughs, are employed to create the core of a nuclear reactor to generate useful heat energy. The reactor incorporates the inherent advantages of a liquid fuel reactor: high specific power, high fuel burnup, inherent safety, ease of control, and simple fuel preparation, processing, reprocessing, and handling. The reactor in addition, has advantages unavailable in other liquid fuel reactors: high breeding potential, low delayed neutron loss, low pumping power requirement, low fission material external holdup, direct fuel-coolant heat exchange capability, and low construction material cost. |
abstract | A variable stop apparatus for arrangement between an X-ray source and an object to be measured in a CT-scanner and a CT-scanner including the variable stop apparatus are provided. The variable stop apparatus includes a stop carrier that is pivotable about a pivot axis. The stop carrier has at least two stops. The at least two stops are in each case configured to be brought into a predetermined angular position by pivoting the stop carrier. The at least two stops are arranged at different longitudinal positions with respect to a longitudinal direction that is defined by the pivot axis. |
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claims | 1. A method for manipulating a workpiece, comprising:engaging a workpiece with a workpiece support in a process chamber under high vacuum;providing an air bearing between a spherical bearing portion connected to the workpiece support and a housing portion disposed in a wall of the process chamber; andscanning the workpiece through an ion beam by manipulating the spherical bearing portion;wherein the spherical bearing portion is configured to enable the workpiece support to be moved with four degrees of freedom, andwherein providing the air bearing comprises pumping gas into a space between the spherical bearing portion and the housing portion. 2. The method of claim 1, wherein scanning the workpiece comprises performing an isocentric scan of the workpiece. 3. The method of claim 1, wherein scanning the workpiece comprises translating the workpiece support with respect to the spherical bearing portion. 4. The method of claim 3, wherein translating the workpiece support comprises rotating the workpiece support with respect to the spherical bearing. 5. The method of claim 1, further comprising maintaining a seal between the process chamber on a first side of the spherical bearing portion and ambient pressure on a second side of the spherical bearing portion, wherein maintaining the seal comprises pumping gas through at least one pumping groove disposed in the housing portion. 6. The method of claim 5, further comprising providing one or more spherical air pads between the spherical bearing portion and the housing portion, the one or more spherical air pads disposed on an atmospheric side of the housing. 7. The method of claim 6, further comprising providing a second air bearing between the spherical bearing portion and a shaft portion, a distal end of the shaft portion disposed in the process chamber, the distal end coupled to the workpiece support, wherein maintaining the seal comprises pumping gas through at least one pumping groove disposed in the spherical bearing portion. 8. A method for manipulating a workpiece, comprising:engaging a workpiece with a workpiece support in a process chamber;providing an air bearing between a spherical bearing portion connected to the workpiece support and a housing portion disposed in a wall of the process chamber; andscanning the workpiece through an ion beam by manipulating the spherical bearing portion via a shaft portion;wherein scanning comprises translating the shaft portion with respect to the spherical bearing portion;wherein the spherical bearing portion and the shaft portion are configured to enable to the workpiece support to be moved with four degrees of freedom; andwherein providing the air bearing comprises pumping gas into a space between the spherical bearing portion and the housing portion. 9. The method of claim 8, wherein scanning the workpiece comprises performing an isocentric scan of the workpiece. 10. The method of claim 8, wherein scanning the workpiece comprises translating the workpiece support along a longitudinal axis of the shaft portion. 11. The method of claim 10, wherein translating the workpiece support comprises rotating the workpiece support with respect to the spherical bearing. 12. The method of claim 8, further comprising maintaining a seal between the process chamber on a first side of the spherical bearing portion and ambient pressure on a second side of the spherical bearing portion, wherein maintaining a seal comprising pumping gas through at least one pumping groove disposed in the housing portion. 13. The method of claim 12, further comprising providing one or more spherical air pads between the spherical bearing portion and a housing portion, the one or more spherical air pads disposed on an atmospheric side of the housing portion. 14. A method for manipulating a workpiece, comprising:engaging a workpiece with a workpiece support in a process chamber;providing an air bearing between a spherical bearing portion connected to the workpiece support and a housing portion disposed in a wall of the process chamber; andscanning the workpiece through an ion beam by manipulating the spherical bearing portion received within a spherical recess in a housing portion, the housing portion connected to a wall of the process chamber;wherein the spherical bearing portion and a shaft portion are configured to enable to the workpiece support to be moved with four degrees of freedom, andwherein providing the air bearing comprises pumping gas into a space between the spherical bearing portion and the housing portion. 15. The method of claim 14, wherein scanning the workpiece comprises performing an isocentric scan of the workpiece. 16. The method of claim 14, wherein scanning the workpiece comprises translating the workpiece support along a longitudinal axis of the shaft portion coupled to the spherical bearing portion. 17. The method of claim 16, wherein translating the workpiece support comprises rotating the workpiece support with respect to the spherical bearing. 18. The method of claim 14, further comprising maintaining a seal between the process chamber on a first side of the spherical bearing portion and ambient pressure on a second side of the spherical bearing portion, wherein maintaining the a seal comprises pumping gas through at least one pumping groove disposed in the housing portion. 19. The method of claim 18, further comprising providing one or more spherical air pads between the spherical bearing portion and the housing portion, the one or more spherical air pads disposed on an atmospheric side of the housing portion. 20. The method of claim 19, further comprising providing a second air bearing between the spherical bearing portion and a shaft portion, a distal end of the shaft portion disposed in the process chamber, the distal end coupled to the workpiece support, wherein maintaining a seal comprises pumping gas through at least one pumping groove disposed in the spherical bearing portion. |
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description | This application is based upon and claims the benefit of priority from Japanese Patent Application No. 2013-147885, filed on Jul. 16, 2013, Japanese Patent Application No. 2013-147886, filed on Jul. 16, 2013, Japanese Patent Application No. 2013-252419, filed on Dec. 5, 2013, and Japanese Patent Application No. 2013-252420, filed on Dec. 5, 2013; the entire contents of which are incorporated herein by reference. Embodiments described herein relate generally to a radiation detector, a scintillator panel, and a method for manufacturing the same. An X-ray detector can be realized as a flat radiation detector based on solid-state imaging elements such as active matrix, CCD, and CMOS. Such an X-ray detector is drawing attention as a new-generation X-ray image detector for diagnosis. A radiographic image or real-time X-ray image is outputted as digital signals by irradiating this X-ray detector with X-rays. The X-ray detector includes a photoelectric conversion substrate for converting light to electrical signals, and a scintillator layer in contact with the photoelectric conversion substrate. The scintillator layer converts externally incident X-rays to light. The light converted from incident X-rays in the scintillator layer reaches the photoelectric conversion substrate and is converted to electric charge. This charge is read as an output signal and converted to digital image signals in e.g. a prescribed signal processing circuit. The scintillator layer may be made of CsI, which is a halide. In this case, incident X-rays cannot be converted to visible light by CsI alone. Thus, as in commonly-used phosphors, an activator is contained to activate excitation of light in response to incident X-rays. In the X-ray detector, the light reception sensitivity of the photoelectric conversion substrate has a peak wavelength around 400-700 nm in the visible range. Thus, in the case where the scintillator layer is made of CsI, Tl is used as an activator. Then, the light excited by incident X-rays has a wavelength around 550 nm. The scintillator layer may be made of a phosphor containing Tl as an activator in CsI, which is a halide. In this case, as in commonly-used phosphors containing an activator, the characteristics of the scintillator layer are significantly affected by the concentration and concentration distribution of Tl serving as an activator. In the X-ray detector including a scintillator layer containing an activator, lack of optimization of the concentration and concentration distribution of the activator incurs characteristics degradation of the scintillator layer. This affects the sensitivity (light emission efficiency) and residual image (the phenomenon in which the subject image of the X-ray image at the (n−1)-th or earlier time remains in the X-ray image at the n-th time) related to the light emission characteristics of the scintillator layer. For instance, in diagnosis using X-ray images, the radiography condition significantly varies with subjects (incident X-rays at a dose of approximately 0.0087-0.87 mGy, because the X-ray transmittance varies with body regions). This may cause a significant difference in the dose of incident X-rays between the (n−1)-th X-ray image and the n-th X-ray image. Here, if the dose of incident X-rays in the (n−1)-th X-ray image is greater than that in the n-th X-ray image, the light emission characteristics of the scintillator layer in the non-subject part of the (n−1)-th X-ray image is changed by the great energy of incident X-rays. This influence remains also in the n-th X-ray image and produces a residual image. In diagnosis using X-ray images, the residual image characteristic is more important than other characteristics of the scintillator layer such as sensitivity (light emission efficiency) and resolution (MTF). Conventionally, there have been proposals for defining the concentration and concentration distribution of the activator of the scintillator layer for the purpose of improving sensitivity (light emission efficiency) and resolution (MTF). Conventional proposals for characteristics improvement of the scintillator layer largely relate to sensitivity (light emission efficiency) and resolution (MTF). There have been few proposals related to overall characteristics improvement including the residual image characteristic. The problem to be solved by the invention is to provide a radiation detector, a scintillator panel, and a method for manufacturing the same capable of improving overall characteristics including the residual image characteristic of the scintillator layer. According to the embodiment, a radiation detector includes a photoelectric conversion substrate converting light to an electrical signal and a scintillator layer being in contact with the photoelectric conversion substrate and converting externally incident radiation to light. The scintillator layer is made of a phosphor containing Tl as an activator in CsI, which is a halide. A concentration of the activator in the phosphor is 1.6 mass %±0.4 mass %, and a concentration distribution of the activator in an in-plane direction and a film thickness direction is within ±15%. Various Embodiments will be described hereinafter with reference to FIG. 1 to FIG. 19. In FIG. 1 to FIG. 4, the basic configuration of a radiation detector 1 is described with reference to first to fourth structure examples. FIG. 5 shows an equivalent circuit diagram of the basic configuration. First, a first structure example of the X-ray detector 1 as a radiation detector is described with reference to FIG. 1 and FIG. 5. As shown in FIG. 1, the X-ray detector 1 is an indirect-type flat X-ray image detector. The X-ray detector 1 includes a photoelectric conversion substrate 2. The photoelectric conversion substrate 2 is an active matrix photoelectric conversion substrate for converting visible light to electrical signals. The photoelectric conversion substrate 2 includes a support substrate 3. The support substrate 3 is an insulating substrate formed from a translucent glass shaped like a rectangular plate. On the surface of the support substrate 3, a plurality of pixels 4 are arranged with spacing from each other in a two-dimensional matrix. Each pixel 4 includes a thin film transistor (TFT) 5 as a switching element, a charge storage capacitor 6, a pixel electrode 7, and a photoelectric conversion element 8 such as a photodiode. As shown in FIG. 5, a plurality of control electrodes 11 are wired on the support substrate 3. The control electrode 11 is a control line along the row direction of the support substrate 3. The plurality of control electrodes 11 are each located between the pixels 4 on the support substrate 3 and spaced in the column direction of the support substrate 3. The gate electrodes 12 of the thin film transistors 5 are electrically connected to these control electrodes 11. A plurality of read electrodes 13 along the column direction of the support substrate 3 are wired on the support substrate 3. The plurality of read electrodes 13 are each located between the pixels 4 on the support substrate 3 and spaced in the row direction of the support substrate 3. The source electrodes 14 of the thin film transistors 5 are electrically connected to these read electrodes 13. The drain electrode 15 of the thin film transistor 5 is electrically connected to each of the charge storage capacitor 6 and the pixel electrode 7. As shown in FIG. 1, the gate electrode 12 of the thin film transistor 5 is formed like an island on the support substrate 3. An insulating film 21 is stacked on the support substrate 3 including the gate electrode 12. The insulating film 21 covers each gate electrode 12. A plurality of island-shaped semi-insulating films 22 are stacked on the insulating film 21. The semi-insulating film 22 is formed from semiconductor and functions as a channel region of the thin film transistors 5. The semi-insulating films 22 are opposed to the respective gate electrodes 12 and cover these gate electrodes 12. That is, the semi-insulating films 22 are provided on the respective gate electrodes 12 via the insulating film 21. The source electrode 14 and the drain electrode 15 are each formed like an island on the insulating film 21 including the semi-insulating films 22. The source electrode 14 and the drain electrode 15 are insulated from and not electrically connected to each other. The source electrode 14 and the drain electrode 15 are provided on opposite sides on the gate electrode 12. One end part of the source electrode 14 and the drain electrode 15 is stacked on the semi-insulating film 22. As shown in FIG. 5, the gate electrode 12 of each thin film transistor 5 is electrically connected to a common control electrode 11 together with the gate electrodes 12 of the other thin film transistors 5 located on the same row. Furthermore, the source electrode 14 of each thin film transistor 5 is electrically connected to a common read electrode 13 together with the source electrodes 14 of the other thin film transistors 5 located on the same column. As shown in FIG. 1, the charge storage capacitor 6 includes an island-shaped lower electrode 23 formed on the support substrate 3. The insulating film 21 is stacked on the support substrate 3 including the lower electrode 23. The insulating film 21 extends from above the gate electrodes 12 of the thin film transistors 5 to above the lower electrodes 23. Furthermore, an island-shaped upper electrode 24 is stacked on the insulating film 21. The upper electrode 24 is opposed to the lower electrode 23 and covers the lower electrode 23. That is, the upper electrode 24 is provided on each lower electrode 23 via the insulating film 21. The drain electrode 15 is stacked on the insulating film 21 including the upper electrode 24. The other end part of the drain electrode 15 is stacked on the upper electrode 24 and electrically connected to the upper electrode 24. An insulating layer 25 is stacked on the insulating film 21 including the semi-insulating films 22, the source electrodes 14, and the drain electrodes 15 of the thin film transistors 5 and the upper electrodes 24 of the charge storage capacitors 6. The insulating layer 25 is formed from e.g. silicon oxide (SiO2) around each pixel electrode 7. A through hole 26 is opened in part of the insulating layer 25. The through hole 26 is a contact hole communicating with the drain electrode 15 of the thin film transistor 5. An island-shaped pixel electrode 7 is stacked on the insulating layer 25 including the through hole 26. The pixel electrode 7 is electrically connected to the drain electrode 15 of the thin film transistor 5 through the through hole 26. A photoelectric conversion element 8 such as a photodiode for converting visible light to electrical signals is stacked on each pixel electrode 7. A scintillator layer 31 is formed on the surface of the photoelectric conversion substrate 2 where the photoelectric conversion element 8 is formed. The scintillator layer 31 converts radiation such as X-rays to visible light. The scintillator layer 31 is formed by depositing a high-brightness fluorescent material in a columnar shape on the photoelectric conversion substrate 2 by vapor phase growth technique such as vacuum evaporation technique, sputtering technique, and CVD technique. The high-brightness fluorescent material is a phosphor such as a halide including cesium iodide (CsI) and an oxide-based compound including gadolinium oxysulfide (GOS). The scintillator layer 31 is formed to have a columnar crystal structure such that a plurality of strip-shaped columnar crystals 32 are formed in the in-plane direction of the photoelectric conversion substrate 2. A reflective layer 41 is stacked on the scintillator layer 31. The reflective layer 41 enhances the utilization efficiency of visible light converted in the scintillator layer 31. A protective layer 42 is stacked on the reflective layer 41. The protective layer 42 protects the scintillator layer 31 from moisture in the atmosphere. An insulating layer 43 is stacked on the protective layer 42. An X-ray grid 44 is formed on the insulating layer 43. The X-ray grid 44 is shaped like a grid for shielding between the pixels 4. In the X-ray detector 1 thus configured, radiation such as X-rays 51 is incident on the scintillator layer 31 and converted to visible light 52 in the columnar crystal 32 of the scintillator layer 31. The visible light 52 travels through the columnar crystal to the photoelectric conversion element 8 of the photoelectric conversion substrate 2 and is converted to electrical signals. The electrical signal converted in the photoelectric conversion element 8 flows to the pixel electrode 7. The electrical signal is carried to the charge storage capacitor 6 connected to the pixel electrode 7. The electrical signal is held and stored in the charge storage capacitor 6 until the gate electrode 12 of the thin film transistor 5 connected to the pixel electrode 7 turns to the driving state. At this time, when one of the control electrodes 11 is turned to the driving state, one row of thin film transistors 5 connected to this control electrode 11 turned to the driving state turn to the driving state. The electrical signal stored in the charge storage capacitor 6 connected to each thin film transistor 5 turned to the driving state is outputted to the read electrode 13. This results in outputting a signal corresponding to a particular row of pixels 4 of the X-ray image. Thus, the signal corresponding to all the pixels 4 of the X-ray image can be outputted by the driving control of the control electrodes 11. This output signal is converted to a digital image signal for output. Next, a second structure example of the X-ray detector 1 is described with reference to FIG. 2. The description uses the same reference numerals as in the first structure example of the X-ray detector 1, and omits the description of similar configurations and operations. The photoelectric conversion substrate 2 has the same structure and operation as that of the first structure example. A scintillator panel 62 is bonded onto the photoelectric conversion substrate 2 via a bonding layer 61. The scintillator panel 62 includes a support substrate 63 transmissive to X-rays 51. A reflective layer 41 reflective to light is formed on the support substrate 63. A scintillator layer 31 including a plurality of strip-shaped columnar crystals 32 is formed on the reflective layer 41. A protective layer 42 for sealing the scintillator layer 31 is stacked on the scintillator layer 31. Furthermore, an X-ray grid 44 shaped like a grid for shielding between the pixels 4 is formed on the support substrate 63. In the X-ray detector 1 thus configured, X-rays 51 are incident on the scintillator layer 31 of the scintillator panel 62 and converted to visible light 52 in the columnar crystal 32 of the scintillator layer 31. The visible light 52 travels through the columnar crystal to the photoelectric conversion element 8 of the photoelectric conversion substrate 2 and is converted to electrical signals. The electrical signal is converted to a digital image signal for output as described above. Next, a third structure example of the X-ray detector 1 is described with reference to FIG. 3. The third structure example of the X-ray detector 1 is similar in configuration to the first structure example of the X-ray detector 1 shown in FIG. 1 except that the scintillator layer 31 does not include the columnar crystals 32. Next, a fourth structure example of the X-ray detector 1 is described with reference to FIG. 4. The fourth structure example of the X-ray detector 1 is similar in configuration to the second structure example of the X-ray detector 1 shown in FIG. 2 except that the scintillator layer 31 does not include the columnar crystals 32. In the X-ray detector 1 of the structures shown in FIG. 1 to FIG. 4, the scintillator layer 31 is made of a phosphor containing Tl as an activator in CsI, which is a halide. Furthermore, the scintillator layer 31 has the following features (1), (2), and (3). (1) The concentration of the activator in the phosphor is 1.6 mass %±0.4 mass %. The concentration distribution of the activator in the in-plane direction and film thickness direction of the phosphor is within ±15%. (2) In at least the region of a unit film thickness of 200 nm or less, the concentration distribution of the activator in the in-plane direction and film thickness direction of the phosphor is within ±15%. Thus, the uniformity is maintained. (3) The scintillator layer 31 is formed by vacuum evaporation technique using two evaporation sources of CsI and TlI. Furthermore, preferably, the scintillator layer 31 has a structure of strip-shaped columnar crystals 32. In the X-ray detector 1 of the first structure example shown in FIG. 1, the correlation of the Tl concentration in the scintillator layer 31 with various characteristics was tested. The result is shown in FIG. 6 to FIG. 8. In this test, the film thickness of the scintillator layer 31 is 600 μm, and the activator is Tl. Furthermore, the correlation of the stacking pitch (formation pitch of a unit film thickness (formation film thickness per rotation of the substrate)) of the scintillator layer 31 with various characteristics was tested. The result is shown in FIG. 9 to FIG. 11. FIG. 6 shows the correlation of the Tl concentration in the scintillator layer 31 with sensitivity ratio. The test condition is such that X-rays are incident at 70 kV and 0.0087 mGy. The sensitivity ratio is the ratio with reference to the sensitivity in the case where the Tl concentration in the scintillator layer 31 is 0.1 mass %. The condition for forming the scintillator layer of each test sample is the same (except the Tl concentration in the scintillator layer 31). As shown in FIG. 6, the sensitivity was maximized for the Tl concentration in the scintillator layer 31 around 1.4 mass %-1.8 mass %. FIG. 7 shows the correlation of the Tl concentration in the scintillator layer 31 with MTF ratio. The MTF ratio represents resolution. The test condition is such that X-rays are incident at 70 kV and 0.0087 mGy. The MTF ratio is the ratio with reference to MTF (at 2 Lp/mm) in the case where the Tl concentration in the scintillator layer 31 is 0.1 mass %. The condition for forming the scintillator layer of each test sample is the same (except the Tl concentration in the scintillator layer 31). As shown in FIG. 7, the result was generally constant up to the Tl concentration in the scintillator layer 31 around 2.0 mass %. FIG. 8 shows the correlation of the Tl concentration in the scintillator layer 31 with residual image ratio. The test condition is as follows. The dose of incident X-rays in the (n−1)-th X-ray image is greater than that in the n-th X-ray image. In the (n−1)-th X-ray image, X-rays are incident at 70 kV and 0.87 mGy. The subject is a lead plate (plate thickness 3 mm). The X-ray image capture interval is 60 sec. In the n-th X-ray image, X-rays are incident at 70 kV and 0.0087 mGy. The subject is none. The X-ray image capture interval is 60 sec. Furthermore, the residual image ratio is the ratio with reference to the residual image in the case where the Tl concentration in the scintillator layer 31 is 0.1 mass %. The condition for forming the scintillator layer of each test sample is the same (except the Tl concentration in the scintillator layer 31). As shown in FIG. 8, the residual image was minimized for the Tl concentration in the scintillator layer 31 around 1.6 mass %. Furthermore, no residual image was observed in the region where the residual image ratio is 0.5 (preferably 0.4) or less and the Tl concentration in the scintillator layer 31 is 1.6 mass %±0.4 mass %. FIG. 9 shows the correlation of the stacking pitch of the scintillator layer 31 with sensitivity ratio. The test condition is such that X-rays are incident at 70 kV and 0.0087 mGy. The Tl concentration in the scintillator layer 31 is 0.1 mass %. The sensitivity ratio is the ratio with reference to the sensitivity in the case where the stacking pitch of the scintillator layer 31 is 200 nm. The condition for forming the scintillator layer of each test sample is the same (except the Tl concentration in the scintillator layer 31). FIG. 10 shows the correlation of the stacking pitch of the scintillator layer 31 with MTF ratio. The test condition is such that X-rays are incident at 70 kV and 0.0087 mGy. The Tl concentration in the scintillator layer 31 is 0.1 mass %. The MTF ratio is the ratio with reference to MTF (at 2 Lp/mm) in the case where the stacking pitch of the scintillator layer 31 is 200 nm. The condition for forming the scintillator layer of each test sample is the same (except the Tl concentration in the scintillator layer 31). FIG. 11 shows the correlation of the stacking pitch of the scintillator layer 31 with residual image ratio. The test condition is such that the dose of incident X-rays in the (n−1)-th X-ray image is greater than that in the n-th X-ray image. In the (n−1)-th X-ray image, X-rays are incident at 70 kV and 0.87 mGy. The subject is a lead plate (plate thickness 3 mm). The X-ray image capture interval is 60 sec. In the n-th X-ray image, X-rays are incident at 70 kV and 0.0087 mGy. The subject is none. The X-ray image capture interval is 60 sec. Furthermore, the Tl concentration in the scintillator layer 31 is 0.1 mass %. The residual image ratio is the ratio with reference to the residual image in the case where the stacking pitch of the scintillator layer 31 is 200 nm. The condition for forming the scintillator layer of each test sample is the same (except the Tl concentration in the scintillator layer 31). As shown in FIG. 9 to FIG. 11, the characteristics tend to be degraded in the region where the stacking pitch of the scintillator layer 31 is 200 nm or more. The light emission wavelength of the scintillator layer 31 has a peak wavelength around 550 nm. The scintillator layer 31 is made primarily of CsI, which has a refractive index of 1.8. The peak wavelength of light emission propagating in the scintillator layer 31 is denoted by λ1. Then, it can be regarded that λ1=550 nm/1.8=306 nm from the relationship between refractive index and wavelength. Thus, in the case where the stacking pitch of the scintillator layer 31 is larger than λ1, the result of FIG. 9 to FIG. 11 is attributable to the increased possibility of the influence of the degradation of optical characteristics (such as scattering and attenuation) associated with e.g. variation of the crystallinity of the scintillator layer 31 and variation of the Tl concentration in the scintillator layer 31. As shown in FIG. 8, the residual image was minimized when the concentration of the activator in the phosphor constituting the scintillator layer 31 is around 1.6 mass %. No residual image was observed in the region of 1.6 mass %±0.4 mass % where the residual image ratio is 0.5 (preferably 0.4) or less. Furthermore, as shown in FIG. 6 and FIG. 7, the characteristics of sensitivity and MTF are also favorable in the region of 1.6 mass %±0.4 mass %. Thus, the concentration of the activator is preferably in the region of 1.6 mass %±0.4 mass %. As shown in FIG. 6 to FIG. 8, the characteristics are nearly stable in the region where the Tl concentration in the scintillator layer 31 is 1.6 mass %±0.4 mass %. Thus, the variation of the characteristics is small even if the Tl concentration in the scintillator layer 31 is varied (approximately ±15%). Even if the concentration of the activator in the phosphor is in the region of 1.6 mass %±0.4 mass %, the characteristics are likely to vary significantly if the concentration distribution of the activator is significantly biased in the in-plane direction and film thickness direction of the phosphor. Thus, the concentration distribution of the activator in the in-plane direction and film thickness direction of the phosphor is preferably within ±15%. The variation of characteristics is small and has little influence if the concentration distribution of the activator is in the variation range of approximately ±15%. Thus, as described above in feature (1), preferably, the concentration of the activator in the phosphor is 1.6 mass %±0.4 mass %, and the concentration distribution of the activator in the in-plane direction and film thickness direction of the phosphor is within ±15%. In at least the region of the phosphor where the unit film thickness is 200 nm or less, the characteristics are likely to vary significantly if the concentration distribution of the activator is significantly biased in the in-plane direction and film thickness direction of the phosphor. Thus, as described above in feature (2), preferably, also in the region of a unit film thickness of 200 nm or less, the concentration distribution of the activator in the in-plane direction and film thickness direction of the phosphor is within ±15%. FIG. 12 is a schematic view of a method for forming the scintillator layer 31. A substrate 72 (corresponding to the photoelectric conversion substrate 2 or the support substrate 63) is placed in the vacuum chamber 71. The film of the scintillator layer 31 is stacked by vacuum evaporation technique. In the vacuum evaporation technique, evaporation particles from the evaporation source 73 of CsI and evaporation particles from the evaporation source 74 of TlI placed in the vacuum chamber 71 are evaporated on the stacking surface of the substrate 72 while rotating the substrate 72. At this time, the Tl concentration distribution in the in-plane direction and film thickness direction per stacking cycle of the scintillator layer 31 can be arbitrarily controlled by controlling the rotation cycle of the substrate 72 and the evaporation of CsI and TlI. Thus, the uniformity of the Tl concentration distribution in the in-plane direction and film thickness direction of the overall scintillator layer 31 is ensured by ensuring the uniformity of the Tl concentration distribution in the in-plane direction and film thickness direction per stacking cycle of the scintillator layer 31 when the scintillator layer 31 is formed. Accordingly, the characteristics, in particular the residual image characteristic, of the scintillator layer 31 can be improved by providing the above features (1)-(3) to the scintillator layer 31 made of a phosphor containing Tl as an activator in CsI, which is a halide. A practical example of the X-ray detector 1 of the first structure example shown in FIG. 1 is now described. In this practical example, the film thickness of the scintillator layer 31 is 600 μm. The stacking pitch of the scintillator layer 31 is 150 nm. The concentration distribution of the activator in the in-plane direction and film thickness direction of the scintillator layer 31 is ±15%. The activator is Tl. Five samples are produced with the concentration of the activator in the scintillator layer 31 being 0.1 mass %, 1.0 mass %, 1.2 mass %, 1.6 mass %, and 2.0 mass %. For these five samples, the subject is radiographed under a particular radiography condition. The radiographed image is processed in a prescribed image processing condition. FIGS. 13A, 13B, 13C, 13D, and 13E show (n-th) X-ray images in this case. The table of FIG. 14 shows the result of the characteristics. In FIG. 14, the sensitivity ratio, the MTF ratio, and the residual image ratio are the values with reference to the case where the Tl concentration in the scintillator layer 31 is 0.1 mass %. The radiography condition is as follows. The dose of incident X-rays in the (n−1)-th X-ray image is greater than that in the n-th X-ray image. In the (n−1)-th X-ray image, X-rays are incident at 70 kV and 0.87 mGy. The subject is a lead plate (plate thickness 3 mm). The X-ray image capture interval is 60 sec. In the n-th X-ray image, X-rays are incident at 70 kV and 0.0087 mGy. The subject is none. The X-ray image capture interval is 60 sec. With regard to the image processing condition, the flat field correction is applied. The window processing is applied (the histogram average of the image ±10%). As shown in FIGS. 13A and 13B, when the concentration of the activator is 0.1 mass % and 1.0 mass %, a residual image is observed in the range enclosed with the dashed line in the figure. As shown in FIGS. 13C, 13D, and 13E, when the concentration of the activator is 1.2 mass %, 1.6 mass %, and 2.0 mass %, no residual image is observed in the range enclosed with the dashed line in the figure. Thus, if the above features (1)-(3) defined in this embodiment are provided to the scintillator layer 31, the residual image characteristic can be improved with the sensitivity and MTF being also favorable. This can improve the performance and reliability of the X-ray detector 1. Next, an embodiment in which the scintillator layer according to the invention is used in a scintillator panel is described. In FIG. 15 to FIG. 19, the basic configuration of the scintillator panel 90 is described with reference to first to fourth structure examples. First, a first structure example of the scintillator panel 90 is described with reference to FIG. 15. The scintillator panel 90 includes a support substrate 91 transmissive to radiation such as X-rays. A reflective layer 92 reflective to light is formed on the support substrate 91. A scintillator layer 93 for converting radiation to visible light is formed on the reflective layer 92. A protective layer 94 for sealing the scintillator layer 93 is stacked on the scintillator layer 93. The support substrate 91 is formed from a material composed primarily of light elements rather than transition metal elements and having good X-ray transmittance. The reflective layer 92 is made of a metal material having high reflectance such as Al, Ni, Cu, Pd, and Ag. The reflective layer 92 reflects light generated in the scintillator layer 93 to the direction opposite to the support substrate 91. Thus, the reflective layer 92 enhances the light utilization efficiency. The scintillator layer 93 is formed by depositing a high-brightness fluorescent material in a columnar shape on the support substrate 91 by vapor phase growth technique such as vacuum evaporation technique, sputtering technique, and CVD technique. The high-brightness fluorescent material is a phosphor such as a halide including cesium iodide (CsI) and an oxide-based compound including gadolinium oxysulfide (GOS). The scintillator layer 93 is formed in a columnar crystal structure such that a plurality of strip-shaped columnar crystals 93a are formed in the in-plane direction of the support substrate 91. In the scintillator panel 90 thus configured, radiation such as X-rays 96 is incident on the scintillator layer 93 from the support substrate 91 side and converted to visible light 97 in the columnar crystal 93a of the scintillator layer 93. The visible light 97 is emitted from the surface of the scintillator layer 93 (the surface of the protective layer 94) on the opposite side from the support substrate 91. FIG. 16 shows a second structure example of the scintillator panel 90. The second structure example of the scintillator panel 90 is similar in configuration to the first structure example of the scintillator panel 90 shown in FIG. 15 except for not including the reflective layer 92. FIG. 17 shows a third structure example of the scintillator panel 90. The third structure example of the scintillator panel 90 is similar in configuration to the first structure example of the scintillator panel 90 shown in FIG. 15 except that the scintillator layer 93 does not include the columnar crystals 93a. FIG. 18 shows a fourth structure example of the scintillator panel 90. The fourth structure example of the scintillator panel 90 is similar in configuration to the second structure example of the scintillator panel 90 shown in FIG. 16 except that the scintillator layer 93 does not include the columnar crystals 93a. FIG. 19 shows a radiography device 100 of e.g. the CCD-DR type based on the scintillator panel 90. The radiography device 100 includes a housing 101. The scintillator panel 90 is placed at one end of the housing 101. A specular reflective plate 102 and an optical lens 103 are placed inside the housing 101. A light receiving element 104 such as CCD is placed at the other end of the housing 101. X-rays 96 are radiated from the X-ray source (X-ray tube) 105 and incident on the scintillator panel 90. The visible light 97 converted in the scintillator layer 93 is emitted from the surface of the scintillator layer 93. The X-ray image is projected on the surface of the scintillator layer 93. This X-ray image is reflected by the reflective plate 102. On the other hand, the X-ray image is collected by the optical lens 103 and applied to the light receiving element 104. The X-ray image is converted to electrical signals in the light receiving element 104 for output. In the scintillator panel 90 of the structures shown in FIG. 15 to FIG. 19, the scintillator layer 93 is made of a phosphor containing Tl as an activator in CsI, which is a halide. Furthermore, the scintillator layer 93 has the following features (1), (2), and (3). (1) The concentration of the activator in the phosphor is 1.6 mass %±0.4 mass %. The concentration distribution of the activator in the in-plane direction and film thickness direction of the phosphor is within ±15%. (2) In at least the region of a unit film thickness of 200 nm or less, the concentration distribution of the activator in the in-plane direction and film thickness direction of the phosphor is within ±15%. Thus, the uniformity is maintained. (3) The scintillator layer 93 is formed by vacuum evaporation technique using two evaporation sources of CsI and TlI. Furthermore, preferably, the scintillator layer 93 has a structure of strip-shaped columnar crystals 93a. As described with reference to FIG. 6 to FIG. 11, the scintillator layer 93 having the above features (1)-(3) defined in this embodiment is used in the scintillator panel 90. Thus, the residual image characteristic can be improved with favorable sensitivity and MTF provided to the scintillator panel 90. This can improve the performance and reliability of the scintillator panel 90. The method for forming the scintillator layer 93 can be made similar to the method for forming the scintillator layer 31 described with reference to FIG. 12. While certain embodiments have been described, these embodiments have been presented by way of example only, and are not intended to limit the scope of the inventions. Indeed, the novel embodiments described herein may be embodied in a variety of other forms; furthermore, various omissions, substitutions and changes in the form of the embodiments described herein may be made without departing from the spirit of the inventions. The accompanying claims and their equivalents are intended to cover such forms or modifications as would fall within the scope and spirit of the invention. |
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abstract | A pressure-relief system for a containment of a nuclear plant has a pressure-relief line which is led through the containment and is closed by a shutoff device, and a wet scrubber being switched into the pressure-relief line lying outside the containment, for the pressure-relief gas flow developing in the pressure-relief operating mode with the shutoff device being open. An effective, reliable operation of the wet scrubber with a compact structural configuration is made possible. This is achieved by a reservoir, arranged in the containment or fluidically connected therewith such that an overpressure, as compared with the outer environment, present in the containment, is transferred to the reservoir, and a feeding line which is led from the reservoir to the wet scrubber and can be closed by a shutoff device, for feeding a liquid active as a scrubbing liquid from the reservoir to the wet scrubber. |
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abstract | An improved method for substrate micromachining. Preferred embodiments of the present invention provide improved methods for the utilization of charged particle beam masking and laser ablation. A combination of the advantages of charged particle beam mask fabrication and ultra short pulse laser ablation are used to significantly reduce substrate processing time and improve lateral resolution and aspect ratio of features machined by laser ablation to preferably smaller than the diffraction limit of the machining laser. |
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abstract | In an exposing method reflecting synchrotron radiation, having a critical wavelength of 8.46 xc3x85, emitted from a radiation generator (SR device) having a deflecting magnetic field of 4.5 T and electron acceleration energy of 0.7 GeV twice through rhodium mirrors having an oblique-incidence angle of 1xc2x0, transmitting the light through a beryllium window of 20 xcexcm and through an X-ray mask prepared by forming an X-ray absorber pattern on a diamond mask substrate of 2 xcexcm in thickness and thereafter irradiating a resist surface provided on a substrate with the light, the resist has a main absorption waveband in the wave range of at least 3 xc3x85 and not more than 13 xc3x85 and contains an element generating Auger electrons having energy in the range of at least about 0.51 KeV and not more than 2.6 KeV upon exposure. |
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061047732 | summary | BACKGROUND OF THE INVENTION Field of the Invention The invention relates to a fuel rod for a nuclear reactor, having a metal cladding tube being filled with nuclear fuel, a metal seal plug, locking plug or stopper being welded to one end of the tube and in particular being formed of the same metal, and an annular bead on the outer surface of the cladding tube at a transition point between the cladding tube and the seal plug. The invention also relates to a welding apparatus for producing the fuel rod. Such a fuel rod is already typical and its annular bead has a sharp annular gradient. In other words, the cross section of the annular bead forms an acute triangle with an acute angle located at the annular gradient. Such a fuel rod can be produced in a welding apparatus with an electrode in which there is a bore for receiving one end of a cladding tube. The bore is chamfered, forming a frustoconical void that tapers toward the inside, on an end of the bore that faces toward a counter electrode which is displaceable relative to the aforementioned electrode and is intended to hold the seal plug. The frustoconical void determines the shape of the annular bead that forms from solidified welding melt when the seal plug is permanently welded to the cladding tube. The welded connection between the cladding tube and the seal plug is considered perfect if the frustoconical void of the electrode for the cladding tube is filled to a predetermined extent during welding with material of the cladding tube and the seal plug, which is ascertainable by experimentation. After the welding, the quality of the weld can be determined by measuring the height of the gradient of the annular bead at the weld between the seal plug and the cladding tube of the fuel rod. In order to produce nuclear reactor fuel assemblies that can be inserted into a nuclear reactor, the fuel rods must be threaded into a mesh of spacers, which as a rule are lattice-like. The annular bead at the weld between the cladding tube and the seal plug of the individual fuel rods is a hindrance to the threading operation. Therefore, after the quality of the weld has been measured and ascertained, the annular beads are mechanically machined down, for instance by being milled off, until it can be assumed that they will no longer hinder the threading of the fuel rods into the spacers. In the mechanical machining down process, damage to the cladding tube can occur, causing a fuel rod to be rejected. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a fuel rod for a nuclear reactor and a welding apparatus for producing the fuel rod, which overcome the hereinafore-mentioned disadvantages of the heretofore-known devices of this general type and which eliminate the operation of mechanical machining down. With the foregoing and other objects in view there is provided, in accordance with the invention, a fuel rod for a nuclear reactor, comprising a metal cladding tube being filled with nuclear fuel and having ends, an outer surface, and a longitudinal axis; a metal seal plug being welded to one of the ends of the cladding tube at a transition point and in particular being formed of the same metal, defining an annular bead on the outer surface of the cladding tube at the transition point; the annular bead having a cylindrical outer jacket surface with jacket lines being substantially parallel to the longitudinal axis of the cladding tube; and the annular bead having material being formed of the metal of the cladding tube and the metal of the seal plug. The outside diameter of the cylindrical jacket surface of the annular bead can be selected from the outset in such a way that threading the applicable fuel rod into cells of spacers without mechanical post-machining of the annular bead is possible. On the other hand, a conclusion as to the quality of the welded connection between the cladding tube and the seal plug of the applicable fuel rod can be drawn from the relative forward feed between the electrode having the cladding tube and the counter electrode having the seal plug as these parts are welded together. In accordance with another feature of the invention, the annular bead has at least two humps on the cylindrical outer jacket surface, each extending along one of the jacket lines of the jacket surface. In accordance with a further feature of the invention, the annular bead is resolidified from a welding melt. In accordance with an added feature of the invention, the cylindrical outer jacket surface of the annular bead is mechanically unmachined or unworked. In accordance with an additional feature of the invention, the cylindrical outer jacket surface of the annular bead has an encompassing depression formed therein in the annular bead containing the material of the cladding tube having the same microscopic structure as in the cladding tube. With the objects of the invention in view, there is also provided a welding apparatus for producing a fuel rod, comprising an electrode having a through bore formed therein for receiving one end of a cladding tube; and a counter electrode being displaceable relative to the electrode for holding a seal plug to be welded the one end of the cladding tube; the electrode having a cylindrical step or shoulder formed therein at an end of the through bore facing toward the counter electrode, the cylindrical step having a diameter being greater than the diameter of the through bore. With such a welding apparatus, a fuel rod according to the invention can be manufactured simply and economically. 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 fuel rod for a nuclear reactor and a welding apparatus for producing the fuel rod, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. |
055740774 | claims | 1. A microwave-energy-absorbing material, comprising orientationally mobile, polar icosahedral molecular units as the source of dielectric loss at microwave frequencies, and a host matrix, said polar icosahedral molecular units blended with said matrix, wherein said molecular units comprise polar carboranes with electronegative substituents on the carborane cage borons opposite to the carbons of said molecular units, wherein said substituent is bromine. 2. A microwave-energy-absorbing material, comprising orientationally mobile, polar icosahedral molecular units as the source of dielectic loss at microwave frequencies, and a host matrix, said polar icosahedral molecular units blended with said matrix, wherein said molecular units comprise polar carboranes with electronegative substituents on the carborane cage borons opposite to the carbons of said molecular units, wherein said substituent is a phosphate ester, ion or sulfate. 3. A microwave-energy-absorbing structure comprising an ion-stabilized gel of crosslinked polymers, into which is loaded polar icosahedral molecules as the source of dielectric loss at microwave frequencies. 4. The structure of claim 3 wherein said gel is synthesized from propylene carbonate in a gel-forming reaction, wherein a diacrylate monomer, propylene carbonate and a photoinitiator are stabilized by solvation with a salt, and the resulting mixture is polymerized via irradiation of ultraviolet light to provide the crosslinked gel. 5. A microwave-energy-absorbing material comprising a foam, wherein said foam comprises macroreticular resins, loaded with polar icosahedral molecular units to increase the dielectric response, wherein said resins are synthesized by the polymerization of a mixture of styrene and divinyl benzene in a poor solvent. 6. A microwave-energy-absorbing material, comprising orientationally mobile, polar icosahedral molecular units as the source of dielectric loss at microwave frequencies, and a host matrix comprising poly cisisoprene, said polar icosahedral molecular units blended with said matrix, wherein said molecular units comprise polar carborane molecules. 7. The material of claim 6 wherein said molecular units comprise ortho-carborane molecular units. 8. The material of claim 6 wherein said molecular units comprise meta-carborane molecular units. 9. A microwave-energy-absorbing material, comprising orientationally mobile, polar icosahedral molecular units as the source of dielectric loss at microwave frequencies, and a host matrix comprising silicone rubber, said polar icosahedral molecular units blended with said matrix, wherein said molecular units comprise polar carborane molecules. 10. The material of claim 9 wherein said molecular units comprise ortho-carborane molecular units. 11. The material of claim 9 wherein said molecular units comprise meta-carborane molecular units. 12. A microwave-energy-absorbing material, comprising orientationally mobile, polar icosahedral molecular units as the source of dielectric loss at microwave frequencies, and a host matrix comprising a high modulus organic polymer blended with said polar icosahedral molecules, said polar icosahedral molecular units blended with said matrix, wherein said molecular units comprise polar carborane molecules. 13. The material of claim 12 wherein said molecular units comprise ortho-carborane molecular units. 14. The material of claim 12 wherein said molecular units comprise meta-carborane molecular units. 15. The material of claim 12 wherein said high modulus organic polymer comprises poly-styrene. 16. The material of claim 12 wherein said high modulus organic polymer comprises polycarbonate. 17. The material of claim 12 wherein said high modulus organic polymer comprises a poly-acrylate. |
summary | ||
050142917 | claims | 1. An X-ray amplifying device comprising means defining an annular space; an exciting gas in said annular space; internal and external circular concentric rings of suitable metallic material circumscribing said annular space and spaced apart to form a channel between them, at least one of said rings forming an X-ray reflection electrode, an entrance and an exit port in said channel, means for applying a suitable difference of electrical potential to the said metallic rings so as to bring them into an excited state favourable to X-ray emission, a suitable X-ray source providing a primary X-ray beam directed so as to enter said entrance port striking against said reflection electrode so as to be reflected at least once before leaving said channel through said exit port so as to cause the X-rays, emitted from the reflection electrode by induced emission, to be superimposed on the X-ray reflected by said reflection electrode thus inceasing the intensity of the X-ray beam leaving the device. 2. A device according to claim 1, in which the applied difference of potential is sufficient to accelerate particles of the exciting gas such that they, in turn, excite the metal rings favourably for X-ray emission. 3. A device according to claim 1, said entrance port and exit port each placed substantially tangent to the said channel between the internal and external metallic rings and at such an angle that after one or more reflections of the incidental rays through entrance port, ray emission takes place through the exit port. 4. A device according to claim 3, in which the concentric metallic rings and the entrance and exit channels are housed in a container that can be emptied of air. 5. A device according to claim 1, in which the external circular metallic ring constitutes part of both a reflective surface and an anticathode for induced emission of X-rays. 6. A device according to claim 1 in which the exciting gas is made up of Xenon or other gas of higher atomic number than that of the metallic electrodes and the difference of potential applied to the metallic rings is of a potential superior to the highest typical potential of ionization of the metal of the metallic-rings. 7. A device according to claim 1, in which, when using a conventional source in tungsten for the primary X-rays, the concentric metallic rings are made of silver or of a silver plated metal, or of tin or other metal of an elevated atomic number, but in any case lower or equal to that of tungsten. 8. A device according to claim 1, in which the two circular concentric electrodes are of metals, of any element suitable for X-ray production, which have either the same atomic number or a lower atomic number than that of the metal of the anticathode producer of the primary X-rays. 9. An X-ray amplifying device comprising means defining an annular space; an exciting gas in said annular space; internal and external circular concentric rings of suitable metallic material circumscribing said annular space and spaced apart to form a channel between them, means for applying a suitable difference of electrical potential to the said metallic rings so as to bring them into an excited state favourable to X-ray emission, wherein the applied difference of potential is sufficient to accelerate particles of the exciting gas such that they, in turn, excite the metal rings favourably for X-ray emission. 10. An X-ray amplifying device comprising means defining an annular space; an exciting gas in said annular space; internal and external circular concentric rings of suitable metallic material circumscribing said annular space and spaced apart to form a channel between them, means for applying a suitable difference of electrical potential to the said metallic rings so as to bring them into an excited state favourable to X-ray. emission, and further comprising an entrance channel and an exit channel, each placed substantially tangent to the said channel between the internal and external metallic rings and at such an angle that after one or more reflections of incidental rays through the entrance channel, ray emission takes place through the exit channel. 11. A device according to claim 10, in which the concentric metallic rings and the entrance and exit channels are housed in a container that can be emptied of air. |
040574687 | summary | BACKGROUND OF THE INVENTION This invention relates to liquid metal cooled fast breeder nuclear reactors and to fuel element sub-assemblies therefor. In fast breeder nuclear reactors it is common to divide the fuel assembly into replaceable fuel element sub-assemblies each comprising a bundle of fuel pins contained within a wrapper or shroud through which liquid metal is flowed in heat exchange with the fuel pins. In one known construction of nuclear reactor the fuel element sub-assemblies are upstanding from a carrier associated with a core supporting diagrid and are arranged in groups and urged into leaning abutment with a central support member for each group of sub-assemblies. The upper regions of the sub-assemblies contain massive steel shielding to form, in combination, an upper shield for the core. Thus with the weight mass being disposed at the upper free end of each sub-assembly there is a tendency for vibration to be set up by coolant flow through the sub-assembly. It is expected that in one envisaged construction of reactor core the amplitude of vibration of the tip of a sub-assembly could be 1.0 mm and, apart from the deleterious effect on the mechanical reliability of the reactor core, such amplitude of vibration could create excessive reactivity noise, that is, it could create slight, variable reactivity changes. SUMMARY OF THE INVENTION According to the invention in a liquid metal cooled fast breeder nuclear reactor having a core comprising a plurality of elongate fuel element sub-assemblies upstanding from core support means, each sub-assembly has inertia damping means at a free upper end region. In a preferred construction of fuel element sub-assembly for a nuclear reactor according to the invention the inertia damping means comprises upper and lower sleeves mounted on a tubular spine, the lower sleeve being rigidly attached to the spine and secured atop a fuel containing tubular wrapper of the sub-assembly whilst the upper sleeve is resiliently attached to the lower sleeve and radially spaced from the spine there being duct means for enabling liquid metal to flood the radial spacing of the upper sleeve and spine when the sub-assembly is submerged in reactor coolant. In use, the device operates on a tuned inertia damping principle using liquid metal as the damping medium; when the upper sleeve vibrates the liquid metal is forced from one side of the spacing to the other and its resistance to this motion provides a damping force. For the optimum damping effect the mass of the upper sleeve needs to be as large as possible in relation to the mass of the lower region of the sub-assembly and from this aspect it is convenient to comprise the upper sleeve of massive steel neutron shielding. Preferably, the resilient attachment of the upper and lower sleeves comprises a bellows unit which is easily constructed and unlikely to break into small pieces should it fail through fatigue. |
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