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abstract | The object of this invention is to provide a method for mitigating a stress corrosion cracking of reactor structural material which makes it possible to suppress the rise in the main steam line dose rate without secondary effects such as a rise in the concentration of radioactive cobalt-60, etc. in the reactor water. Hydrogen and a reductive nitrogen compound containing nitrogen having a negative oxidation number (for example, hydrazine) are injected into the core water of boiling water nuclear power plant. By injecting the reductive nitrogen compound containing nitrogen having a negative oxidation number into the core water, the stress corrosion cracking of structural material of reactor can be mitigated without side reactions such as a rise in the concentration of cobalt-60, etc. |
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claims | 1. A pressurized water nuclear reactor pressure vessel comprising:an upper support plate;a removable upper head;a plurality of rod cluster control assembly upper guide tubes, each guide tube extending from the upper support plate toward the removable upper head; anda plurality of in-core instrumentation assembly column extensions, each column extension continuously extending from and being supported by a horizontal section of the support plate and extending toward the removable upper head to an elevation substantially at or above the rod cluster control assembly upper guide tubing, and each column extension connected by a bracket to one of the upper guide tubes and each column extension provides support for in-core instrumentation over substantially the entire length of the column extension. 2. The pressurized water reactor pressure vessel of claim 1 wherein coolant is conveyed to the reactor pressure vessel by a reactor coolant pump having a rotation frequency and wherein an assembly of the upper guide tube attached by the bracket to the upper mounted instrumentation columns have natural resonant vibration frequencies substantially different from the excitation frequency of the reactor coolant pump. 3. The pressurized water reactor pressure vessel of claim 1 wherein each column extension is connected by bracket to a different one of the upper guide tubes. 4. The pressurized water reactor pressure vessel of claim 1 wherein the column extensions are respectively connected to or adjacent the top of the upper guide tubes. 5. The pressurized water reactor pressure vessel of claim 1 wherein the bracket is welded to the upper guide tube at one end and the column extension is screwed into the bracket at another end. 6. The pressurized water reactor pressure vessel of claim 1 wherein each of the column extensions is fastened at one end to the upper support plate. 7. The pressurized water reactor pressure vessel of claim 1 wherein one end of each of the column extensions is aligned with its own penetration in the upper head. 8. The pressurized water reactor pressure vessel of claim 1 wherein at least some of the in-core instrument assembly column extensions provide moveable in-core detectors access to a nuclear fuel assembly within a core within the pressure vessel. 9. The pressurized water reactor pressure vessel of claim 1 wherein a lower end of the column extensions are respectively fastened to a support column that extends between the upper support plate and an upper core plate. 10. The pressurized water reactor pressure vessel of claim 1 wherein the lower end of the column extensions are respectively screwed into the support columns. 11. The pressurized water reactor pressure vessel of claim 1 wherein at least some of the column extensions extend above the upper guide tubes. 12. The pressurized water reactor pressure vessel of claim 1 wherein the instrumentation assembly column extensions end in funnels between the upper support plate and the removable upper head. 13. The pressurized water reactor pressure vessel of claim 1 further comprising:a core including a plurality of fuel assemblies;an upper core plate situated over the fuel assemblies; anda plurality of support columns extending vertically between the upper core plate at a lower end of the support columns and the upper support plate at an upper end of the support columns, the support columns respectively in line and communicating with a corresponding one of the in-core instrumentation column extensions. 14. The pressurized water reactor pressure vessel of claim 1 wherein the instrumentation assembly column extensions are substantially straight and substantially rigid and guide an instrumentation assembly substantially through the instrumentation assembly column extension's entire length. 15. A pressurized water reactor power generating facility having a pressure vessel comprising:an upper support plate;a removable upper head;a plurality of rod cluster control assembly upper guide tubes, each upper guide tube extending from the support plate toward the removable upper head; anda plurality of in-core instrumentation assembly column extensions, each column extension continuously extending from the upper support plate toward the removable upper head to an elevation substantially at or above the rod cluster control assembly upper guide tubing, and each column extension connected by a bracket to one of the upper guide tubes and each column extension provides support for in-core instrumentation over substantially the entire length of the column extension. |
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039716970 | abstract | Bombarding a cesium heat pipe with high energy particles causes a spallation reaction which produces vapors of .sup.123 Xe and contaminants. The contaminants are removed in a dry ice cold trap while the .sup.123 Xe condenses in a liquid nitrogen trap where it decays to .sup.123.sub.I. |
claims | 1. An apparatus that acts as a shield for radiopharmaceuticals and protects from radioactivity comprising: a) a first body with a first hollow core that is open on a first edge and a second edge of said first body, said first hollow core for housing a hypodermic syringe; b) a second body with a second hollow core that is open on a first edge of said second body, said second hollow core for housing said hypodermic syringe; c) a third body with a third hollow core that is open on a first edge of said third body, said third hollow core for housing said hypodermic syringe; d) said hypodermic syringe capable of containing a radiopharmaceutical; e) a first connection means wherein said first body releasably communicates with said second body for providing protection from radioactivity emitted by the radiopharmaceutical; f) a second connection means wherein said first body releasably communicates with said third body for providing protection from said radioactivity; and g) said second body further comprising a means for compressing said hypodermic syringe to eject said radiopharmaceutical from the hypodermic syringe while said first body is in communication with said second body. 2. The apparatus as claimed in claim 1 wherein said first body, second body and third body are constructed from a plurality of radiation shielding materials. claim 1 3. The apparatus as claimed in claim 1 wherein said means for compressing said hypodermic syringe comprises: claim 1 a) an internal sleeve, slidably and rotatably communicating with the interior surface of said second body and capable of partially extending outside of said second body through slots formed in said second body""s second edge; b) at least one disk guide inside said second body and parallel to the long axis of said second body and extending the length of said second body; c) an actuator cap fixedly attached to said internal sleeve and positioned to block said radiation emitted from the second edge of said second body, said actuator cap engaged by said at least one disk guide; d) a disk having an at least one guide notch in slidable communication with said at least one disk guide such that the disk is prevented from rotating relative to said second body; e) said disk further having at least one engagement notch; f) at least one disk engagement tooth on the inside surface of said internal sleeve, said tooth positioned such that when the internal sleeve is selectively rotated to a position, said internal sleeve can be extended from and retracted into said second body without bringing said at least one engagement tooth into communication with the disk and when the internal sleeve is selectively rotated to any other position said at least one engagement tooth engages the disk when the actuator cap is compressed; g) wherein said at least one engagement notch is sized slightly larger than said at least one disk engagement tooth. 4. The apparatus as claimed in claim 1 wherein said means for compressing compresses said hypodermic syringe by being partially extended from said second body, selectively rotated relative to said second body and then compressed into said second body. claim 1 5. The apparatus as claimed in claim 1 wherein said first connection means and said second connection means are selected from the group consisting of threaded connection, locking nut and compression flange. claim 1 6. The apparatus as claimed in claim 1 wherein said third body is removed from said first body, permitting operation of said means for compressing to compress said hypodermic syringe to eject said radiopharmaceutical from the hypodermic syringe, while providing protection from said radiation. claim 1 7. The apparatus as claimed in claim 1 wherein said hypodermic syringe is housed in an insert. claim 1 8. The apparatus as claimed in claim 7 wherein said insert further comprises a first section and a second section wherein said second section is detachable from said first section. claim 7 9. An apparatus that acts as a shield for radiopharmaceuticals and protects individuals from radioactivity comprising: a) a first body with a first hollow core that is open on a first edge and a second edge of said first body, said first hollow core for housing a hypodermic syringe; b) a second body with a second hollow core that is open on a first edge of said second body, said second hollow core for housing said hypodermic syringe; c) a third body with a third hollow core that is open on a first edge of said third body, said third hollow core for housing said hypodermic syringe; d) said hypodermic syringe capable of containing a radiopharmaceutical; e) a first connection means wherein said first body releasably communicates with said second body for providing protection from radioactivity emitted by the radiopharmaceutical; f) a second connection means wherein said first body releasably communicates with said third body for providing protection from said radioactivity; g) said third body further comprising means for extending said hypodermic syringe from said first and third bodies to permit measurement of said radiopharmaceutical in said hypodermic syringe and providing protection from said radioactivity; and h) said second body further comprising means for compressing said hypodermic syringe to eject said radiopharmaceutical from the hypodermic syringe while said first body is in communication with said second body. 10. The apparatus as claimed in claim 9 wherein said first body, second body and third body are constructed from a plurality of radiation shielding materials. claim 9 11. The apparatus as claimed in claim 9 wherein said means for extending comprises two rod connectors and means to securely fasten said rod connectors to said hypodermic syringe. claim 9 12. The apparatus as claimed in claim 9 wherein said means for extending comprises a chain. claim 9 13. The apparatus as claimed in claim 9 wherein said means for extending comprises a cable. claim 9 14. The apparatus as claimed in claim 9 wherein said hypodermic syringe is housed in insert. claim 9 15. The apparatus as claimed in claim 14 wherein said insert further comprises a first section and a second section, wherein said second section is detachable from said first section. claim 14 16. The apparatus as claimed in claim 9 wherein said second body is removable from said first body allowing said radiopharmaceutical in said hypodermic syringe to be measured in a well counter. claim 9 17. The apparatus as claimed in claim 9 wherein said third body is removable from said first body for said individual to manipulate said hypodermic syringe to inject a patient with said radiopharmaceuticals and be protected from said radiation. claim 9 18. The apparatus as claimed in claim 9 wherein said means for compressing said hypodermic syringe comprises: claim 9 a) an internal sleeve, slidably and rotatably communicating with the interior surface of said second body and capable of partially extending outside of said second body through slots formed in said second body""s second edge; b) at least one disk guide inside said second body and parallel to the long axis of said second body and extending the length of said second body; c) an actuator cap fixedly attached to said internal sleeve and positioned to block said radiation emitted from the second edge of said second body, said actuator cap engaged by said at least one disk guide; d) a disk having an at least one guide notch in slidable communication with said at least one disk guide such that the disk is prevented from rotating relative to said second body; e) said disk further having at least one engagement notch; f) at least one disk engagement tooth on the inside surface of said internal sleeve, said tooth positioned such that when the internal sleeve is selectively rotated to a position, said internal sleeve can be extended from and retracted into said second body without bringing said at least one engagement tooth into communication with the disk and when the internal sleeve is selectively rotated to any other position said at least one engagement tooth engages the disk when the actuator cap is compressed; g) wherein said at least one engagement notch is sized slightly larger than said at least one engagement tooth. |
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abstract | The invention includes a heat exchanger provided at a position higher than a primary containment vessel; a condensate storage tank disposed below the heat exchanger and above an upper end of a reactor core placed in a reactor pressure vessel; a non-condensate gas discharge line connected to an upper section of the condensate storage tank and to a suppression pool; a second condensate discharge line connected to a position below that section of the condensate storage tank to which a first end of the non-condensate gas discharge line is connected, and to the suppression pool; and a condensate return line connected to a position below that section of the condensate storage tank to which a first end of the second condensate discharge line is connected, and to a side portion of the reactor pressure vessel, the side portion being above the upper end of the core. |
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claims | 1. A pressurized water reactor containment internal passive heat removal system with at least one cooling water circulation circuit, comprising:a heat exchanger located inside the containment and comprising an upper header and a lower header interconnected by heat-exchange tubes,a riser pipeline and a downtake pipeline connected to the heat exchanger,a cooling water supply tank located above the heat exchanger outside the containment and connected to the downtake pipeline, anda steam relief valve connected to the riser pipeline, located in the water supply tank and connected to the same hydraulically, wherein the heat exchanger is divided into heat exchanger sections each having an upper header section and a lower header section, each of the upper header sections and the lower header sections satisfying the relationship:L/D≤20,where L is the length of the respective upper or lower header section,D is the header bore of the respective upper or lower header section;wherein the riser pipeline is configured so that the riser section height hrs is no less than:hrs=(ΔPcres−Δρheghhe)/Δρrsg, ΔPcres=Δρrsghrs+Δρheghhe,where Pcres is the circuit total hydraulic resistance,hhe is the heat exchanger height,g is the gravity factor,Δρrs=ρcw−(ρ′(1−x)+ρ″x)Δρhe=ρcw−ρhw ρcw is the downtake pipeline water density,ρhw is the riser pipeline water density within the heat exchanger height range,ρ′,ρ″ are the water and steam saturation density, andx is the mean mass steam quality of the two-phase mixture in the riser section. 2. A system according to claim 1, wherein at least a part of the riser pipeline from the upper headers of the heat exchanger sections to the steam relief valve is inclined upwards from horizontal at an angle of at least 10°. 3. A system according to claim 2, wherein the riser pipeline includes sections with an inclination angle of less than 10° in relation to horizontal, the length of such sections is Lsec1 and the bore is Dsec1 meeting the following criterion:Lsec1/Dsec1≤10. 4. A system according to claim 1, wherein at least a part of the downtake pipeline has a downward inclination to an angle of at least 10° in relation to horizontal. 5. A system according to claim 4, wherein the downtake pipeline includes sections with an inclination angle of less than 10° in relation to horizontal, the length of such sections is Lsec2 and the bore is Dsec2 meeting the following criterion:Lsec2/Dsec2≤10. 6. A system according to claim 1, wherein the heat-exchange tubes have a height allowing to meet the criteria of turbulent convection on the heat exchanger outer surface, namely:Ra>4·1012,where R a = gl 3 S c ν 2 · ρ w - ρ c ρ c ,Ra is the Rayleigh criterion,g is the gravity factor,l is the heat exchanger tube height,ν is the steam-air kinematic viscosity coefficient,ρw is the steam-air medium density on the outer wall of the heat exchanger tubing,ρc is the steam-water medium density in the containment, S c = ν D dif is the Schmidt number, andDdif is the steam diffusion factor. 7. A system according to claim 1, wherein the heat exchanger section has a single-row vertical bundle. |
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042343852 | claims | 1. A nuclear fuel element comprising a fuel body formed of nuclear fuel material and a clad covering the fuel body, the clad being formed of an austenitic stainless steel consisting essentially of nickel, chromium, silicon in an amount of more than 0 to less than 0.7% by weight, carbon and titanium in amounts satisfying the relationship: EQU 17[C]+27[Ti].gtoreq.2.1 2. A nuclear fuel element as defined in claim 1, wherein the upper limit of the titanium content is 0.3% by weight. 3. A nuclear fuel element as defined in claim 1, wherein said austenitic stainless steel contains silicon in an amount of 0.5% by weight or more. 4. A nuclear fuel element as defined in claim 1, wherein said austenitic stainless steel contains 6 to 26% by weight of nickel and 9 to 26% by weight of chromium. 5. A nuclear fuel element as defined in claim 1, wherein the final heat-treatment is carried out at a temperature of more than 1120.degree. C. up to 1200.degree. C. 6. A nuclear fuel element as defined in claim 1, wherein the nuclear fuel material is selected from the group consisting of the oxides, the nitrides and the carbides of uranium, plutonium and thorium, and a mixture thereof. 7. A nuclear fuel element comprising: a fuel body formed of a nuclear fuel material selected from the group consisting of the oxides, nitrides and carbides of uranium, plutonium and thorium, and a mixture thereof; a clad covering the fuel body, the clad being formed of an austenitic stainless steel consisting essentially of 6 to 26% by weight of nickel, 9 to 26% by weight of chromium, 0.5 to less than 0.7% by weight of silicon, carbon and titanium in amounts satisfying the relationship: EQU 17[C]+27[Ti].gtoreq.2.1 8. A nuclear fuel subassembly comprising a plurality of nuclear fuel elements according to claim 7. |
053696760 | summary | BACKGROUND OF THE INVENTION The present invention relates generally to nuclear reactors, and, more specifically, to a mechanism for refueling a reactor core in a pressure vessel of a boiling water reactor. A boiling water reactor (BWR) includes a reactor core submerged in water in a closed reactor pressure vessel, with the vessel being disposed in a containment vessel above which is disposed a refueling pool of water. During a refueling cycle, the upper head of the pressure vessel is removed along with internal components of the vessel to provide access to the reactor core therein. Water completely fills the open vessel up to the level of the pool disposed thereabove, and a conventional mobile refueling gantry or platform is movable above the pool for carrying replacement fuel bundles to the reactor core and removing spent fuel bundles therefrom. The platform typically travels longitudinally on rails and has a trolley which moves transversely between the rails, and a telescoping mast with a grapple at the bottom end thereof is carried by the trolley so that fuel bundles may be accurately moved underwater to and from the reactor core. The core typically includes a two-dimensional horizontal array of square cells in which square fuel bundles are longitudinally inserted and withdrawn therefrom. Since the elevation distance from the refueling platform to the reactor core is typically relatively large, for example about 30 meters, the difficulty and time required to complete the refueling process is relatively high. Since the fuel bundle fits into its mating cell with close tolerances, a significant amount of operator time is required for close-in accurate positioning the fuel bundle therein. And, in the event of any water currents occurring during the refueling operation, the difficulty of accurately inserting a fuel bundle into its mating cell is further increased. In a typical refueling cycle, about 30 percent of the fuel bundles in the reactor core are replaced. And, a significant portion or all of the remaining fuel is often shuffled or transferred from one position in the core to another for obtaining improved fuel burnup to extend the life thereof. The conventional refueling platform, therefore, must not only remove spent fuel bundles and install replacement fuel bundles, but must also shuffle the remaining fuel bundles between cells to complete the entire refueling operation. Of course, this takes a substantial amount of time which increases costs, as well as increases the amount of radiation exposure time. SUMMARY OF THE INVENTION A reactor refueling mechanism includes a bridge, a trolley on the bridge, and an elevator on the trolley, with the elevator including a mast with a grapple at the lower end thereof. The bridge spans a reactor core and is rotatable therearound, with the trolley being translatable over the bridge. The elevator is also rotatable on the trolley, and the mast is vertically movable for positioning the grapple. The combined movement capability of the mechanism allows the shuffling of fuel bundles between cells in the reactor core, and the accurate placement of square fuel bundles in complementary square cells in the core. |
046577270 | summary | BACKGROUND OF THE INVENTION In response to the accident at Three Mile Island Unit 2, the federal government has required the development of symptom-related emergency operating procedures by utilities which operate nuclear power plants. The main objective of such procedures is to ensure that proper mitigating actions are implemented by licensed operators in the event of an accident. In addition, the criteria for implementing such mitigating actions identified in the emergency operating procedures must be based solely upon indications or symptoms observable by licensed operators using instrumentation within the control room of a nuclear station. The federal government has required the development and installation of instrumentation which can be used by operators in the control room to identify symptoms which are indicative of transient or accident conditions. Prior to the accident at Three Mile Island Unit 2, the typical response by an operator in a control room during a transient was to attempt to identify the event in progress and then to locate and implement a specific event mitigating procedure associated with the identified event. Since the accident at Three Mile Island Unit 2, the federal government has required utilities to develop emergency operating procedures which can by used by operators to mitigate emergency events independent of identification of their causes. The development of these symptom-related emergency operating procedures was intended to remove the possible errors an operator might make in trying to identify an event prior to attempting to mitigate it. Errors in event identification are possible by trained and licensed operators due to the vast numbers and combinations of possible indications which could appear on control room instrumentation due to different types of transients and accidents. The federal government and nuclear industry have acknowledged that the accident at Three Mile Island Unit 2 could have been mitigated had the operators utilized built-in plant features and responded to indications which identified a need for mitigating actions, rather than attempting to identify the event and to implement the proper mitigating procedures for the named event. The means to classify emergency events so as to properly activate state and local governments has not in general improved in the same manner as emergency operating procedures since the accident at Three Mile Island Unit 2. Current federal regulations regarding event classification specify that four levels of emergency events be established and referenced in the emergency response plans for nuclear generating stations. Current federal guidance documents present general emergency class descriptions and detailed lists of example initiating conditions which were considered to be representative of each specific emergency class. As a result of following this guidance utilities have typically developed means to classify emergency events within nuclear power stations by requiring the licensed operator to identify the event in progress and to correlate the identified event to the listing associated with each emergency class presented in the federal guidance. This approach to emergency classification is currently employed by the majority of operating nuclear power stations and is consistent with federal requirements and guidance. This method of event classification requires the correct identification of the emergency event to ensure proper classification and to facilitate appropriate responses by state and local governments to ensure public protection. The result of this approach has been to introduce or maintain the existence of the same source of error, i.e., failure to properly identify the event, which was most significant in the accident at Three Mile Island Unit 2. SUMMARY OF INVENTION Emergency classification must be independent of the event cause and subsequent development so as to minimize the effect of potential errors associated with nuclear plant operators' responses to perceived causes of events which may not be correct and consistent with protecting the general public. Emergency event classification methods and procedures must be consistent with emergency operating procedures. The present invention provides the means to classify emergency events which is independent of event identification, which confirms the area of hazard and auguments emergency operating procedures developed by the utilities operating nuclear generating stations subsequent to the accident at Three Mile Island Unit 2, and constitutes a major improvement in public safety. In addition, the present invention utilizes a computer to accomplish the comparison of symptoms of an emergency event to a logic matrix which corresponds to the various classifications for emergency events, and thus greatly assists operators in event classification. The present invention classifies actual and potential emergency conditions at commercial nuclear generating stations. This emergency event classification process is directly related to three barriers to fission product release to the environment. Since 1970, the design of nuclear power plants has been governed by the General Design Criteria specified in the Code of Federal Regulations, Title 10, Part 50, Appendix A. In particular, Criteria Nos. 10, 13, 14, and 16 identify three fission product barriers which are intented to prevent the uncontrolled release of radioactive material to the environment. The three fission product barriers are the reactor core (intergrity of the fuel cladding), the reactor coolant system pressure boundary, and the containment. The functional integrity of any of the three boundaries is sufficient to prevent the uncontrolled release of radioactive material to the environment. The loss of function any one barrier constitutes a significant reduction in the level of safety of a nuclear generating station. The implementation of the process requires rigorous analysis to determine appropriate values which characterize the functional integrity of the fission product barriers. This process establishes functional performance criteria for each fission product barrier such that the symptoms associated with degradation of any single barrier, or with degradation of several barriers or combinations of barriers are unambiguous, unique and identifiable. A computer is programmed to use the indications of barrier degradation to identify the magnitude of the hazard to the general public which exists when these indications are present. Using the computer, the nuclear power plant operator properly classifies emergencies at nuclear stations and automatically advises offsite authorities of the hazard associated with plant emergency conditions. Definitions regarding the operability of each fission product barrier are included in the license to operate each nuclear generating station. However, the definitions associated with conditions stated in operating licenses are appropriate only during plant operation and cannot be directly utilized to define emergency event classes. The process of implementation of the fission product barrier approach to emergency event classification is the determination of and use of functional indications in a nuclear generating station's control room of fission product barrier integrity. The process is a unique concept which is applicable to all power reactors which are designed in accordance with the Code of Federal Regulations, Title 10, Part 50, Appendix A, General Design Criteria. The process directly relates the loss of fission product barrier function to a specific emergency event classification. Events which are off-normal and which do not represent a loss of function of any of the three fission product barriers, but are of sufficient interest to warrant activation of a nuclear plant''s emergency plan, are classified as a "Unusual Events". Events resulting in the loss of function of one fission product barrier are classified as "Alerts". Events resulting in the loss of function of two fission product barriers are classified as "Site Area Emergencies". Events resulting in the loss of all three fission product barriers are classified as "General Emergencies". The criteria stated within a nuclear generating station's emergency plan for fission product barrier function are properly determined using the present invention to provide margin between limiting conditions for operation associated with characteristics of fission product barrier function (the normal upper limit of certain critical parameters monitored during normal plant operation) and minimum criteria for emergency plan activation and event classification. Such a margin precludes activation of the emergency plan due to simply exceeding a limiting condition for operation associated with a fission product barrier. Additionally, proper selection of the functional criteria for barrier function according to the present invention allows the licensed operator to recognize in a nuclear plant's control room the display of unique symptoms associated with the breach of individual fission product barriers, and all the possible combinations of breaches of fission product barriers. The operator can then readily identify the functional status of all fission product barriers employing the computer, whose output is verified by indications available within each plant's control room. The main objective of emergency preparedness efforts for commercial nuclear generating stations is to limit the radiation dose which might be received by the general public in the event of an accident having offsite radiological consequences. Federal emergency preparedness regulations require classification of any emergency condition according to a graded system commensurate with the hazard presented to the public. Radiation dose to the general public located in the vicinity of a nuclear generating station can be characterized approximately in the following relationship: EQU DOSE.alpha.f(.phi.,t.sub.o,u/p,t.sub.d).times.f(w).times.f(s).times.f(L.sub .F).times.f(L.sub.RCS).times.f(L.sub.C) where, during an event, relatively fixed functions are: f(.phi.,t.sub.o,u/p,t.sub.d) is the source term, a value related to the amount of radioactive material which exists in the fuel, the form of that material and the volatility of the material. The source term is related to reactor power, (.phi.); to the time of operation, (t.sub.o); to fuel characteristics, (u/p); and time after shutdown, (t.sub.d); among other factors; PA1 f(w) is a function related to meteorological conditions at the time of release of radioactive material; PA1 f(s) is a function of site demographic characteristics; PA1 f(L.sub.F) is a function of the leak rate of the reactor fuel cladding; PA1 f(L.sub.RCS) is a function of the leak rate of the reactor coolant system; PA1 f(L.sub.C) is a function of the leak rate of containment systems. And where relatively variable functions are: The above relationship is a qualitative relationship which illustrates how the barrier functions relate to offsite dose calculations, and is not intended to present a strict quantitative relationship. The reactor fuel is contained within the fuel cladding within the reactor coolant system, which is itself contained within the containment system. The fuel cladding (essentially a metallic shell which encapsulates the fuel), the reactor coolant system and the containment system constitute three barriers to release of radioactive material from the reactor fuel to the environment. The complete functioning of any one of these three barriers presents a sufficient obstacle to prevent the release of radioactive materials to the environment. If the numerical value of any one of the three functions f(L.sub.F), f(L.sub.RCS), or f(L.sub.C), presented above is equal to zero or is very nearly zero, the offsite radiation dose to the general public is also very nearly zero or equal to zero since all of the other functions have finite values. The values of f(.phi., t.sub.o,u/p,t.sub.d) and f(s) are determined by considerations unrelated to any nuclear power plant transient or accident and can be considered constants. f(w) may be determined by weather characteristics; it cannot be changed by any action by nuclear station operators and can be considered a constant for purposes of implementation of the present system. The only terms which are influenced by specific transients or emergency events are f(L.sub.F), f(L.sub.RCS), and f(L.sub.C). To present a signficant hazard to the public an accident must significantly influence the leak rate, i.e., the means by which and rate at which radioactive materials are being transported across any of the three barriers. The operators of nuclear power stations are required to determine the magnitude of hazard to the health and safety of the public which exists during any emergency situation and are required to classify the hazard into four categories, progressing from the less serious to the most serious hazard to the public. While it would seem that the quantification of f(L.sub.F), f(L.sub.RCS), and f(L.sub.C) would significantly aid in the determination of the hazard level, the representation or calculation of each of these functions in an absolute manner is extraordinarily difficult due to the complexity of the processes and the number of parameters which relate to status of each barrier. It is possible, however, to empirically assess whether each barrier exists (is functional) and is adequate to provide the degree of protection sought. To accomplish this assessment it is necessary to develop a functional definition of each fission product barrier and then, during any event, to determine if the function is being maintained. This determination of functional requirements for the fission product barriers constitutes the first part of the process of implementation of the fission product barrier emergency event classification and response system. The steps taken, in real time, by the nuclear power plant operator and the computer to utilize nuclear plant instrumentation to identify the functional status of fission product barriers so as to classify an event is the second part of the subject process. Considerations associated with quantification of the function of the fission product barriers, f(L.sub.F), f(L.sub.RCS), and f(L.sub.C), are now discussed. f(L.sub.F) Function of the Reactor Fuel as a Fission Product Barrier The fuel of a light water nuclear power reactor is comprised of ceramic pellets which are enclosed in metal tubes, referred to as fuel cladding. The length of these tubes is typically between 10 and 14 feet, with a length of 12 feet being most common. Each tube is a pressure vessel and as such is a leak-tight enclosure. Operational limitations are placed upon nuclear power plants so as to prevent the degradation of the integrity of the fuel cladding. During operation of the reactor the fuel material is transformed by the fission process to yield a distribution of various elements (fission products) some of which are gaseous. A portion of the gaseous fission products migrates to the void between the fuel matrix and the cladding tube and mixes with inert gases located in that void area or gap. Individual fuel rods are verified to be leak tight upon initial fabrication. During operation within a reactor, mechanical wear, internal pressure generation, metal fatigue and creep due to pressure variations result in minor degradation of the fuel cladding. During accident conditions, changes in the environment around the fuel due to changes in coolant flow rate, temperature and pressure may result in significant stresses upon the cladding material such that rupture, local melting, and chemical interactions with other materials can occur which may result in the loss of integrity of individual tubes containing fuel. The loss of that integrity results in the transport of radioactive material from within the tube to the area outside of the tube, i.e., to the reactor coolant system in which all fuel is contained. Typical power reactors contain tens of thousands of individual fuel tubes assembled in bundles called fuel elements. Due to the number of such fuel tubes it is not possible to monitor each tube for indication of pressure or any other physical parameter so as to determine the integrity of individual tubes. Integrity is inferred by the determination of the concentration of radioactive materials within the reactor coolant system using plant instrumentation. Since the only source of fission products is from the fuel, the detection of fission products within the reactor coolant system indicates the existence of degradation of one or more fuel tubes. Low levels of degradation have been considered in the design of nuclear power plants and limitations regarding the maximum allowable reactor coolant system radioactive materials concentrations have been established for each reactor so as to assure adequate functioning of the fuel cladding as a barrier to significant radioactive material transport and possible release of radioactive material to the environment during normal operation. The establishment of a numerical value for the concentation of radioactive material in the reactor coolant system in effect establishes criteria for consideration that the integrity of the fuel cladding as a barrier to fission product release is adequate and therefore that the barrier is functional. The invention establishes the criteria for selecting a particular radioactive material concentration as an indication that the fuel barrier is functional for the purpose of emergency event classification. The amount of radioactive material located within the fuel cladding is overwhelmingly the largest source of radioactive material within a nuclear power plant. The gaseous fission products located within the gap between the fuel pellets and the cladding inner surface amount to millions of curies of radioactive material. The allowable concentration of radioactive material present within the reactor coolant may amount to thousands of curies with normal concentrations being typically a few hundred curies or less. The process establishes the functional definition of the fuel as a fission product barrier by examining in detail many characteristics of a specific nuclear power plant related to plant systems, equipment performance characteristics, equipment locations, operating procedures, instrumentation displays within the control room, accident analysis results, and normal plant operating limitations. f(L.sub.RCS) Function of the Reactor Coolant System as a Fission Product Barrier The reactor coolant system of a nuclear power plant operates at high temperature and pressure so as to provide thermal energy in the form of heated steam to a turbine generator for the purpose of generating electricity. The reactor coolant system contains the reactor coolant water which has an average temperature in the range of 550 degrees Fahrenheit. All reactor coolant systems include instrumentation to inform the operator in the control room that adequate coolant inventory exists within the system. Minor leakage from the reactor coolant system occurs due primarily to the large number of mechanical seals on the system and the high differential pressure across those seals. The normal reactor coolant inventory is maintained by pumping systems which provide additional coolant to the system as necessary to maintain proper inventory. In the event of significant degradation of the reactor coolant system pressure boundary, additional pumping systems may be required to operate to maintain an adequate coolant inventory within the reactor coolant system. The reactor coolant system is located within a pressure vessel referred to as the containment. The environmental conditions and concentration of radioactive materials within the containment are monitored by the reactor operator. Degradation of the reactor coolant system pressure boundary results in the transport of reactor coolant to areas within the plant which would not normally experience the existence of radioactive materials in the concentration normally present within the reactor coolant system. In addition, since reactor coolant is at a high temperature, the transport of a significant amount of reactor coolant across the reactor coolant system pressure boundary results in significant energy transfer as well as the transfer of mass. Since the transfer of energy and mass will be to an enclosed system, the environmental conditions, radioactive material concentrations and mass of liquid located within the enclosed recipient system will reflect the transport of reactor coolant. The function of the reactor coolant system as a fission product barrier is manifested by the absence of the transport of radioactive materials, energy and mass into enclosed systems of a nuclear power plant which are designed to contain significant leakage or rupture of reactor coolant. In addition the ability to maintain reactor coolant inventory employing normal operating systems is also indicative of the integrity of the reactor coolant system and its ability to function as a fission product barrier. The specific characteristics of a breach of the reactor coolant system fission product barrier coupled with the operating characteristics of the reactor coolant system determine the specific combination of indications which will be manifested. Regardless of the type of failure or type of reactor considered, the invention establishes a characteristic description of symptoms exhibited by a loss of function of the fission product barrier associated with the reactor coolant system. f(L.sub.C) Function of the Containment as a Fission Product Barrier The containment of a nuclear power plant is a large pressure vessel designed to contain the energy and mass release resulting from a major rupture of the reactor coolant system, among other considerations. The environmental conditions within the containment provide information to the nuclear plant operator relative to the function of the reactor coolant system as a fission product barrier. Similarly the environmental conditions in the buildings which surround or are contiguous with the containment provide information regarding the integrity of the containment, and thus of its ability to function as a fission product barrier. Such buildings are equipped with monitoring equipment which functions continuously to inform the operator of the concentration of radioactive materials within those areas. The radioactive material concentration within plant buildings is low in comparison with the radioactive material concentration in containment. The identification of the presence of significant radioactive material concentrations beyond the containment fission product barrier is indicative of a degraded barrier. The transport of radioactive material to areas beyond the containment boundary constitutes the lack of containment and lack of functioning of the containment as a fission product barrier. Federal requirements for nuclear power stations specify enclosure of the reactor coolant system totally within the containment. The process compares concentrations of radioactive material within containment during an emergency event and the concentration of radioactive material being transported to the environment to define the function of containment as a fission product barrier. |
047675942 | abstract | An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs. |
claims | 1. A computer program product comprising a non-transitory computer readable storage medium having readable program code embodied in the medium, the computer program product includes at least one component operable to:measure a change in at least one electrical characteristic of a first sensing device caused by heat generated at a first heating device;calculate a temperature of the first heating device using the measured change in the at least one electrical characteristic;determine an offset between the first heating device and at least a second heating device, each having a different number of contacts; andextrapolate the offset to zero contacts. 2. The computer program product of claim 1, wherein the first heating device and the first sensing device are arranged on the SiGe island. 3. The computer program product of claim 1, wherein the first sensing device, the first heating device, and a common source contact arranged on a SiGe island constitute an apparatus, and the common source contact leads to a source of both the first heating device and the first sensing device. 4. The computer program product of claim 1, wherein the calculating is based on a temperature versus power level relationship for the first heating device using the measured change in the at least one electrical characteristic of the of the first sensing device at different power levels and different distances from the first heating device. 5. The computer program product of claim 1, wherein the at least one electrical characteristic is a sub-threshold voltage slope. 6. The computer program product of claim 1, wherein the measuring includes measuring a series of measurements between the first sensing device and the first heating device at varying amounts of power applied to the first heating device. 7. The computer program product of claim 1, wherein the measuring includes measuring a series of measurements between the first sensing device and the first heating device at varying distances. 8. The computer program product of claim 1, wherein the first heating device and the first sensing device are a field effect transistor. 9. The computer program product of claim 1, wherein the at least one electrical characteristic comprises drain current versus gate bias voltage. 10. The computer program product of claim 1, wherein the at least one electrical characteristic comprises sub-threshold voltage slope swing. 11. The computer program product of claim 1, wherein the at least one component is further operable to calibrate the first sensing device by measuring a particular electrical characteristic of an active region of the first sensing device held at a known ambient temperature. 12. The computer program product of claim 1, wherein:the measuring includes:measuring the change in the at least one electrical characteristic of the first sensing device caused by the heat generated at the first heating device with a first number of contacts;measuring the change in the at least one electrical characteristic of the first sensing device caused by the heat generated at the first heating device with a second number of contacts; andthe calculating includes:calculating the temperature of the first heating device with the first number of contacts and the second number of contacts using the measured change in the at least one electrical characteristic; andextrapolating results of the measurement obtained with the first number of contacts and the second number of contacts to the zero contacts. 13. The computer program product of claim 1, wherein the measuring includes establishing an amount of temperature change per contact between the first heating device and the second heating device having the different number of contacts and the calculating includes extrapolating results of the measurement step to the zero contacts to determine an actual device temperature without an offsetting effect of the contacts. |
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abstract | Soft X-rays are very suitable for the examination of biological samples by means of an X-ray microscope. The X-rays are generated by focusing an electron beam onto a fluid jet, thus producing a very small electron focus on the jet and hence a very small monochromatic X-ray spot. The electron spot can be obtained by means of a standard electron microscope (a SEM) or by means of a standard electron gun for a cathode ray tube (a CRT gun). The imaging optical elements in the X-ray microscope may be Fresnel zone plates. |
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claims | 1. A particle beam apparatus defining an optical axis and comprising:an illuminating system for illuminating an object to be positioned in an object plane with a beam of charged particles which splits into a null beam and higher diffraction orders at said object;an objective arranged along said optical axis for imaging said object illuminated by said illuminating system;said illuminating system being configured to generate, during operation, an annularly-shaped illuminating aperture in a plane fourier transformed to said object plane;said objective having a focal plane facing away from said object plane;a phase-shifting element mounted in said focal plane or a plane conjugated thereto;said phase-shifting element including an einzel lens having first and second outer electrodes arranged along said optical axis and an inner electrode arranged therebetween;said electrodes being arranged and charged with electrical potential, during operation, so as to cause the potential at said optical axis to correspond to the potential at said first and second outer electrodes of said einzel lens; and,wherein said phase-shifting element further comprises an ancillary electrode on or near said optical axis. 2. The particle beam apparatus of claim 1, wherein said ancillary electrode is charged with the potential of said first and second outer electrodes. 3. The particle beam apparatus of claim 2, further comprising a manipulator for holding said ancillary electrode and positioning said ancillary electrode perpendicularly to the direction of said optical axis. 4. The particle beam apparatus of claim 1, wherein said phase-shifting element is configured to impart, during operation, a phase shift to said null beam relative to said higher diffraction orders; and, said phase-shifting element is further configured to, during operation, not influence or only slightly influence said higher diffraction orders running closer to said optical axis than said null beam. 5. A particle beam apparatus defining an optical axis and comprising:an illuminating system for illuminating an object to be positioned in an object plane with a beam of charged particles which splits into a null beam and higher diffraction orders at said object;an objective arranged along said optical axis for imaging said object illuminated by said illuminating system;said illuminating system being configured to generate, during operation, an annularly-shaped illuminating aperture in a plane fourier transformed to said object plane;said objective having a focal plane facing away from said object plane;a phase-shifting element mounted in said focal plane or a plane conjugated thereto;said phase-shifting element including an einzel lens having first and second outer electrodes arranged along said optical axis and an inner electrode arranged therebetween;said electrodes being arranged and charged with electrical potential, during operation, so as to cause the potential at said optical axis to correspond to the potential at said first and second outer electrodes of said einzel lens;wherein said inner electrode comprises a plurality of segments;potentials of opposite polarity are applied to respective ones of each two of said segments lying mutually opposite each other with reference to said optical axis; and,wherein said plurality of segments are electrically insulated from each other. 6. The particle beam apparatus of claim 5, further comprising a deflection system for time sequentially generating said annularly-shaped illumination aperture; said deflection system being mounted in a plane conjugated to said object plane on the side thereof facing toward said illumination system; and, a voltage supply configured to cause a potential to be applied to said segments alternately or rotatingly. 7. The particle beam apparatus of claim 5, wherein said inner electrode comprises two segments lying opposite each other with respect to the optical axis and having respective edges facing toward said optical axis; and, said edges are at different distances from said optical axis. 8. The particle beam apparatus of claim 7, further comprising a deflection system for time sequentially generating said annularly-shaped illumination aperture; said deflection system being mounted in a plane conjugated to said object plane on the side thereof facing toward said illumination system; and, a voltage supply configured to cause a potential to be applied to said segments alternately or rotatingly. 9. A particle beam apparatus defining an optical axis and comprising:an illuminating system for illuminating an object to be positioned in an object plane with a beam of charged particles which splits into a null beam and higher diffraction orders at said object;an objective arranged along said optical axis for imaging said object illuminated by said illuminating system;said illuminating system being configured to generate, during operation, an annularly-shaped illuminating aperture in a plane fourier transformed to said object plane;said objective having a focal plane facing away from said object plane;a phase-shifting element mounted in said focal plane or a plane conjugated thereto;said phase-shifting element including a carrier; and, an annularly-shaped electrode accommodated in said carrier;said annularly-shaped electrode having a side facing away from said optical axis in a radial direction;said electrode having an edge on said side thereof;said edge and said carrier conjointly defining an annular gap therebetween wherein said phase-shifting element imparts a phase shift to the beam of charged particles split into higher diffraction orders in response to a potential applied thereto; and,wherein said phase-shifting element does not influence, or only slightly influences, said null beam. 10. The particle beam apparatus of claim 9, wherein said annularly-shaped electrode is a first annularly-shaped electrode; said phase-shifting element further comprises a second annularly-shaped electrode; and, said first and second annularly-shaped electrodes are electrically insulated from each other. 11. The particle beam apparatus of claim 10, wherein said second annularly-shaped electrode is arranged radially outside of said first annularly-shaped electrode; and said carrier is at a potential and, during operation, said potential of said carrier is applied to said second annularly-shaped electrode; and wherein said particle beam apparatus further comprises a voltage source for applying a potential to said first annularly-shaped electrode deviating from said potential of said carrier. 12. The particle beam apparatus of claim 11, wherein one of said first and second annularly-shaped electrodes comprises a plurality of annularly-shaped segments; and wherein said particle beam apparatus further comprises: a deflection system for time sequentially generating said annularly-shaped illumination aperture; said deflection system being mounted in a plane conjugated to said object plane on the side thereof facing toward said illumination system; and, a voltage supply configured to cause a potential to be applied to said segments alternately or rotatingly. 13. The particle beam apparatus of claim 9, wherein said phase-shifting element comprises a plurality of said annularly-shaped electrodes accommodated in said carrier; said annularly-shaped electrodes have respective sides facing away from said optical axis in a radial direction; said electrodes have respective edges on corresponding ones of said sides thereof; and, said edges and said carrier conjointly define respective annular gaps therebetween. 14. The particle beam apparatus of claim 13, wherein each one of said plurality of annularly-shaped electrodes comprises a plurality of annularly-shaped segments. |
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059463651 | summary | The invention relates to a process for producing a guide tube of a nuclear reactor fuel assembly, a mandrel for forming the guide tube, and the guide tube obtained. Fuel assemblies, and in particular fuel assemblies for water-cooled nuclear reactors, usually comprise a framework in which are inserted and maintained fuel rods constituting a cluster in which the rods are parallel to one another. The framework in particular comprises guide tubes arranged parallel to the rods of the cluster which constitute both structural elements of the framework and elements for guiding absorber rods and control rods used for regulating in service the reactivity of the core of the nuclear reactor constituted by juxtaposed fuel assemblies. The fuel assemblies are disposed in the core of the nuclear reactor in a vertical position, i.e. in such position that the rods of the cluster of the assembly and the guide tubes are vertical. The guide tubes guiding the absorber rods of the control clusters of the nuclear reactor have, in an end part constituting the lower end part of the guide tubes of the assemblies in the service position in the core, a reduced diameter or a narrowing of the section which has for purpose to create a region in which the absorber rods of the control rods of the reactor are braked or slowed down in the case of a dropping of the control rods subsequent to a stoppage of the nuclear reactor which may be a programmed stoppage or a stoppage for an incidental cause. This device for braking the control rods by a throttling of the cooling fluid of the reactor between the absorber rods of the control bar and the lower part of reduced diameter of the guide tubes operates by a braking effect named a dash-pot effect. The lower part of the guide tubes of the fuel assemblies is highly stressed upon the dropping of the control rods, in particular by compression of the cooling fluid in contact with the internal walls of the guide tubes of the fuel assembly in their region of reduced diameter. In the case of transitional periods of operation of the nuclear reactor, the assembly may be slightly raised by the cooling fluid and the force of inertia of the fuel assembly when it descends may cause a flexion and/or torsion of the part of reduced diameter of the guide tubes which constitutes a weak part of the tube, when the reduction in the diameter of the lower part of the guide tube is obtained by a necking of the tube, the wall thickness being substantially constant throughout the length of the tube. It has therefore been proposed to employ guide tubes whose lower end part or base is reinforced in such a manner as to withstand the stresses created by the dropping of the control rods and in the course of the transitional periods. A first solution may comprise reinforcing the lower part of the guide tube by a tubular sleeve which is coaxial with the guide tube, placed against and welded to the guide tube at both ends. However, the welds of the reinforced parts of the guide tubes of zirconium alloy which result in residual stresses, may diminish the mechanical resistance and corrosion resistance of the guide tubes. Further, defects in the alignment may occur between the two tube sections to be assembled. It was therefore proposed in FR-A-2714516 to produce a guide tube with a reinforced base in one piece. The guide tube has a cylindrical outer surface having of substantially constant diameter, except for end regions for fixing the guide tube to the terminal elements of the fuel assembly, and the wall of the tube has a thickness which is increased in a lower part whose length may be between 10 and 30% of the total length of the tube. The guide tube is obtained by producing a tubular blank whose thickness is substantially constant and this thickness is reduced in a fraction of its length, between 70 and 90%, while maintaining a constant inside diameter. The part of the blank which had maintained its original thickness is then upset toward the interior so as to constitute a tube having a constant outside diameter. The reduction in the thickness and the upsetting may be effected by rotary hammering operations. This method of obtaining guide tubes of zirconium alloy has been found to be delicate to carry out. Further, between the part of the tube, termed the main or body part, having a first thickness which was obtained by reduction in the thickness of the blank and the second part of the tube which was upset toward the interior, there is a transition region having an internal surface in the shape of a conical chamfer whose vertex angle has a value of around 10.degree.. The presence of a discontinuity in the internal cylindrical surface of the guide tube may reduce the quality of the guiding of the absorber rods and limit the dropping speed of the control rods. Further, this transition region may be a weakened region of the guide tube. For producing tubes sheathing the rods of the fuel assemblies of nuclear reactors of zirconium alloy, a tubular blank is formed in the usual manner by rolling in a pilgrim or pilger rolling mill. The rolling permits obtaining the sheathing tube with its final dimensions in the course of a plurality of successive rolling operations. This process may also be employed for the shaping of guide tubes of zirconium alloy. However, such a process has never been employed heretofore for producing a guide tube having a cylindrical outside surface of constant diameter, a body or main part of the guide tube having a first thickness and a lower end part of the guide tube having a second thickness exceeding the thickness of the main part. An object of the invention is to provide a process for producing a guide tube for a nuclear reactor fuel assembly having a substantially constant outside diameter and comprising at least one main part which has a first wall thickness and a reinforced part, in a fraction of the length of the guide tube, which has a second wall thickness exceeding said first wall thickness, comprising rolling a tubular blank on a mandrel in a pilgrim rolling mill. For this purpose, the process further comprises: effecting, in a first stage, a rolling of a first section of the blank on a first part of the mandrel with a reduction of the outside diameter of the blank to the outside diameter of the guide tube and a reduction of the wall thickness of the blank to one of said first wall thickness and second wall thickness, PA1 displacing the mandrel in the axial direction of the blank, and PA1 effecting, in a second stage, a rolling of a second section of the blank on a second part of the mandrel spaced from said first part in the axial direction of the mandrel, with a reduction of the outside diameter of the blank to the outside diameter of the guide tube and a reduction of the wall thickness of the blank to the other of said first wall thickness and second wall thickness. In order to explain the invention, there will now be described, by way of a non-limitative example with reference to the accompanying drawings, the process according to the invention, the forming mandrel employed for carrying out the process and the guide tube obtained. |
053234308 | description | DETAILED DESCRIPTION OF EMBODIMENTS OF THE INVENTION In a vessel 1 (FIG. 1) of the Dodewaard nuclear reactor in the Netherlands, provided with an inlet nozzle 2 for water and an outlet nozzle 3 for steam, is provided a reactor core 4, in which first control means or rods 5 are movable. In the vessel 1 there is formed an interface I between a liquid phase (L) and steam or gas phase (S), which may be rather turbulent. The interface I is far above the core 4 and above the chimney 8. Secured to the inner wall of the vessel 1 is a gauge 6 in which sensors are provided which are connected to the outside through a nozzle 7 in the top of the vessel. Inside the gauge 6, BICOTH sensor wires are provided according to the arrangement and design of FIG. 2A. For further details of the sensors, see Ara et al.'s U.S. Pat. No. 4,423,639. Further, low level sensor wires are provided for responding to very low levels of the water in the reactor. Inside the water level gauge 6, there exists an interface between liquid and steam phase, which corresponds accurately to the collapsed level in the vessel. When the interface is situated between two thermocouple junctions of a BICOTH sensor there will be a different temperature at the uppermost thermocouple junction of the pair than at the undermost thermocouple junction as a result of poor heat transfer to the surrounding steam area. This temperature difference produces a positive voltage output of the sensor in situations wherein the interface is between two junctions of a thermocouple pair, or a zero voltage output wherein it is under or above two such junctions. The sensor output signals are supplied to a signal processing and conditioning unit. FIG. 2A shows a BICOTH arrangement and design for the measuring of levels of the water-steam interface between +70 cm and -40 cm relative to a zero level which is situated exactly 55 cm above the top of the chimney. In this arrangement of FIG. 2A, eight BICOTH sensors are used. In another embodiment of the present invention according to FIG. 4, a sensor known as a TRICOTH sensor is used in which a central heater wire 21 is insulated by an insulating sheet 22 from a heater sheet 23, outwardly of which eight wires 24 are uniformly embedded in a sheet 25 of Al.sub.2 O.sub.3, which is enclosed by a cladding 26. In this embodiment the heating occurs uniformly to the eight wires such that they will all obtain substantially the same amount of heat. This arrangement also avoids electrical interference between the heater wire and the sensor lines. Apart from a common wire 24 of Alumel and a wire tc1 of Chromel, six of the wires 24 each comprise three thermocouple materials, such that with the arrangement of FIG. 5, as can be seen from FIG. 6, 23 different levels can be measured. The wire tc1 is used just to measure water temperature, forming a thermocouple at the bottom of FIG. 5 with the common return wire com. The wires 1a, 1b, 2a, com, tc1, 2b, 3a, 3b are the eight wires 24 in FIG. 5. At the bottom, the eight wires are interconnected. Each crossbar on the wires in FIG. 5 represents one of the above-mentioned thermocouple junctions. The table at the right in FIG. 6 indicates that 23 levels are uniquely distinguishable by the ternary code structure at the left in FIG. 6. The TRICOTH sensor is stand-alone; only one sensor needs to be used, having 6 ternary coding signal lines, as shown in FIGS. 4, 5 and 6. It uses three types of thermocouple materials, alumel, constantan, and chromel (FIG. 5), and produces positive signals, active zero signals, and negative signals in each sensor line, such that a given water level on the outside of the TRICOTH sensor generates uniquely a 6-digit ternary code (e.g., level 16 generates the code -+-0-0 as shown in FIG. 6). The TRICOTH sensor has a central coaxial heater which is the heat source over the length of the sensor and which is a central component for the generation of the sensor codes. Compared to the BICOTH system consisting of 8 different sensors, the stand-alone TRICOTH sensor has advantages of simplicity as well as digital and analog precision. This is explained as follows. Consider the TRICOTH system in an upright position, the water level at interval 16 (see FIG. 6), and the central heater switched on. Assume a stepwise axial temperature transition inside the sensor at the position of the water level, i.e., a lower temperature, almost the water temperature, below the water level, and a higher temperature (20 degrees C. higher for example) above the water level. Then, only the ternary code -+-0-0 is produced. However, due to axial conduction of heat, a smooth temperature transition curve exists inside the sensor instead of a stepwise transition. This temperature transition function moves with the water level. Therefore, analog signal transition functions (FIGS. 7 and 8) are created in the sensor lines. When the thermocouple junction of a sensor line is in range of the transition function, then the normalized amplitude of the sensor signal is an accurate analog measure of the position of the water level relative to that junction. Therefore, for all level intervals, the combination of ternary codes and analog signal information provides unique and continuous water level readings over the whole measuring range of the hybrid digital-analog TRICOTH sensor. |
summary | ||
claims | 1. A method for heat treating a metal tube or pipe, comprising:heat treating a metal tube or pipe which is accommodated in a heat treatment furnace, the metal tube or pipe being laid down on a plurality of cross beams, the plurality of cross beams being arranged along a longitudinal direction of the metal tube or pipe such that a distance between adjacent ones of the cross beams is in a range of 200 to 2500 mm, the metal tube or pipe being in contact with the cross beams, whereinas the cross beams, cross beams having a convex top surface are used, the convex top surface having a radius of curvature in a range of 50 to 300 mm. 2. A method for heat treating a metal tube or pipe comprising:heat treating a metal tube or pipe, which is accommodated in a heat treatment furnace, the metal tube or pipe being laid down on a plurality of the cross beams with spacers having a convex top surface interposed therebetween, the plurality of cross beams being arranged along a longitudinal direction of the metal tube or pipe such that a distance between adjacent ones of the cross beams is in a range of 200 to 2500 mm, the metal tube or pipe being in contact with the spacers, whereinthe convex top surface has a radius of curvature in a range of 50 to 300 mm. 3. The method for heat treating a metal tube or pipe according to claim 1, wherein the metal tube or pipe has a composition consisting of, in mass %, C: 0.15% or less, Si: 1.00% or less, Mn: 2.0% or less, P: 0.030% or less, S: 0.030% or less, Cr: 10.0% to 40.0%, Ni: 8.0% to 80.0%, Ti: 0.5% or less, Cu: 0.6% or less, Al: 0.5% or less, and N: 0.20% or less, the balance being Fe and impurities. 4. A heat treatment furnace that performs heat treatment on a metal tube or pipe, the metal tube or pipe being accommodated in the heat treatment furnace in a state where the metal tube or pipe is laid down on a plurality of cross beams, the plurality of cross beams being arranged along a longitudinal direction of the metal tube or pipe such that a distance between adjacent ones of the cross beams is in a range of 200 to 2500 mm, the metal tube or pipe being in contact with the cross beams whereinas the cross beams, cross beams having a convex top surface are used, the convex top surface having a radius of curvature in a range of 50 to 300 mm. 5. A heat treatment furnace, that performs heat treatment on a metal tube or pipe, the metal tube or pipe being accommodated in a heat treatment furnace in a state where the metal tube or pipe is laid down on a plurality of cross beams with spacers having a convex top surface interposed therebetween, the plurality of cross beams being arranged along a longitudinal direction of the metal tube or pipe such that a distance between adjacent ones of the cross beams is in a range of 200 to 2500 mm, the metal tube or pipe being in contact with the spacers, whereinthe convex top surface has a radius of curvature in a range of 50 to 300 mm. 6. The method for heat treating a metal tube or pipe according to claim 2, wherein the metal tube or pipe has a composition consisting of, in mass %, C: 0.15% or less, Si: 1.00% or less, Mn: 2.0% or less, P: 0.030% or less, S: 0.030% or less, Cr: 10.0% to 40.0%, Ni: 8.0% to 80.0%, Ti: 0.5% or less, Cu: 0.6% or less, Al: 0.5% or less, and N: 0.20% or less, the balance being Fe and impurities. |
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claims | 1. A passive containment spray system comprising:a spray coolant storage unit installed within a containment accommodating a reactor vessel and configured to store coolant and air therein, the air within the spray coolant storage unit fluidly communicates with air within the containment such that the spray coolant storage unit maintains equilibrium of pressure with the containment;a spray pipe that is installed within the containment in such a manner that when an accident occurs, the coolant supplied from the spray coolant storage unit is sprayed into the containment through the spray pipe due to an increase in pressure within the containment;a connection pipe having one end inserted into the spray coolant storage unit to provide a flow path along which the coolant flows, and the other end connected to the spray pipe to supply the coolant passively to the spray pipe when the pressure within the containment increases due to an occurrence of an accident and a flow of the coolant occurs therein, anda check valve installed in the spray pipe such that steam discharged into the containment or air is prevented from being introduced through the spray pipe into the connection pipe, the check valve being opened in a direction toward the spray pipe by a flow of the coolant from the spray coolant storage unit such that the coolant within the spray coolant storage unit flows through the spray pipe, whereinthe connection pipe includes:an upward flow path portion inserted into the spray coolant storage unit and extending up to a predetermined height such that the flow of the coolant from the spray coolant storage unit to the spray pipe is prevented from occurring within a normal plant operation pressure range for the containment; anda downward flow path portion extending downward from the upward flow path portion and connected to the spray pipe such that the coolant is supplied continuously to the spray pipe therethrough due to a difference of a gravitational head of water, andthe check valve is configured to form a pressure difference between an inside of the connection pipe and a target space of spraying in the containment until the check valve is opened when the flow of coolant occurs at a height of the upward flow path portion or above,the passive containment spray system further comprises:a diverged pipe having one end connected to the downward flow path portion and the other end exposed to air; andan isolation valve installed in the diverged pipe, and configured to be opened by an operator to prevent an occurrence of the flow of the coolant from the spray coolant storage unit by forming an equilibrium of pressure between an inside of the connection pipe and an outsider of the connection pipe when a nuclear power plant is in a normal plant operation condition, or when the spray coolant storage unit is in maintenance,an intermediate cavity unit installed around the other end of the connection pipe to enhance the flow of the coolant that occurs in a direction from the spray coolant storage unit to the connection pipe, the intermediate cavity unit generating a difference in pressure from the spray coolant storage unit, and connected to the spray pipe to supply the coolant that passes through the connection pipe to the spray pipe,wherein the spray coolant storage unit and the intermediate cavity unit do not share a common wall. 2. The passive containment spray system of claim 1, whereinthe upward flow path portion is configured to provide a flow path along which the coolant flows when the pressure within the containment increases; andthe downward flow path portion is configured to provide a flow path along which the coolant flows when the pressure within the containment increases and the flow of the coolant occurs at the height of the upward flow path portion or above. 3. The passive containment spray system of claim 2, wherein a cross-sectional area provided in the upward flow path portion and a cross-sectional area provided in the downward flow path portion are different from each other to facilitate discharging of gas that is filled in the connection pipe. 4. The passive containment spray system of claim 1, further comprising a check valve installed in a pipe that is connected to an upper portion of the intermediate cavity unit, and opened to discharge gas within the intermediate cavity unit when pressure within the intermediate cavity unit is greater than that within the containment due to the coolant through the connection pipe. 5. The passive containment spray system of claim 4, further comprising an orifice installed in the pipe to limit an amount of flowing fluid discharged through the check valve. 6. The passive containment spray system of claim 4, wherein at least one of the check valve installed in the pipe connected to the upper portion of the intermediate cavity unit and the cheek valve installed in the spray pipe is provided in plurality to prevent the passive containment spray system from malfunctioning due to a single failure. 7. The passive containment spray system of claim 1, further comprising at least one spray nozzle connected to the spray pipe to spray the coolant into the containment therethrough. 8. The passive containment spray system of claim 1, wherein the spray coolant storage unit is installed at a predetermined height inside of the containment to allow for spraying of the coolant due to a gravitational head of water, and is maintained in an opened state to achieve equilibrium of pressure between the spray coolant storage unit and the containment. 9. A nuclear power plant comprising:a reactor vessel;a containment that is installed outside of the reactor vessel such that radioactive material is prevented from releasing from the reactor vessel to outside of the containment; anda passive containment spray system,wherein the passive containment spray system includes:a spray coolant storage unit installed within a containment accommodating the reactor vessel and configured to store coolant and air therein, the air within the spray coolant storage unit fluidly communicates with air within the containment such that the spray coolant storage unit maintains equilibrium of pressure between the spray coolant storage unit and with the containment,a spray pipe that is installed within the containment in such a manner that when an accident occurs, the coolant supplied from the spray coolant storage unit is sprayed into the containment through the spray pipe due to an increase in pressure within the containment building, anda connection pipe having one end inserted into the spray coolant storage unit to provide a flow path along which the coolant flows, and the other end connected to the spray pipe to supply the coolant passively to the spray pipe when the pressure within the containment increases due to an occurrence of an accident and a flow of the coolant occurs therein, anda check valve installed in the spray pipe such that steam discharged into the containment or air is prevented from being introduced through the spray pipe into the connection pipe, the check valve being opened in a direction toward the spray pipe by a flow of the coolant from the spray coolant storage unit such that the coolant within the spray coolant storage unit flows through the spray pipe, whereinthe connection pipe includes:an upward flow path portion inserted into the spray coolant storage unit and extending up to a predetermined height such that the flow of the coolant from the spray coolant storage unit to the spray pipe is prevented from occurring within a normal plant operation pressure range for the containment; anda downward flow path portion extending downward from the upward flow path portion and connected to the spray pipe such that the coolant is supplied continuously to the spray pipe therethrough due to a difference of a gravitational head of water, andthe check valve is configured to form a pressure difference between an inside of the connection pipe and a target space of spraying in the containment until the check valve is opened when the flow of coolant occurs at a height of the upward flow path portion or above,the passive containment spray system further comprises:a diverged pipe having one end connected to the downward flow path portion and the other end exposed to air; andan isolation valve installed in the diverged pipe, and configured to be opened by an operator to prevent an occurrence of the flow of the coolant from the spray coolant storage unit by forming an equilibrium of pressure between an inside of the connection pipe and an outsider of the connection pipe when a nuclear power plant is in a normal plant operation condition, or when the spray coolant storage unit is in maintenance,an intermediate cavity unit installed around the other end of the connection pipe to enhance the flow of the coolant that occurs in a direction from the spray coolant storage unit to the connection pipe, the intermediate cavity unit generating a difference in pressure from the spray coolant storage unit, and connected to the spray pipe to supply the coolant that passes through the connection pipe to the spray pipe,wherein the spray coolant storage unit and the intermediate cavity unit do not share a common wall. |
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claims | 1. An apparatus comprising:one or more processors; anda memory coupled to the processors comprising instructions executable by the processors, the processors operable when executing the instructions to:identify an inputted reference failure rate for a monitored device, the inputted reference failure rate being a first quotient of an amount of failures associated with a population of the monitored device and an amount of time, the inputted reference failure rate associated with an expected operating temperature for the monitored device and an expected communication capacity utilization for the device;communicate with the monitored device while the monitored device is in field operation for determining an actual operating temperature for the monitored device and an actual communication capacity utilization;determine a temperature stress adjustment factor using the expected operating temperature and the actual operating temperature;determine an electrical stress adjustment factor using the expected communication capacity utilization and the actual communication capacity utilization, wherein the processors are further operable to determine the electrical stress adjustment factor by calculating a difference of the expected communication capacity utilization and the actual communication capacity utilization; andoutput an instantaneous failure rate that is a first mathematical product of the inputted reference failure rate, the temperature stress adjustment factor and the electrical stress adjustment factor. 2. The apparatus of claim 1 wherein determining the temperature stress adjustment factor further includes determining a difference of a first reciprocal of the expected operating temperature and a second reciprocal of the actual operating temperature. 3. The apparatus of claim 2 wherein the processors are further operable to determine the temperature stress adjustment factor by calculating a second mathematical product of the difference and a second quotient of a predetermined thermal activation energy for the monitored device and Boltzmann's constant. 4. The apparatus of claim 1 wherein the processors are further operable to determine the electrical stress adjustment factor by calculating a second mathematical product of the difference and a predetermined electrical stress characteristic of the monitored device. 5. The apparatus of claim 1 wherein the apparatus is integrated with a router and the monitored device is a communication component located in the router. 6. The apparatus of claim 5 wherein the outputted instantaneous failure rate corresponds to the communication component. 7. The apparatus of claim 6 wherein the processors are further operable to use the instantaneous failure rate for the communication component to determine an instantaneous failure rate for the router. 8. The apparatus of claim 1 wherein the instantaneous failure rate for the monitored device is outputted to a display device. 9. The apparatus of claim 1 wherein the processors are further operable to:check the integrity of non volatile memory used to store the inputted reference failure rate, the expected operating temperature and the expected communication capacity utilization;initialize random access memory (RAM) with previously stored values;define a reliability sampling period or interval; andstart background tasks. 10. The apparatus of claim 1 wherein the monitored device is located remotely with respect to the apparatus and the apparatus is operable to communicate with the monitored device over a network. 11. The apparatus of claim 1 wherein the processors are further operable to communicate with a traffic byte counter coupled to the monitored device to determine the actual communication capacity utilization. 12. The apparatus of claim 1 wherein the processors are further operable to determine a cumulative reliability indication value by summing the instantaneous failure rate with other instantaneous failure rates for the monitored device that are determined by the apparatus periodically according to a predefined period. 13. The apparatus of claim 1 wherein the processors are further operable to output a field-adjusted Mean Time Between Failures (MTBF) for the monitored device that is determined by adjusting a predetermined MTBF for the monitored device using the actual operating parameters that are measured while the monitored device is used for non-testing purposes in a field environment. 14. The apparatus of claim 1 wherein the outputted instantaneous failure rate is associated with a Mean Time Between Failures (MTBF) for the monitored device. 15. A system comprising:means for inputting a first predetermined Mean Time Between Failures (MTBF) for a monitored device, the first predetermined MTBF based on expected environmental conditions and expected usage parameters for the device;means for measuring actual temperature while the device is being operated in the field;means for identifying actual electrical stress on the monitored device by comparing the actual communication capacity utilization to an expected communication capacity utilization;means for determining a second field-adjusted MTBF for the monitored device, the second field-adjusted MTBF determined by adjusting the first predetermined MTBF according to both a first factor representing a difference between the field-measured actual temperature and an expected operating temperature and a second factor representing a difference between the identified electrical stress and expected electrical stress for the device, the second field-adjusted MTBF being different than the first predetermined MTBF; andmeans for outputting the second field-adjusted MTBF. 16. The system of claim 15 wherein the measurements of the actual temperature and the actual communication capacity utilization are initiated automatically after passage of a predefined time interval. 17. The system of claim 16 wherein the measurements are taken over a predefined duration. 18. The system of claim 15 further comprising:means for determining an instantaneous failure rate by adjusting a reference failure rate for the device according to measurements;means for automatically re-determining the instantaneous failure rate after passage of a predefined time interval; andmeans for identifying a cumulative reliability indication value for the monitored device by summing the initial instantaneous failure rate and the re-determined instantaneous failure rate. 19. The system of claim 15 wherein the expected environmental conditions include expected operating humidity and expected operating ambient temperature. 20. The system of claim 15 wherein the expected usage parameters are based on an expected amount of power cycles applied to the monitored device. 21. The system of claim 15 wherein the expected temperature is an expected operating temperature. 22. The system of claim 15 wherein the expected temperature is an expected ambient temperature. 23. A computer readable medium with instructions embedded therein for causing a processor to implement a reliability determination process including:an initialization module for directing implementation of an initialization process;a determination process and a field condition reliability analysis process for determining one or more operational parameters of a component, the operational parameters determined at least in part by calculating an electrical stress adjustment factor that represents a difference between an expected communication capacity utilization for the component and an actual communication capacity utilization that is based on an output of a traffic byte counter corresponding to the component;a reliability determination runtime module for interfacing with an operating system to calculate one or more field-adjusted Mean Time Between Failures (MTBFs) by adjusting a reference MTBF for the component using the operational parameters and to calculate one or more cumulative reliability index values based on the field-adjusted MTBFs; andan output module for causing the calculated cumulative reliability index values to be displayed to a user. 24. The computer readable medium of claim 23 wherein said initialization module includes instructions for:checking the integrity of non volatile memory;initializing random access memory (RAM) with previously stored values;defining a reliability sampling period or interval; anda background module for starting background tasks. 25. The computer readable medium of claim 24 wherein said background module includes instructions for implementing reliability associated firmware activities. 26. The computer readable medium of claim 24 wherein said background module divides the background tasks into multiple background threads that operate separately. 27. A method comprising:identifying a reference failure rate for a device, the reference failure rate usable for calculating a time between failures statistic associated with the device and based on expected operating parameters for the device;measuring actual operating parameters for the device while the device is operated for non-testing purposes in a field environment for the device, the actual operating parameters including a measured actual communication capacity utilization for the device;determining an electrical stress adjustment factor according to a comparison of the actual communication capacity utilization to an expected communication capacity utilization for the device;determining a custom failure rate by adjusting the reference failure rate based at least in part on the determined electrical stress adjustment factor;outputting a signal for displaying the custom failure rate; andoutputting a field-adjusted Mean Time Between Failures (MTBF) for the device that is determined by adjusting a predetermined MTBF for the device using the actual operating parameters that are measured while the device is operated for the non-testing purposes in the field environment. 28. The method of claim 27 wherein the measured operating parameters are transferred over a network for remote analysis. 29. The method of claim 27 wherein the device is a communication component located in a router. 30. The method of claim 27 wherein the custom failure rate is an instantaneous failure rate for the device measured at a first time and the custom failure rate is summed with other instantaneous failure rates for the device that are measured at second other times to generate a cumulative reliability indication. 31. The method of claim 27 wherein measuring actual operating parameters includes monitoring the actual communication capacity utilization for the device. |
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044118627 | abstract | The present invention relates to a spacer grating for fuel element in a nuclear reactor of the PWR type, said grating being constituted by intersecting plates. Over the greater part of the walls equipped with springs, double springs comprise two active parts on either side of the wall on which the double spring is placed, allowing the symmetrical action of this spring on two different pencils inside two adjacent cells. On the other walls equipped with springs, single springs comprise one active part only, disposed in one of the two cells separated by the wall on which is placed the spring of which the other part is not capable of coming into contact with a fuel pencil. The invention is particularly applicable to the manufacture of fuel assemblies in nuclear reactors of the PWR type. |
abstract | According to various embodiments, an emissions calculation system calculates the emissions resulting from transporting a shipment that includes an individual or group of packages. In certain embodiments, the calculated emissions are based on estimated amounts of fuel used in the transportation process based on historical fuel usage and shipment data. In addition, various embodiments of the system generate and display (or otherwise make available) one or more reports of the calculated emissions information. For example, the reports may provide a total amount of carbon dioxide emitted for a particular shipment or group of shipments. In addition, the report may provide a breakdown of the amount of emissions resulting from various operational activities in the transportation process (e.g., resulting from pickup, delivery, and/or transportation between carrier facilities, and resulting from stationary operational activities, such as sorting or movement within each carrier facility). |
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039327485 | summary | BACKGROUND OF THE INVENTION The present invention relates to a method of determining the distance between an area under fire and the muzzle of the weapon by means of autoradiography of the irradiated firing residues on this area. Determining the distance between a fire arm and the point of impact is a very important problem in criminal investigation. In doubtful cases of suicide, justifiable self-defense, or even when determining whether a specific hole was caused by the impact of a bullet, the immediate environment of the point of impact is a very important source to be investigated. A large number of methods of identification of traces of shots are known; but, so far neither chemical analysis nor purely quantitative determination through neutron activation analysis (NAA) permit a satisfactory assessment of the distance to be made. Among other disadvantages, these methods do not operate in the non-destructive mode because the irradiated filters with the traces of shots had to be destroyed for measurement so that the spurious activity of the matrix was removed. Autoradiography has already been used to identify impacts of bullets as such by irradiating a piece of material fired at and afterwards contacting it with X-ray film. This film was then processed into prints or radiographs, respectively. However, this type of investigation of the hole made by a bullet does not allow any conclusions to be drawn as to the range from which a shot was fired. SUMMARY OF THE INVENTION The method according to the present invention in a general sense relates to the detection of the traces of antimony following the bullet and ejected by the weapon. These are always contained in the primer or in the bullet in addition to barium and lead. These particles accumulate on the material stopping a bullet as the bullet penetrates, and are retained and partly burned into this area. The characteristic distribution of such particles in the area around a bullet hole is unambiguous and cannot be placed there in any other way. As has been shown in qualitative analyses of the bullet, powder, primer and shell, the main source of the traces of antimony is the bullet, some of it stemming from the powder as well as the primer. Most probably a certain quantity is rubbed off the bullet as the bullet very rapidly passes the rifling on the inside of the barrel, follows the bullet as a powder or in molten form and is precipitated around the point of impact. As far as the analysis of traces at the point of impact is concerned, it is a particularly fortunate coincidence that the element antimony is very easy to ascertain after neutron activation. The present invention is concerned with a non-destructive and thus reproducible method which can be used to determine the distance between the point of impact of a bullet and a weapon. The structure of the background at which the bullet was fired should not be able to falsify the result, and the evaluation should be very simple and quick. In the present invention, this problem is solved in the method outlined above in that several carrier areas are fired at from various distances. The carrier areas afterwards are activated by neutron irradiation and then contacted with a film sensitive to nuclear radiation. The series of autoradiographs produced after development of the film are compared with the autoradiograph produced in the same way of the trace on the area under fire of the shot to be investigated and are used as a distance standard for that trace. In this method the carrier areas may be made out of the same material as the area under fire. For the case of a highly radioactive carrier area containing, for instance, antimony, it is suggested in accordance with the present invention that the residues of the shot after irradiation are transferred from the carrier areas and/or the area under fire onto a new carrier which is then used to produce the autoradiographs. In accordance with the present invention a very advantageous method of transfer is contacting the area containing the residues of the shot with an adhesive foil under pressure, afterwards separating the area and the foil from each other and covering the adhesive side of the foil with another foil. It is furthermore suggested in the present invention that the total density and the number of particles be measured either visually or by means of image analyzers in evaluating the autoradiographs as the distance standard. |
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052981967 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to the extraction of tritium and preparation of radioactive waste for disposal. More particularly, the present invention relates to tritium recovery during the process of heating and melting irradiated nuclear targets. 2. Discussion of Background Tritium is a radioactive isotope of hydrogen. Because it is frequently found in the form of a diatomic gas or bound in a water molecule, it is relatively mobile and difficult to recover and contain once it has escaped into the environment. Preventing the escape of tritium is an important priority in the operation of facilities where tritium is created as a product or a by-product. Tritium is created when lithium metal in target bundles is irradiated in a nuclear reactor. Nuclear reactor targets, after use, are referred to as "spent". Spent target bundles must be stored in carefully constructed disposal cells for a few hundred to thousands of years depending on the length of time the radioisotopes remain significantly radioactive and on government requirements. The volume required for long-term storage of radioactive waste should be minimized because of cost and other considerations. Using heat to drive tritium from nuclear targets for recovery is well known. See for example both U.S. Pat. No. 4,532,102 and 3,100,184 issued to Cawley and Abrahams, respectively. A prior art furnace, crucible and target bundle is illustrated in FIG. 1. A target bundle after heating is illustrated in its crucible in FIG. 2. A target bundle 20 is placed in a crucible 22 that holds it and encloses the bundle laterally, the top of the crucible is covered with a perforated plug 24 and has an integral hook 26. Crucible 22 is lowered into a retort 34 by a lid 36 that seals against a top flange 38 of retort 34. Retort 34 is positioned in a furnace 50 having electric resistance heaters 52 inside its wall 54. An inlet 56 and an exit 58 allow the circulation of coolant to prevent the shell 60 of the furnace from overheating. An off-gas pipe 62 allows tritium to be drawn from the interior of retort 34. After heating, bundle 20 occupies a much smaller volume than before heating, as depicted in FIG. 2 where a melted bundle 70 is shown in a crucible 72. The melted bundle, slumped to the bottom of the crucible, is disposed of still in the crucible. Although gases from the furnace are drawn off during heating, opening the furnace causes some tritium to escape to the environment. The current furnace, therefore, is unsatisfactory because it allows too much tritium to escape. Furthermore, the melted bundle comprises only about one tenth of the volume of the crucible. Thus, a substantial fraction of the interior of the crucible, as sent to a waste repository, is empty space. SUMMARY OF THE INVENTION According to its major aspects and broadly stated, the present invention is a crucible and sleeve assembly for holding and laterally enclosing an object, such as a target bundle, during transport to and from an inverted furnace where the object is heated to extract recoverable gases. When the object is heated, it gives off gases and vapors of radioactive materials that are drawn from the furnace interior through a trap filled with metal shavings. The hot vapors of radioactive materials condense on the shavings. Gases passing through the trap are collected in tanks downstream as the product. After the object is heated, it occupies less volume than before it is heated. The crucible is sized to hold and laterally enclose the object after it is heated so that, when a lid is sealed to it, the object is fully enclosed and can be transferred to a disposal or storage facility, the sealed crucible have substantially no empty space within. The crucible and sleeve assembly, on the other hand, is dimensioned to hold and laterally enclose the object before it is heated. Several components of the invention individually and in cooperation act to prevent the loss of gases to the environment. The furnace is inverted so that gases lighter than air tend to remain in the furnace when it is opened to remove the object. The portion of the furnace that includes the opening is housed within a containment. Airlocks prevent contamination of the atmosphere of the containment when the object is introduced into or taken from the containment. Finally, strippers remove gases from the containment atmosphere that have escaped from the inverted furnace. The crucible and sleeve assembly is an important feature of the present invention. The two cooperate to securely hold and laterally enclose the object before it is heated. Although a single container can serve the same function, being able to separate the sleeve from the crucible after the object has been heated and its volume is reduced to the preselected dimension of the crucible alone results in a substantial reduction in the volume of the material actually disposed of and enables the reuse of the sleeve. The crucible and sleeve fit together easily and securely so that full lateral support without leakage is provided. Both have flanges that engage each other, and the sleeve has a skirt that depends from the flange into the crucible. The inverted furnace is another important feature of the present invention. By inverting the furnace, the prevention of the escape of the gases given off as a result of heating, such as tritium when a nuclear target bundle is heated, is substantially simplified. The trap of the present invention is another important feature. The trap contains metal shavings on which the radioactive vapors condense, thus simplifying recovery of radioactive vapors that should not be vented to the atmosphere. Other features and advantages of the present invention will be apparent to those skilled in the art from a careful reading of the Detailed Description of a Preferred Embodiment presented below and accompanied by the drawings. |
claims | 1. A process for the chemical stabilization of a uranium carbide composite material corresponding to:UCx+yC with x≧1 and y>0, placed in a stabilization chamber, comprising the following stages:a stage of rise in temperature of the internal temperature of said chamber to a temperature of between approximately 380° C. and 550° C., said chamber being fed with an inert gas consisting of one or both of argon and nitrogen;a stage of isothermal oxidative treatment at said temperature of between approximately 380° C. and 550° C., said chamber being placed under O2 partial pressure; anda stage of controlling the completion of the stabilization of said composite material which comprises monitoring of the amount of molecular oxygen consumed and/or of carbon dioxide given off or of carbon dioxide and carbon monoxide given off, until at least the achievement of a value of an input set point for the molecular oxygen, of a minimum threshold value for said amount of carbon dioxide or of threshold values for the carbon dioxide and carbon monoxide. 2. The process for the chemical stabilization of a uranium carbide composite material as claimed in claim 1 wherein the stage of controlling the completion of the stabilization additionally comprises monitoring of variation in weight of the composite material based on carbon and uranium in the chamber, an increase in weight being correlated with the oxidation of uranium carbide in progress. 3. The process for the chemical stabilization of a uranium carbide composite material as claimed in claim 1, wherein the stage of controlling the completion of the stabilization is carried out with the application of a rise in temperature of the internal temperature of said chamber and the monitoring of CO2 given off. 4. The process for the chemical stabilization of a uranium carbide composite material as claimed in claim 1, comprising the introduction of a water vapor partial pressure into said chamber before and/or during and/or after the isothermal oxidative treatment stage. 5. The process for the chemical stabilization of a uranium carbide composite material as claimed in claim 4, wherein the stage of controlling the completion of the stabilization further comprises detection of H2 as marker for monitoring an end of oxidation in said chamber. 6. The process for the chemical stabilization of a uranium carbide composite material as claimed in claim 2, wherein the stage of controlling the completion of the stabilization comprises an operation of overpressurizing a plurality of reaction gases present in said chamber so as to accelerate the end of the oxidation of said composite material. 7. The process for the chemical stabilization of a uranium carbide composite material as claimed in claim 6, wherein the stage of controlling the completion of the stabilization additionally comprises a cycle of an operation of overpressurizing and an operation of underpressurizing the reaction gases present in said chamber. 8. The process for the chemical stabilization of a uranium carbide composite material as claimed in claim 2, in which said composite material exhibits a morphology of powder or of porous or dense pellet. 9. The process for the chemical stabilization of a uranium carbide composite material as claimed in claim 2, comprising a preliminary stage of determination of an optimum oxidation temperature by thermogravimetric analysis of a sample of UCx+yC composite material. |
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claims | 1. An apparatus for preventing triple junction breakdown, comprising:a first metal electrode;a second metal electrode;an insulator disposed between the first metal electrode and the second metal electrode, wherein the insulator has at least one surface between the first metal electrode and the second metal electrode that is exposed to a vacuum;a first conductive layer located between the first metal electrode and the insulator, wherein the first conductive layer prevents triple junction breakdown from occurring at an interface of the first electrode, insulator and vacuum; anda second conductive layer located between the second metal electrode and the insulator opposite the first conductive layer, wherein the second conductive layer prevents triple junction breakdown from occurring at an interface of the second electrode, insulator and vacuum. 2. The apparatus according to claim 1, wherein the first and second conductive layers comprise metal particles doped into the insulator. 3. The apparatus according to claim 1, wherein the first and second conductive layers are deposited on the insulator. 4. The apparatus according to claim 1, wherein the first and second conductive layers are bonded onto the insulator. 5. The apparatus according to claim 4, wherein the first and second conductive layers are glued onto the insulator. 6. The apparatus according to claim 1, wherein the first and second conductive layers are joined to the insulator at an atom level without formation of a microscopic gap. 7. The apparatus according to claim 1, further comprises a first O-ring and a second O-ring, wherein the first O-ring is sandwiched between the first conductive layer and the first metal electrode and the second O-ring is sandwiched between the second conductive layer and the second metal electrode. 8. An apparatus for preventing triple junction instability in an ion implanter, comprising:a first metal electrode;a second metal electrode;an insulator disposed between the first metal electrode and the second metal electrode, wherein the insulator has at least one surface between the first metal electrode and the second metal electrode that is exposed to a vacuum that transports an ion beam generated by the ion implanter;a first conductive layer located between the first metal electrode and the insulator, wherein the first conductive layer prevents triple junction breakdown from occurring at an interface of the first electrode, insulator and vacuum; anda second conductive layer located between the second metal electrode and the insulator opposite the first conductive layer, wherein the second conductive layer prevents triple junction breakdown from occurring at an interface of the second electrode, insulator and vacuum. 9. The apparatus according to claim 8, wherein the first and second conductive layers comprise metal particles doped into the insulator. 10. The apparatus according to claim 8, wherein the first and second conductive layers are deposited on the insulator. 11. The apparatus according to claim 8, wherein the first and second conductive layers are bonded onto the insulator. 12. The apparatus according to claim 11, wherein the first and second conductive layers are glued onto the insulator. 13. The apparatus according to claim 8, wherein the first and second conductive layers are joined to the insulator at an atom level without formation of a microscopic gap. 14. The apparatus according to claim 8, further comprises a first O-ring and a second O-ring, wherein the first O-ring is sandwiched between the first conductive layer and the first metal electrode and the second O-ring is sandwiched between the second conductive layer and the second metal electrode. 15. A method for preventing triple junction instability in an ion implanter, comprising:providing a first metal electrode;providing a second metal electrode;disposing an insulator between the first metal electrode and the second metal electrode, wherein the insulator has at least one surface between the first metal electrode and the second metal electrode that is exposed to a vacuum that transports an ion beam generated by the ion implanter;providing a first conductive layer located between the first metal electrode and the insulator, wherein the first conductive layer prevents triple junction breakdown from occurring at an interface of the first electrode, insulator and vacuum; andproviding a second conductive layer located between the second metal electrode and the insulator opposite the first conductive layer, wherein the second conductive layer prevents triple junction breakdown from occurring at an interface of the second electrode, insulator and vacuum. 16. The method according to claim 15, wherein the providing of the first and second conductive layers comprises doping metal particles into the insulator. 17. The method according to claim 15, wherein the providing of the first and second conductive layers comprises depositing the first and second conductive layers on the insulator. 18. The method according to claim 15, wherein the providing of the first and second conductive layers comprises bonding the first and second conductive layers onto the insulator. 19. The method according to claim 18, wherein the bonding comprises gluing the first and second conductive layers onto the insulator. 20. The method according to claim 15, wherein the providing of the first and second conductive layers comprises joining the first and second conductive layers to the insulator at an atom level without formation of a microscopic gap. 21. The method according to claim 15, further comprising providing a first O-ring and a second O-ring, wherein the first O-ring is sandwiched between the first conductive layer and the first metal electrode and the second O-ring is sandwiched between the second conductive layer and the second metal electrode. |
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054611857 | description | DETAILED DESCRIPTION OF THE INVENTION As shown in FIG. 1, a suitable glass forming material is heated to the melting point of the material and mixed with a sacrificial oxide in zone 1, forming a molten glass mixture containing at least substantially totally dissolved sacrificial metal oxide. Molten glass mixture is forwarded to zone 2 where radioactive waste, such as spent nuclear reactor fuel rods and their metal cladding, is added to the molten mixture. It is an important feature of the invention that the sacrificial oxide is present in the molten glass bath when the feed of radioactive waste is introduced, so that the sacrificial oxide may readily oxidize the metal of the waste and aid in dissolution of metal oxides. The amount of sacrificial metal oxide provided in the molten glass mixture is selected so that the ratio of sacrificial metal oxide to the elemental metal and any carbon of the waste material is sufficient to insure that any metal from the waste or its cladding will be in the oxide state or converted thereto and thus soluble in or dissolved into the molten glass. The sacrificial oxide (MO) reacts with the metal Me, e.g., plutonium, of the waste and/or cladding (and with any oxidizable carbon present) according to the following general reaction: EQU MO+Me.fwdarw.M+MeO (1) Reaction mixture may then be forwarded to zone 3, where the metal M is separated from the sacrificial oxide. If the sacrificial oxide chosen comprises a metal M having a relatively low elemental melting point, the separation may be achieved simply, such as by density or decantation. If the metal M has a high melting point, decantation or filtration may be used. Molten glass mixture containing the converted radioactive waste is forwarded to zone 4 where it is transferred to containers and allowed to solidify. The metal M, separated in zone 3, is forwarded to zone 5 where it is reoxidized and returned to use in zone 1. In the illustration of FIG. 2, a ceramic lined (e.g., fired alumina) furnace 10 contains a molten bath or mass of glass (designated generally as A) generated and maintained by a heat source (induction heater) 11. As alternate construction, furnace 10 may employ a cold wall, wherein a coating of solidified glass from the bath lines the cooled inner wall surface and protects the furnace, or both a ceramic liner and cold wall construction may be utilized. The volume of glass, which may, for example, comprise or consist of lead borosilicate glass, is maintained in furnace 10 by addition of glass frit via tube 12. The optimum formula for dissolution, in the case of this glass, will range from about 1 mole of PbO per mole of B.sub.2 O.sub.3 to about 3 moles of PbO per mole of B.sub.2 O.sub.3. Sacrificial metal oxide, in this case PbO, is supplied to the bath via tube 13. Spent fuel rods (uranium oxide clad with zirconium) are fed to the melted glass bath through an entry 14. Unmelted frit (and perhaps waste) forms a crust on the melt. The stoichiometric ratio of PbO to the elemental metal of the fuel cladding is maintained at about 2. The PbO reacts with the zirconium according to the reaction: EQU 2PbO+Zr.fwdarw.2Pb+ZrO.sub.2 (2) The uranium oxide and zirconium oxide are soluble in the glass bath, while the lead, molten at the temperature of the bath (about 1000.degree. C.) forms a separate lower layer in the furnace 10. The separate layer serves several functions in the bath. Normally, the glass and all expected wastes have lower densities than lead; waste heavier than the glass will float at the glass-lead interface while being dissolved. This feature serves to protect the bottom liner from physical and chemical attack by the unprocessed waste or-the sacrificial oxide. The lead layer also provides a thermal and corrosion barrier between glass and liner. The molten lead is removed from the furnace through outlet 15 and transferred by line 16 to oxidation zone 17. Oxygen is blown through the lead, forming solid lead oxide, which is easily separated from the surface of the molten lead, as shown. The separated solid lead oxide is then sent via line 18 for use in furnace 10 through entry 13. Concurrently, molten lead borosilicate glass containing converted radioactive waste values is removed through line 19 to canister 20 and allowed to cool and solidify. In a further embodiment, not shown, one or more additional glass baths, connected serially, may be provided. In such case, the first bath serves as a rough oxidizing and dissolution zone. Periodically or continuously, molten glass containing dissolved and/or oxidized waste is removed from a suit, able location in the first glass bath and sent to a second bath (optionally smaller) where waste, glass, glass precursors, PbO, or B.sub.2 O.sub.3 may be added as desired to produce a glass product having predetermined desired composition or characteristics. A second bath thus serves to optimize or fine tune the concentrations of the components of the glass-waste composite. Elemental metal from the superficial oxide may or may not be separated from the second bath. If separated therefrom, the metal may be sent to the first bath or the oxidation zone. Additional serial glass baths may be employed as desired or needed to produce good glass. In FIG. 3, a sacrificial oxide, such as Fe.sub.2 O.sub.3, is selected and mixed with a suitable glass, such as sodium borosilicate glass, and the mixture is heated to fusion in zone 50, the ferric oxide dissolving in the molten glass. The molten glass containing the ferric oxide is passed to zone 51 where the waste material is added, with additional heat being supplied. The ferric oxide is provided in an amount sufficient to allow oxidation of all the elemental metal of the waste material, including any assembly or cladding, while insuring that little, if any, of the ferric oxide is converted all the way to iron. Stated differently, enough Fe.sub.2 O.sub.3 is employed so that the iron in the mixture is not totally reduced but remains as FeO. Since the FeO is soluble in the molten borosilicate glass, separation of the iron values is not necessary, and the molten glass mixture containing metal oxides derived from waste and/or cladding values and FeO is transferred in zone 52 to suitable containers. In any embodiment herein where CO.sub.2 is generated from oxidation of carbon, it may simply be removed as off-gas, as illustrated in FIG. 2. As mentioned, the invention has particular application to the treatment of weapons grade plutonium. Preferably, the plutonium is fed, along with other waste material, e.g., spent light-water reactor fuel, in the procedure of FIG. 2. For illustration only, the molar ratio of spent fuel to plutonium should be at least about 6, preferably from about 10 to about 100, e.g., 20 to 1, and the final glass concentration of plutonium as oxide in the glass composite may be from about 0.1 percent to about 5.0 percent by weight. Addition of spent fuel with weapons grade plutonium serves to isotopically dilute the weapons grade plutonium with reactor grade plutonium, combines the weapons grade plutonium with highly radioactive materials, and produces a final glass material having fewer safety, security, and proliferation concerns. |
claims | 1. A method of forming a fuel rod for a nuclear reactor, the method comprising:disposing a powder comprising particles of a fuel material on a substrate;exposing the powder to energy from an energy source to form a first layer of a nuclear fuel comprising inter-granular bonds between adjacent particles of the fuel material in the first layer;disposing additional powder comprising particles of the fuel material over the first layer of the nuclear fuel; andexposing the additional powder to energy from the energy source to form a second layer of the nuclear fuel comprising inter-granular bonds between adjacent particles of the additional powder and between particles of the additional powder and the first layer of the nuclear fuel to form the nuclear fuel to have a void fraction greater than about 0.20. 2. The method of claim 1, further comprising:disposing a powder comprising particles of a cladding material around the powder comprising particles of the fuel material; andexposing the particles of the cladding material to energy from the energy source to form cladding around the nuclear fuel. 3. The method of claim 2, further comprising exposing the cladding and the nuclear fuel to annealing conditions to densify the cladding without densifying the nuclear fuel. 4. The method of claim 1, wherein disposing a powder comprising particles of a fuel material on a substrate comprises disposing a powder comprising between 1.0 weight percent and 15.0 weight percent uranium and between 85.0 weight percent and 99.0 weight percent zirconium on the substrate. 5. The method of claim 1, wherein disposing additional powder comprising particles of the fuel material over the first layer of the nuclear fuel comprises disposing additional powder comprising a different composition of the fuel material than the powder of the first layer of the nuclear fuel. 6. The method of claim 5, wherein exposing the additional powder to energy from the energy source to form a second layer of the nuclear fuel comprises forming the second layer to have a different composition than the first layer. 7. The method of claim 6, wherein exposing the additional powder to energy from the energy source to form a second layer of the nuclear fuel comprises forming the first layer to comprise a different amount of enriched nuclear fuel than the second layer. 8. The method of claim 1, further comprising exposing the nuclear fuel to annealing conditions to densify at least a portion of the nuclear fuel. 9. The method of claim 8, wherein exposing the nuclear fuel to annealing conditions comprises exposing the nuclear fuel to at least one of a spark plasma sintering process or a hot isostatic pressing process. 10. The method of claim 1, wherein disposing a powder comprising particles of a fuel material on a substrate comprises disposing a powder comprising particles of the fuel material and particles of a burnable poison material on the substrate. 11. The method of claim 1, further comprising disposing particles of a diffusion barrier material on the substrate, wherein exposing the powder to energy from an energy source to form a first layer of a nuclear fuel comprises exposing the particles of the diffusion barrier material to energy from the energy source to form a diffusion barrier around the fuel material. 12. The method of claim 1, wherein disposing a powder comprising particles of a fuel material on a substrate comprises disposing a powder comprising particles of the fuel material and at least one dopant on the substrate. 13. The method of claim 12, wherein disposing a powder comprising particles of the fuel material and at least one dopant on the substrate comprises disposing a powder comprising particles of the fuel material and at least one dopant selected from the group consisting of carbon nanotubes, zirconium oxide, aluminum oxide, and combinations thereof on the substrate. 14. The method of claim 1, wherein disposing a powder comprising particles of a fuel material on a substrate comprises disposing a powder comprising particles of uranium and zirconium on the substrate. 15. The method of claim 1, wherein exposing the powder to energy from an energy source to form a first layer of a nuclear fuel comprises forming the first layer of the nuclear fuel to define an aperture through a central portion of the first layer of the nuclear fuel. 16. The method of claim 1, wherein disposing a powder comprising particles of a fuel material on a substrate comprises disposing a powder comprising about 10 atomic percent uranium and about 90 atomic percent of at least one of zirconium and tungsten on the substrate. 17. The method of claim 1, wherein exposing the powder to energy from an energy source to form a first layer of a nuclear fuel comprises forming the first layer of the nuclear fuel to exhibit a composition varying with a radial distance from a center of the first layer. 18. The method of claim 1, further comprising disposing a powder comprising particles of a cladding material around the powder comprising particles of the fuel material, an enrichment of the nuclear fuel greater proximate the cladding material than at portions distal from the cladding material. 19. The method of claim 1, wherein exposing the powder to energy from an energy source to form a first layer of a nuclear fuel comprises exposing the powder to energy to form a first layer of an article comprising a bundle of fuel rods. 20. A method of forming a portion of a nuclear reactor, the method comprising:exposing powder comprising at least a nuclear fuel material and a cladding material to energy to form a first layer of an article;disposing additional powder over the first layer of the article to form a layer of the additional powder over the first layer of the article; andexposing the additional powder to energy to form a second layer of the article, wherein the second layer of the article has a same cross-sectional shape as the first layer of the article, the same cross-sectional shape being rotationally offset from the first layer of the article. 21. The method of claim 20, wherein exposing a powder comprising at least a nuclear fuel material and a cladding material to energy to form a first layer of an article comprises forming the first layer of the article to have a theoretical density less than about 90% theoretical density. 22. The method of claim 20, wherein exposing powder comprising at least a nuclear fuel material and a cladding material to energy to form a first layer of an article comprises exposing a powder comprising uranium and at least one of a metal oxide and carbon nanotubes to the energy. 23. The method of claim 20, wherein exposing powder comprising at least a nuclear fuel material and a cladding material to energy to form a first layer of an article comprises exposing a powder comprising at least a nuclear fuel material and a cladding material to energy to form a first layer of an article comprising a bundle of fuel rods. |
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058964290 | description | DESCRIPTION OF PREFERRED EMBODIMENTS A. Introduction A pair of methods is used to safely and reliably evaluate the thickness and integrity of the liners of the hearth of a blast furnace. The first method involves the transport of gamma rays (specifically, bremsstrahlung radiation) directed into the hearth-wall liner. The second involves the die away of thermal neutrons directed into the hearth-wall liner. Either method can be used independently to evaluate the condition of a hearth-wall liner. However, the two methods are used to greatest advantage when they are used together because of the complementary information that each provides. The term, "gamma rays," may be used interchangeably herein with "photons" and "electromagnetic radiation." However, the term, "gamma rays," most correctly refers to radiation produced by nuclear processes. An embodiment of the apparatus for performing the disclosed methods is illustrated in FIG. 2. The radiation source 32 is a small compact 6 MeV electron linear accelerator (LINAC) directed toward the wall 13 opposite the molten iron 36. The hearth-wall liner 14 typically comprises carbon and is surrounded along its exterior by a one-to-two-inch-thick steel shell 11. The radiation source 32 is one meter long with a weight of approximately 40 kg and is designed for field use. It produces gamma rays of up to 6 MeV in energy. Moreover, a beryllium target 33 may be moved into the path of the gamma rays, as shown by the broken-line image, to produce neutrons for the decay-time measurements. Alternatively, the radiation source 32 may be supplemented or replaced by a small sealed pulsed neutron generator. The source 32 is surrounded by gamma-ray detectors 34 positioned to detect gamma radiation emitted from the wall 13 as a result of the gamma rays and neutrons directed into the wall 13. Any suitable gamma-ray detector may be used. For example, the detectors may be semiconductor detectors, such as germanium or cadmium zinc telluride. Alternatively, the detectors may be scintillators, such as sodium iodide (thallium-doped), barium fluoride, bismuth germanate, one of the rare earth oxyorthosilicates (cerium-doped) or a rare earth aluminate (cerium-doped) porovskite, for example, with an associated photosensitive device such as a photomultiplier tube (PMT) or avalanche photo diode (APD). As another alternative, the detectors may be gas detectors, such as high pressure xenon. The radiation/neutron source 32 and its associated detectors 34 are portable and may be moved vertically up the wall 13 of the furnace as well as horizontally to traverse the entire area to be scanned. The cylindrical side wall of the crucible-shaped hearth 12 of a blast furnace 10, illustrated in FIG. 1, is particularly well suited for evaluation by these methods. B. Gamma-Ray Transport In the first of the two methods for evaluating the hearth-wall liner, high-energy bremsstrahlung radiation 38 is directed at the hearth sidewall 13 from outside the furnace 10, and the radiation emitted back out of the wall 13 is measured and analyzed. As noted previously, the bremsstrahlung radiation is produced by a LINAC and comprises gamma ray photons of at least 1.02 MeV, and, preferably, between about 3 and 8 MeV, emitted by accelerated electrons after colliding with the nuclei of atoms in a target within the LINAC. The gamma rays penetrate the hearth 12, allowing testing to be performed without draining the furnace, without placing probes in the furnace, without breaching the integrity of the outer steel shell 11 and without suspending equipment down into the furnace from above. Using gamma rays as described herein, the test can be performed without any disruption of furnace operation. As the photons collide with the atoms that comprise the hearth-wall liner 14, and, perhaps also, the iron bath 36, as shown in FIG. 3, each photon 40 may interact with an atom to produce at least one of the following three results: pair production, Compton scattering or photoelectric absorption. First, if the energy of the photon is above 1.02 MeV, the photon may be completely absorbed by the nucleus resulting in pair production. Pair production occurs when the energy of the photon is converted into an electron and a positron, each with a rest-mass energy of 511 keV and a kinetic energy that depends on the energy of the incident gamma ray. A positron is a particle with the mass of an electron but with a positive charge. As the positron and electron move away from their point of origin, each loses energy by ionization. When the positron energy drops low enough, the positron combines with an electron, and both are annihilated to produce a pair of 511 keV photons radiated in opposite directions. Pair production is illustrated by the photon-interaction representation, A, in FIG. 3. Annihilation is illustrated by the representation, B, along the same path. The probability of pair production is directly proportional to the square of the atomic number of the medium. Though the exact energy dependence of pair production is complex, a reasonable approximation of the probability of pair production per unit track length, .SIGMA..sub.pair, at higher energies is given by: EQU .SIGMA..sub.pair =K.sub.pair .rho..sub.nuc Z.sup.2 lnE.sub..gamma., where K.sub.pair is a constant, .rho..sub.nuc is the nuclear density, Z is the atomic number and E.sub..gamma. is the energy of the incident photon. Second, the photon may collide with an electron and transfer all of its energy to the electron in the form of kinetic energy. In such a case, the electron is ejected from its atom, and the photon disappears in what is known as the photoelectric effect. An example of the photoelectric effect is designated by the photon path representation, C, in FIG. 3. The ejection of the electron leaves a vacancy in one of the atom's orbitals which will typically be filled by an electron in an outer orbital of higher energy level. As the outer electron changes orbitals to fill the vacancy, energy is released in a photon which can be easily recognized because its energy will be that of the difference in energy between the respective orbitals of the atom. The ejected electron has an energy which is the difference between the energy of the incident gamma ray and the binding energy of the electron in its atomic shell. Like pair production, photoelectric absorption increases with increasing atomic number. The absorption probability, .SIGMA..sub.a, per track length of photons is approximately given by: EQU .SIGMA..sub.a =K.sub.a Z.sup.3.6 .SIGMA..sub.e /E.sub..gamma..sup.3.15, where .rho..sub.e is the electron density. Third, the photon may collide with an electron and transfer some of its energy to the electron, ejecting the electron from its orbit. Meanwhile, the photon deflects and continues through the medium, albeit at a lower energy. This process occurs mainly with electrons in the outer orbital of an atom and is known as Compton scattering. Compton scattering dominates at energies above those where photoelectric absorption is dominant yet below those where pair production becomes important. Compton scattering is illustrated by the photon-interaction representation, D, in FIG. 3. The difference between Compton scattering and photoelectric absorption is that the photon loses only part of its energy when it undergoes Compton scattering. The likelihood of scatter per unit track length, .SIGMA..sub.s, is a function only of electron density and photon energy and is given by: EQU .SIGMA..sub.s =K.sub.s .rho..sub.e /E.sub..gamma.. The energy of the photon after it is scattered is inversely related to the angle by which it is scattered. After the photon scatters, the reduced-energy photon will likely interact with another atom in one of the three ways discussed. Because carbon, of which the hearth-wall liner 13 of the hearth 12 is comprised, has a relatively low atomic number, 6, pair production and photoelectric absorption rarely occur in carbon. Instead, a photon traveling through carbon is much more likely to undergo Compton scattering. In contrast, the iron of the bath has a much higher atomic number, 26. As a result, the probability of pair production in iron is at least a factor of ten larger than in carbon, especially where the photon energy is greater than 3-4 MeV. Thus, we expect pair production at the surface of the iron and subsequent production of photons with the energy characteristic of this process, 511 keV. The equations, provided above, are utilized to establish a strategy for distinguishing carbon from iron. In a thick piece of carbon, the dominant form of photon interaction will be Compton scattering, whereas in iron, pair production will be much larger. These differences are even more pronounced when one accounts for the difference in density between iron and carbon in establishing absolute rates. Accordingly, the radiation emitted from the wall will vary according to the thickness of carbon that the photon must pass through before reaching the iron. The signature of the photons detected at or near the surface of the wall can then be matched to radiation patterns characteristic of walls having hearth-wall liners of varying thicknesses to closely estimate the actual thickness of the hearth-wall liner measured. The gamma-ray source and detectors are then moved to different locations on the sidewall for repeated measurements to better evaluate the entirety of the sidewall. The exact details of the multiple scatter and its dependence on carbon thickness can best be predicted using simulation, as for example by the MONTE CARLO N-PARTICLE CODE.TM. (MCNP). MCNP is designed to simulate the transport of neutrons, photons and/or electrons through a medium or plurality of media. When a user enters the appropriate description (geometry and materials) of the media and the source, the code sequentially simulates each of the individual probabilistic events that comprise the transport of the particle through the medium. When simulating the interaction or sequence of interactions between a photon and the medium, the code is used to statistically sample a probability distribution of each of the possible forms of photon/atomic interaction, described above, using a random number generator. As the simulated particle proceeds from a source through each simulated interaction, the code is used to track the decrease in energy and the change in direction that each interaction produces until the eventual death of the particle in some terminal category, such as absorption, pair production, or escape. As the histories of more simulated particles are collected, the release of the various types of detectable radiation is tallied to produce a model estimating actual behavior in a system having the specified characteristics. By modifying the characteristics of a carbon hearth-wall liner that are input to the MCNP--particularly, the thickness of the carbon--a range of results can be obtained. Since the area tested will normally be surrounded by a steel shell, the simulation should be configured to include the characteristics of this shell so that its influence can be accounted for. After confirming the accuracy of these results by comparing the MCNP values with actual test data, the MCNP values can be matched to field measurements to provide a reliable estimate of carbon thickness. Alternatively, algorithms can be empirically developed that relate changes in the spectral response to changes in the carbon thickness. Using MCNP, simulations of gamma ray transport through a hearth wall with a 5-cm-thick steel shell and a carbon liner having a thickness of either 90 or 100 cm were conducted. The simulated energy spectra produced at a point 20 cm from the point on the wall where radiation is incident are shown in FIGS. 4 and 5. The peak of the simulated energy spectrum, between 200 and 300 keV is approximately 5% greater in the simulation for the 100-cm-thick carbon hearth-wall liner, shown in FIG. 5, than the peak for the 90-cm-thick carbon hearth-wall liner. These results validate the premise that the methods of this invention can be used to differentiate between hearth-wall liners of different thicknesses. As an alternative to MCNP, a similar simulation program, known as COG.TM., developed at Lawrence Livermore National Laboratory may be used, as may other Monte Carlo programs. C. Neutron Decay Time A second procedure for inspecting the carbon hearth-wall liner of a blast furnace involves directing neutrons into the hearth-wall liner and measuring the time between emission and absorption of the neutron. The neutron scatters through the medium of the wall, and sometimes through the molten iron, losing energy until it reaches thermal energies. At thermal energies, the neutron diffuses through the medium until captured by a nucleus, thereby producing a gamma ray that can the furnace outside the furnace. The basis for this approach is the fixed rate at which thermal neutrons will be absorbed in a given medium. The rate at which the neutrons are absorbed depends on the macroscopic thermal-neutron-absorption cross section, .SIGMA., of the medium. The macroscopic thermal-neutron-absorption cross section is the average microscopic thermal-neutron-absorption cross section per unit volume of the medium. The rate at which thermal neutrons are absorbed in a medium is given by: EQU dN/dt=N.sub.0 e.sup.-.SIGMA.vt, where N.sub.0 is the number of neutrons at time t=0 and v is the velocity of the neutrons. The major advantage of using this technique in this particular context is that the neutron absorption cross sections for carbon and iron are dramatically different. The microscopic thermal neutron absorption cross section for iron is 2.56 barns (10.sup.-24 cm.sup.2), while that for carbon is 3.4 millibarns. Thus, the sensitivity of a measurement of the thermal neutron decay time for iron in a predominantly carbon environment is about a factor of a thousand. As a result, the detection of a 5 percent change in the thermal decay time should allow the operator to detect an infiltration of iron into carbon at about 50 parts per million. The measurement is performed by producing a short burst of high-energy neutrons. The burst in this situation should be about one millisecond in duration, though narrower burst widths are quite acceptable. The high-energy neutrons can be provided using the same compact electron accelerator used to produce the bremsstrahlung radiation for the gamma-ray transport measurement. The neutrons are supplied by placing a beryllium secondary target in the path of the gamma rays. Alternatively, the neutrons can be supplied by a small sealed neutron generator, such as those used in the oil well logging industry. The high-energy neutrons scatter in the furnace material, gradually losing energy until they approach thermal energies. While at thermal energies, the neutrons diffuse through the furnace material until they are captured by nuclei. Each capture event is accompanied by the essentially instantaneous production of one or more gamma rays that are detected by a gamma-ray detector outside the furnace, where the gamma-ray detector is of a type mentioned previously. The detection of gamma rays is recorded as a function of time relative to the end of the burst of neutrons. Following a sufficient interim, additional neutron bursts are provided to produce more gamma rays, which are detected and recorded to build the necessary level of statistical accuracy for the lifetime determination. The number of gamma rays detected as a function of time is fit by an appropriate mathematical function to determine the decay time of neutrons directed into the wall. Moreover, the neutron source and detectors are moved to different locations on the sidewall, where the measurements are repeated and a better evaluation of the entirety of the sidewall is obtained. The recorded measurements are then fed to a personal computer 35 to determine the amount of infiltrated iron by correlating the neutron decay times (or the time dependence of the gamma-ray spectrum) with known iron infiltration parameters. The proper correlation can be established through mathematical analysis, actual testing, and/or Monte Carlo simulation, using codes such as MCNP. Once the computer 35 has performed this correlation, a value for the estimated thickness of the hearth-wall liner is transmitted to a display 37. Neutron decay measurements have been used to probe earth formation properties behind steel casings. However, they have not been applied in an industrial application, as considered here. The expected depth sensitivity in this environment is expected to be about 70-90 cm which is sufficiently deep to monitor the region from the outer surface of the furnace to a depth near the iron-carbon interface. D. Equivalents While this invention has been particularly shown and described with references to preferred embodiments thereof, it will be understood by those skilled in the art that various changes in form and details may be made therein without departing from the scope of the invention as defined by the appended claims. For example, and without limitation, infiltration of other molten metals, besides iron may be analyzed in a similar fashion utilizing the apparatus and methods described herein. |
abstract | A collimator for a radiotherapy apparatus, comprising a block of radiation-attenuating material having a front face forming the leading edge of the block and at least one main rear face defining the trailing edge of the block, in which the or each rear face is substantially planar in the direction of the depth of the block and non-parallel to the front face. The collimator may form part of a radiotherapy apparatus, and methods of operation of such apparatus are described. |
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description | The fuel element for high temperature pebble bed nuclear reactors (hereinafter referred to as HTR) is a graphite sphere of 60 mm diameter that is produced by molding of special A 3-type graphite. The fuel element for HTR consists of a fuel-comprising core with 50 mm diameter that is surrounded by a fuel-free shell of 5 mm thickness. The core of the fuel element sphere (FE-sphere) is seamlessly bonded to the shell and thus forms a unit with it. The fuel is homogeneously distributed in the form of coated fuel particles within the sphere core. The coated particles are spheres of about 0.5 mm diameter (fuel kernels) preferably consisting of uranium oxide. These kernels are coated multiple times with layers of pyrolytic carbon and silicon carbide in order to retain the fission products which are generated during reactor operation. The spherical fuel kernels are preferably obtained by the gel supported precipitation method. This method includes forming drops of a solution of uranyl nitrate with additives of polyvinyl alcohol and tetrahydrofurfuryl alcohol by vibrating nozzles. The solution is then solidified to spherical fuel kernels of ammonium diuranate (ADU) by using NH3 as well as NH4OH. After washing, drying, reducing and sintering, UO2-fuel kernels of high density having desired diameters are obtained. The coating of the fuel kernels with pyrolytic carbon and silicon carbide is usually performed in fluidized bed units. These units (furnaces) consist of a vertically standing graphite tube with a conical bottom which is heated from outside by means of a graphite resistance heater. Several nozzles are placed to the end at the top point of the core. They provide the carrier gas argon or hydrogen as well as coating gases into the unit for fluidized bed operation. The pyrolytic carbon layers are deposited by thermal decomposition of ethine or a mixture of ethine and propine from the gaseous phase at temperatures between 1000° C. and 1400° C. In case of coating with silicon carbide, methyl trichlorosilane preferably serves as the coating gas. In this case the deposition temperature is slightly higher and amounts to 1500° C. Depending on the coating conditions, several layers of different densities and structures with different physical and mechanical properties are obtained. This method is inter alia published in SM-111/15, Symposium “Advanced HighTemperature Gas-Cooled Reactors” in Jülich on 21. to 25. October 1968 and in “Recent Development in the Manufacture of Spherical Fuel Elements for High Temperature Reactors” (Hackstein, K. G., Hrovat, M., Spener, G., Ehlers, K.) and KFA Report, Jülich, 687-RW (August 1970), “Entwicklung von beschichteten Brennstoffteilchen” (H. Nickel) as well as in the German patent DE 102 04 166 and the German publication DE 101 44 352 A1. The HTR fuel element spheres have to meet many requirements: high geometrical density of the graphite matrix good mechanical strength properties low young modulus low coefficient of thermal extension good heat conductivity and high stability upon irradiation with fast neutrons. In order to achieve these properties molded bodies manufactured from carbon have to be subjected to a graphitizing process at temperatures between 2700° C. and 3000° C. As the molded fuel element spheres comprise coated fuel particles within the sphere core after molding, a graphitizing process at temperatures above 2700° C. cannot be done. The decisive reasons for that are: Already at temperatures above 2100° C. the uranium diffuses from the fuel kernels into the coating layers of the particles and further into the graphite matrix of the fuel element sphere. The uranium that diffused into the porous graphite matrix of the fuel element sphere outside the coating would have led to impermissibly high contamination of the cooling gas with the released fission products during reactor operation. Diffusion of uranium in graphite is described in Journal of Nuclear Materials, 19 (1966), pages 53-58 by Hrovat, M. and Spener, G. Furthermore, the pyrolytic carbon layers change their structure at temperatures above 2100° C. Thereby anisotropy of the crystallographic orientation of the pyrolytic carbon increases significantly. Consequently, there is the risk of the coated particles losing their mechanical integrity in the reactor very early. This would lead to the risk of radioactive fission products being spontaneously released. The results are described in KFA-Report “Jül-868-RW” (June 1972), “Über die Änderung der Anisotropie der kristallographischen Orientierung in Pyrokohlenstoffhüllschichten durch Glühung und Neutronenbestrahlung” by Koizlik, K. The pertinent literature teaches that graphite only keeps its dimensional stability and mechanical integrity upon radiation with fast neutrons in temperatures above 1000° C., if it is highly crystalline and isotropic. The irradiation process and the corresponding results are inter alia described in: GA-Report (March 1970), “Irradiation Behaviour of Nuclear Graphites at Elevated Temperatures”, by Engle, G. B. and PNWL-1056 Report (1969), Pacific Northwest Laboratory Richland/Wash., by Helm, J. W. In order to ensure dimensional stability and mechanical integrity of the molded fuel element spheres during the whole residence time in the reactor, in spite the heat treatment being limited to only about 2000° C., a special graphite has been developed. The special graphite is referred to in technical literature as A 3-graphite matrix. A 3-graphite matrix is based on natural graphite. Natural graphite shows extremely high crystallinity. Its primary particles are, however, lamellar with hexagonal crystalline order (“Syngonie”) and are thus highly anisotropic. In order to achieve the required isotropy of the physical properties of the matrix the fuel element spheres are molded in a rubber die, preferably a silicon rubber. The cylindrical rubber die is setup of several parts having an ellipsoid cavity in the center for reception of the mixture of molding powder and fuel. This cavity has such measures that already at a pressure of above 5 MPa a sphere is formed. The rubber die is inserted into a steel die of the compactor and is compressed between lower and upper punches. For manufacturing the fuel element spheres a mixture of graphite molding powder and overcoated particles are pre-molded into manageable sphere cores. Then the pre-molded sphere core is embedded into graphite molding powder in a second rubber die and is molded at elevated pressure to yield a permeable, air permeable sphere. This remolded sphere is then molded in a third rubber die in vacuum to the desired density. For carbonization of the binder the fuel element spheres are heated in inert gas atmosphere to 800° C. within 18 hours and are finally annealed in vacuum at about 2000° C. The A 3-graphite matrix consists of 72.7% by weight of natural graphite, 18.2% by weight of petroleum coke (graphitized in powder form at 3000° C.) and 9.1% by weight of binder coke. The method is described in German patent and publication pamphlets DE 198 37 989 C2 and DE 102 53 205 A1. In order to not only provide fuel element spheres that are isotropic but also show hardly any property gradient, the fuel element spheres are molded in a third molding step at a high pressure of 300 MPa to a density of 1.92 g/cm3, which is approximately 99% of the theoretical density. Upon pressure release the density decreases to a value of 1.8 g/cm3, decreases further upon heat treatment and reaches a minimum value of 1.6 g/cm3 at 280° C. At this temperature the binder resin starts to carbonize while producing gaseous crack products. Adjustment of the required porosity is done by adding a portion of electro graphite powder. Thereby a nearly pressureless degassing of the matrix is achieved, thus avoiding cracks in the matrix. While the carbonization of the resin proceeds, the graphite matrix begins to shrink and reaches a relatively high final density of 1.72 g/cm3 at around 850° C. Weight loss of the sphere matrix by carbonizing of the resin amounts to about 9% by weight. Optimization of the A 3-graphite matrix is described in KFA-Report, Jül.-969-RW, June 1973, “Über die Entwicklung eines Matrixmaterials zur Herstellung gepresster Brennelemente für Hochtemperaturreaktoren” by Hrovat, M., Nickel, H. and Koizlik, K. During the seventies and eighties of the twentieth century more than a million of molded A 3 fuel element spheres have been used in the pebble bed reactor AVR in Jülich and the thorium high temperature reactor (THTR) in Schmehausen/Uentrop. The fuel element spheres have proven themselves in continuous operations and have shown immaculate behaviour. The nuclear reactors of later generations additionally make further demands on HTR fuel elements. The fuel element spheres have to stay intact at full reactor power and may not release impermissible fission products in case a hypothetical accident happens and, for example, a complete cooling breakdown and/or uncontrolled break-in of air, water or water vapor into the reactor core happens. In order to provide for these prerequisites A 3-fuel element spheres have to have improved corrosion resistance against oxygen or water vapor. In order to determine corrosion resistance a standard testing procedure is conducted. In this standard testing procedure the fuel element spheres are heated to 1000° C. in an inert gas atmosphere comprising water vapor, and resulting weight loss is determined. The reaction gas is a mixture of argon with 1% volume of water vapor. This mixture is manufactured in a moisturizing container which is filled with water. During this manufacturing procedure the argon gas that bubbles through the water is saturated with water vapor. The volumetric flow rate of the reaction gas is 150 l/hr and is selected such that only about 20% of available oxygen reacts with the graphite matrix of the sphere in the given test conditions. The corrosion rate is the burn-up of the graphite per hour in milligrams per square centimeter of sphere surface. The value is determined at 1000° C. using A 3 fuel element spheres, the value is in the range of from 1 to 1.25 mg/cm2 per hour. The reference value for nuclear-pure ATJ reactor graphite of Union Carbide Corporation (UCC), which has been graphitized at 3000° C., is significantly lower at 0.7 mg/cm2 per hour. Long term corrosion tests revealed that within the A 3 fuel element spheres preferably the binder coke which has formed from the phenol formaldehyde resin reacted with the water vapor and thus led to selective burn-up of the matrix. In contrast to natural graphite and graphitized petroleum coke the binder coke showed a significantly higher chemical affinity and consequently a higher reaction rate of the oxidation with water vapor. In order to prove the selective oxidation of the binder coke the oxidized fuel element spheres which had been subjected to the corrosion test procedure where subsequently subjected to an attrition test. For testing the attrition the fuel element spheres were moved into a rotating barrel which rotated at 55 rotations per minute. A chamfered threshold of 2 mm height on the inner surface of the barrel provided for the fuel element spheres permanently being in motion and not sliding on the inner surface of the barrel. Holes on the bottom and lid of the barrel provided for an unresisted exit of attrited graphite matrix. The attrited graphite matrix was analyzed for crystallinity by X-Ray microstructure analysis. The crystallite size was 90 nm and hence constitutes a very high value and can only be attributed to the graphite components of natural graphite and graphitized petroleum coke. The corresponding Lc value of the binder coke is approximately one order of magnitude lower and could not be determined. The absence of binder coke in the attrited graphite matrix is a proof for selective burning up of coke during the corrosion test procedure (oxidation). Selective burning up of binder coke is thus the main reason for loss of strength in the graphite matrix in connection with shaving of the surface of the fuel element spheres. Procedure for increasing corrosion resistance in graphite molded bodies are known from DE 41 27 693 A1, DE 27 18 143 and DE 12 69 559. In these documents the molded bodies are improved in terms of corrosion resistance by applying protective layers comprising SiC and/or ZrC after completion. Such subsequent coating is not done for the fuel element spheres made from A 3 graphite and, consequently, there is no coating that could hamper selective burning-up of the binder coke, which decisively causes corrosion of the sphere shaped fuel elements. Furthermore, subsequent compression processes are labour-intensive and expensive. It is the object of the present invention to provide a new concept for fuel element spheres and their production, meeting the prerequisites of high temperature pebble bed nuclear reactors of later generations. This object is solved by the subject-matter of the patent claims. The object is especially solved by the fuel element spheres comprising a fuel-free shell of silicon carbide (SiC) and/or zirconium carbide (ZrC) as well as natural graphite and graphitized petroleum coke, the shell having an average nominal thickness of at least 1 mm, preferably at least 2 mm and most preferably at least 3 mm. It is further preferred that the average nominal thickness is in a range of from 1 to 5 mm, even further preferred in a range of from 2 to 5 mm and most preferred in a range of 2 to 4 mm, wherein 3 mm is a possible embodiment. Determination of the average nominal thickness of the fuel-free layer is done by means which are known to the person having skill in the art. The values given above include a range of tolerance (due to accuracy of measurement) of +/−0.5 mm. The proportion of silicon carbide in the fuel-free shell is in the range of from 6 to 14% by weight, further preferred in the range of from 8 to 12% by weight, even further preferred in the range of from 9 to 11% by weight and most preferred in the range of from 9 to 10% by weight, wherein 10% by weight is a possible embodiment. The proportion of zirconium carbide in the fuel-free shell is in the range of from 10 to 30% by weight, further preferred in the range of from 15 to 30% by weight, even further preferred in the range of from 19 to 25% by weight and most preferred in the range of from 20 to 23% by weight, wherein 22.3% by weight is a possible embodiment. The main aspect of the present invention is the use of the high chemical affinity of the binder coke. This has in the past been a disadvantage in corrosion tests. However, in the tests it has surprisingly been found that the chemical affinity of the binder coke can be used by adding a silicon and/or zirconium compound to the molding powder for the shell. Similar to oxidation with water vapor the silicon and/or zirconium compound selectively reacts with the carbon of the binder coke during annealing of the fuel element spheres in vacuum at a maximum temperature of 2000° C. Thereby almost only the proportion of binder coke of the A3 graphite matrix which is accountable for corrosion is reacted to the corrosion resistant SiC or ZrC. Both carbides SiC and ZrC are well proven reactor materials with cubic crystalline structure (“Syngonie”) and thus inherently isotropic. SiC and ZrC are characterized by high hardness, high mechanical strength and very good corrosion resistance. By application of SiC or ZrC in the production of A 3 graphite matrix such properties of the fuel element spheres like density, load at brake and especially corrosion resistance are significantly improved and the prerequisites of fuel element spheres for pebble bed reactors of later generations are met. As a consequence of improved corrosion resistance and mechanical strength properties the thickness of the fuel free shells of the fuel element spheres can be decreased. Thereby the relative volume of the fuel containing sphere core is increased and subsequently the fuel temperature is decreased. A lower fuel temperature improves retention capability for fission products of the coated particles significantly. In the production of sphere shaped fuel elements according to the present invention the same graphite molding powder is used for the shell as well as for production of the fuel-containing core. During production of the fuel elements the graphite matrix is formed from the molding powder, which graphite matrix is consequently the same for shell and core. The fuel elements according to the present invention can thus be described by the feature of similar or identical composition of the graphite matrix in the fuel-containing core and fuel-free shell. This feature is important especially for distinction from other sphere shaped fuel elements which do not comprise this identity of graphite matrix. Such fuel elements preferably comprise a shell of electro graphite and are for example described in the publication: “Fuel Development for THTR”, G. Spencer, M. Hrovat and L. Rachor, Proceedings of the Conference “Fuel cycles of the HTGR”, Brussels, June 1965. In order to describe the similarity of the graphite matrices it shall be noticed that for production of the fuel-containing core and the fuel-free shell the same graphite molding powder is used, which finally forms said graphite matrix. The term “similarity” is used with respect to the fact that silicon or zirconium compounds are added to the graphite molding powder which is used for pre-molding of the fuel-free shell, in the above-mentioned amounts. The following examples shall further describe the provision of the fuel element spheres according to the present invention and their novel conception without restricting the scope of the invention: Production of the graphite molding powder has been done in two separate homogenization batches: The molding powder for the sphere core and the molding powder for the sphere shell. For production of the molding powder for the sphere core a nuclear-pure natural graphite was pre-mixed with petroleum coke which had been graphitized at 3000° C. in a weight ratio of 4:1 in dry condition. Relative to the graphite components 20% by weight of phenol formaldehyde binder resin dissolved in methanol was added and homogenized in a kneader mixer at room temperature. The material to be kneaded was dried at 105° C. in vacuum (P<50 hPa) and afterwards broken down in a hammer mill with sieves adjusted to 1 mm. For production of the molding powder for the sphere shell all process steps except preparation of the SiO2 suspension remained unchanged. The proportion of SiO2 powder amounted to 83.4% by weight relative to the binder resin. The starting compounds had the following properties: natural graphite with labeling FP of the supplier Kropfmühl, bulk density 0.4 g/cm3, grain density 2.26 g/cm3, BET-surface 2 m2/g, crystallite size Lc=100 nm, average particle size 10 to 20 μm, ash content 200 ppm and boron equivalent from impurities of the ash <1 ppm. graphitized petroleum coke with the labeling KRB <0.1 mm of the supplier Ringsdorff, graphitizing temperature 3000° C., bulk density 0.65 g/cm3, grain density 2.2 g/cm3, BET-surface 1.2 m2/g, crystallite size Lc=60 nm, average particle size 30 to 40 μm, ash content 10 ppm and boron equivalent form impurities of the ash <1 ppm. phenol formaldehyde resin of the type Novolak with the labeling 4911 of the supplier Bakelite, condensation agent HCl, molecular weight 690, softening point 101° C., pH-value=6, acid value=7.5, free phenol 0.12% by weight, coke yield 50%, solubility in methanol 99.97% by weight, ash content 160 ppm and boron equivalent from the impurities of the ash 1 ppm. In order to increase molecular weight the resin has been subjected to steam distillation after condensation. SiO2 powder, finely ground commercial SiO2 powder with an average partide size of from 1 to 5 μm and a purity of 99.95%. The coated fuel particles with a diameter of about 0.9 mm were overcoated with a part of the graphite molding powder which had been produced for the sphere core in a rotating drum under addition of small amounts of nebulized resin solvent. This procedure was done until the particles were covered by a porous overcoating layer which was about 0.2 mm thick. The coated UO2-fuel particles had a core diameter of 0.5 mm and a density of 10.6 g/cm3. The fuel kernels were coated four times, first with a buffer layer of pyrolytic carbon (thickness 95 μm, density 1.05 g/cm3), then with a dense pyrolytic carbon layer (thickness 40 μm, density 1.90 g/cm3), afterwards with a dense SiC layer (thickness 35 μm, density 3.19 g/cm3) and finally with a dense pyrolytic carbon layer (thickness 40 μm, density 1.90 g/cm3). The coated particles, overcoated with molding powder, were dried and mixed portion after portion with further graphite molding powder in a weight ratio of 1:2.23. A portion of this mixture weighing 164 g which constitutes 29.3 g coated fuel particles, was charged into the first rubber die. This rubber die was molded in a steel die at 5 MPa. In a rubber die with ellipsoid shaped cavity of 205 cm3 and an axial ratio of 1:1.17 a manageable sphere with a diameter of about 62 mm and a density of 1.2 g/cm3 was obtained. This sphere was embedded into lose layer of graphite molding powder with SiO2 additive in a second rubber die. Within an axial ratio of 1:1.14 the volume of the ellipsoid shaped cavity was 295 cm3. After molding of the die at 15 MPa a sphere with about 68 mm diameter, a weight of 240 g and a density of 1.45 g/cm3 was obtained. This pre-molded sphere was placed in a third accurately fitting rubber die and subjected to a final molding in vacuum (P<120 hPa) at high pressure of 300 MPa. Under the pressure of 300 MPa the density of the graphite matrix was 1.94 g/cm3. With the chosen composition of the graphite molding powder this density equals the value of 99% of the theoretical density. After load release the density of the graphite matrix decreased from 1.94 g/cm3 to 1.82 g/cm3. To carbonize the binder the spheres were heated to 800° C. in a nitrogen purge for 18 hours and finally annealed in vacuum (P<10−2 hPa) at 1900° C. During this process step the binder coke which had been formed from the phenol formaldehyde resin according to the present invention reacted with SiO2 to SiC. Except for the replacement of the SiO2 powder by ZrO2 powder the further production steps of the molding powder production remain unchanged, i.e. as described in example 1. The ZrO2 proportion in the methanol resin solution was 167% by weight relating to the binder resin. The applied ZrO2 powder with the labeling TZ of the supplier Toyo Soda had an average particle size of about 1 μm and a purity of 99.99%. After heat treatment and machining of the fuel element spheres to a diameter of 60 mm the following properties were determined: geometrical density of the fuel-free sphere shell, thickness of the fuel-free shell (thickness was determined by means of X-Ray analysis), load at brake, this parameter was determined by crushing the fuel element spheres between two steel plates and measuring the load at brake, SiC or ZrC amount in the sphere shell (determination was done by chemical analysis and X-Ray micro structure analysis) and corrosion resistance (the fuel element spheres were subjected to the standard oxidation test procedure). The results are listed in the following table and compared to the values of the A 3-fuel element spheres. with SiCwith ZrCA 3-fuelfuel elementproportionproportionelementsphere/propertyin the shellin the shellspheresdensity of the fuel-1.781.911.72free shell (g/cm3)thickness of the fuel-3 ± 0.53 ± 0.55 ± 0.5free shell (mm)load at break parallel293124to direction of moldingload at break vertical272923to direction of moldingSiC proportion in the10——shell (% by weight)ZrC proportion in the—22.3—shell (% by weight)SiC volume fraction in6.8——the shell (% by volume)ZrC volume fraction in—8.6—the shell (% by volume)Corrosion rate0.410.391.24-1.1at 1000° C. and 1%by volume H2O— vapor(mg/cm2 per hour) It can be seen from the table that density and load at break of the fuel element spheres are significantly improved by application of SiC or ZrC. It is emphasized that corrosion resistance is particularly improved. With values of 0.41 and 0.39 mg/cm3 per hour the corrosion rate is almost decreased by a factor of 3 when compared to the respective value for A 3 fuel element spheres and even by the factor of 1.7 when compared to the reactor graphite ATJ of the supplier UCC. The SiC proportion in the fuel element sphere shell of 10% by weight amounts to 5.32 g. This value is relatively low and confirms nearly a value of 5.28 g of the SiC coating of 23.300 fuel element particles in a sphere. These 23.300 particles together comprise 14 g uranium (all together). Because of the relatively high density of ZrC when compared to SiC the volume fractions of both carbides in the sphere shell only differ insignificantly. |
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description | This application claims priority to U.S. Provisional Patent Application Ser. No. 60/644,077, filed Jan. 14, 2005, and entitled “Synchronous Raster Scanning Lithographic System.” The entire application of U.S. Provisional Patent Application Ser. No. 60/644,077 is incorporated herein by reference. This invention was made with Government support under Grant Number HR0011-05-9-0001, modification number P00003, awarded by the Defense Advanced Research Projects Agency (DARPA). The Government has certain rights in this invention. The section headings used herein are for organizational purposes only and should not to be construed as limiting the subject matter described in the present application. The present invention relates generally to high throughput image formation. Many known imaging systems use single beam illumination to image complex shapes. For example, single beam illumination imaging systems are used in many products, such as cathode ray tube systems and plotters. Single beam illumination imaging systems are also used in lithographic systems for fabricating mask works for projection and contact lithography. In addition, single beam illumination imaging systems are used for direct exposure systems that write images directly on to a work piece. Recently, multi beam imaging systems have been used to increase imaging throughput. For example, multi-beam imaging systems are currently used in high throughput dot matrix printers, ink jet printers, and laser jet printers. While the present teachings are described in conjunction with various embodiments and examples, it is not intended that the present teachings be limited to such embodiments. On the contrary, the present teachings encompass various alternatives, modifications and equivalents, as will be appreciated by those of skill in the art. For example, some aspects of the imaging system of the present invention are described in connection with lithography systems used for fabricating semiconductor devices. However, it is understood that the imaging systems of the present invention have many applications and are not limited to lithography systems. For example, the imaging systems of the present invention can be used for direct writing technologies, such as ink, thermal, and laser printing technologies and also for phosphorous screen technologies. In addition, the imaging systems of the present invention can be used with single or multi-beam (pen) systems. It should be understood that the individual steps of the methods of the present invention may be performed in any order and/or simultaneously as long as the invention remains operable. Furthermore, it should be understood that the apparatus and methods of the present invention can include any number or all of the described embodiments as long as the invention remains operable. The term “photosensitive material” is defined herein to mean a material that is sensitive to a particle beam. The particle beam can include any type of particles, such as photons, electrons, ions, or X-ray. The exposure time of a single pixel Te of a photosensitive material having a know sensitivity can be expressed as: T e = S R E D ,where SR is defined as the sensitivity of the photosensitive material (e.g. uC/cm2 for electron sensitive materials and mJ/cm2 for photon sensitive materials), and ED is defined as the current density of the exposure beam (e.g. A/cm2 for electron beam based writing systems and mW/cm2 for optical beam based writing systems). For a single beam system, the maximum total cycle time to expose (write) a lithographic pattern in a given area can be expressed as:TC=Te*Npixels,where Te is the exposure time of a single pixel, and Npixels is the number of pixels in the exposure area. The number of pixels in the exposure area, where the required positional resolution GR is ≧spot size or spot resolution (Rmin) can be expressed as: N pixels = LW R m i n 2 ,where L is the length of the exposure area, and W is the width of the exposure area. However, when the positional resolution GR, is smaller than the spot size or spot resolution Rmin, such as in the case of direct write lithography, the number of potential pixels is then a function of the grid resolution and can be expressed as: N pixels = LW G R 2 . Thus for a single beam system with precise positional control requirements, the cycle time can be expressed as: T C = T e ( LW G R 2 ) . Multiple exposure beams can be used to reduce the cycle time. Multi-beam lithography systems according to the present invention use synchronous scanning architectures that move multiple beams or pens synchronously, at the same velocity. Each beam or pen can write (expose) independently. The cycle time is reduced by the number of beams Nbeams in the exposure area, which can be expressed as: N beams = LW D beam 2 ,where Dbeam is the distance between the beams assuming X and Y symmetry. The cycle time TC for multiple beams exposing a desired exposure area can be expressed as: T C = T e ( D beam G R ) 2 or T C = S R E D ( D beam G R ) 2 . Thus, where high pixel positional requirements exist, GR<Rmin, the total cycle time TC changes inversely as the square of the positional resolution GR. In order to double the lithographic positional resolution in these systems (decrease GR by a factor of two) and maintain a constant cycle time TC, one must either increase the current density ED by a factor of four or increase the number of beams by a factor of four. There are, of course, practical limitations on the magnitude of the current density and the number of beams that can be implemented. These practical limitations pose a significant impediment to obtaining high throughput from multi-beam synchronous scanning lithography systems that also maintain high positional resolution. An imaging system according to the present invention uses multiple exposure beams in combination with the threshold characteristics of the exposure medium to provide high image fidelity and high imaging throughput. To demonstrate the robustness of the methods and apparatus of the present invention, simulations were performed with a relatively poorly resolved spot energy. FIG. 1A is a simulation of a spot energy profile 100 for a poorly resolved 16 nm square electron beam that is used in the simulations described herein. FIG. 1B is a simulation of a dose profile 150 of the spot energy profile 100 shown in FIG. 1A that is scanned in one direction at a constant velocity along a single axis. The dose profile 150 was simulated for a beam that is energized for approximately three times the spot size, or 48 nm in the example presented herein. The desired final image is created by scanning the electron beam in two dimensions at increments equal to the desired grid resolution GR. FIG. 2A is a simulation of a multi-scan dose profile 200 of the energy profile 100 shown in FIG. 1A that is scanned at the desired grid resolution GR increments in two dimensions at a constant velocity along one axis according to the present invention. The electron beam is energized for approximately three times the spot size, which is 48 nm in both dimensions for the example presented herein. The simulation of the multi-scan dose profile 200 shows a high contrast dose profile, which indicates that a high contrast image can be formed. FIG. 2B is a cross section 250 of the multi-scan dose profile 200 shown in FIG. 2A that compares the starting energy profile shown in FIG. 1A with an ideal square wave profile. The cross section 250 shown in FIG. 2B indicates that the multi-scan method of the present invention is not sensitive to the starting beam profile. This result is in contrast to properties of known scanning methods which have lithographic resolutions that are particularly sensitive to the starting beam profile. Throughput of lithographic systems is typically a function of several parameters, such as the required dose, resolution, placement accuracy and the number of beams used in the system. Lithographic systems according to the present invention, which use multiple exposing beams and the threshold response of the developing mediums, do not have placement accuracies that depend on the system throughput. Thus, in one embodiment, a lithographic system according to the present invention uses a multi-beam source and the threshold characteristics of the exposure medium to provide significantly faster cycle time than known multi-beam lithography systems. In one embodiment of the invention, the exposure beam is overlapped during scanning so that the total spot current density ED is increased by the number of overlapped scans (exposures) NP compared to the nominal spot resolution, where NP can be expressed as: N p = ( R m i n G R ) 2 . Therefore, the cycle time TC of a lithographic system according to the present invention that uses multiple exposure beams and the threshold characteristics of the exposure medium can be expressed as: T C = S R E D N p ( D beam G R ) 2 or T C = T e ( D beam R m i n ) 2 or T C = S R E D ( D beam R m i n ) 2 . Thus, the cycle time TC of lithographic systems using the multi-beam method of the present invention is inversely proportional to the square of the spot size or spot resolution Rmin and is independent of the positional grid requirement. The spot size or spot resolution Rmin is typically much larger than the grid resolution GR. Therefore, the cycle time TC of lithographic systems using the multi-beam method of the present invention is significantly less than prior art lithography systems. The reduction in cycle time TC is achieved in part because the total dose to fully expose a given position is achieved with overlapping exposure beams. The following example illustrates the reduction in cycle time TC which can be achieved by using the multi-beam method according to the present invention. The cycle time TC of a known scanning method of exposing a resist having a resist sensitivity SR equal to 10 uC/cm2, a current density ED equal to 100 A/cm2, a spot size Rmin equal to 0.025 um, and a minimum placement resolution GR equal to 0.002 um is equal to 7,812 seconds (0.46 cycles per hour). In contrast, the cycle time TC using a multi-beam method according to the present invention with the same parameters is 50 seconds (72 cycles per hour). Thus, lithographic systems using the multi-beam method of the present invention can achieve a throughput that compares with the throughput of lithographic systems that have the minimum placement resolution GR equal to the spot size Rmin. In addition to reducing the cycle time TC, lithography systems of the present invention can use synchronous scanning to provide smooth blending of fields. The synchronous scanning also can be used to reduce the effects of noise, asymmetry, aberrations, and other defects in the beam that can be present in the writing spot (pen). Furthermore, the synchronous scanning of the present invention can be used to provide dose control. FIG. 3A illustrates a spot energy distribution 300 of a poor quality writing beam that would not be suitable for high resolution applications in known lithography systems. FIG. 3B illustrates a simulation of a multi-scan dose profile 350 using the spot energy distribution 300 shown in FIG. 3A. The simulation of the multi-scan dose profile 350 indicates that a well formed energy density profile can be achieved by using the method of performing multiple scans according to the present invention. In one embodiment of the invention, dose profiles overlap forming a blended dose profile. Complex images can be formed by blending dose profiles. Also, stitching and/or merging of patterns that cross boundaries between each beam in a multi beam system can be facilitated by blending dose profiles. FIG. 4A illustrates an exposure pattern 400 that facilitates dose blending. FIG. 4B illustrates a simulation of a multi-scan energy density profile 450 that indicates blending of dose profiles in the exposure pattern shown in FIG. 4A. The simulation of the multi-scan energy density profile 450 shown in FIG. 4B indicates blending of the cross patterns that produces a complex continuous dose profile that is difficult or impossible to achievable with known lithography systems. FIG. 5A illustrates a block diagram of a synchronous scanning multi-beam lithography apparatus 500 according to the present invention. A source array 502 includes beam generating columns arranged on predetermined centers that generate a plurality of beams of exposing radiation. The source array 502 can generate multiple beams of any type of exposing radiation including optical beams, electron beams, ion beams, and X-ray beams. The number of desired beams generated by the source array 502 is determined by the desired area to be lithographically defined or printed with the exposing radiation. A plurality of beam modulators modulates at least some of the plurality of exposure beams. In some embodiments, the source array 502 includes the plurality of beam modulators as indicated in FIG. 5A and as described in connection with the multi-electron beam source 800 shown in FIG. 8. In many embodiments, each of the plurality of beams of exposing radiation is modulated by one of a plurality of beam modulators. In many embodiments, the synchronous multi-beam lithography apparatus 500 includes a processor 504 or data generator having outputs that are electrically connected to control inputs of the plurality of beam modulators. The processor 504 or data generator generates electrical signals that instruct the plurality of beam modulators to modulate the plurality of beams of exposing radiation according to a desired exposure pattern. The block diagram 500 also shows a synchronous beam deflector 506 that simultaneously deflects all beams in the Y-axis direction (shown in FIG. 5B) by a predetermined distance. In many embodiments, the synchronous beam deflector 506 simultaneously deflects all beams in the Y-axis direction at a constant velocity. In many embodiments, the processor 504 or data generator includes at least one output that is electrically connected to a control input of the beam deflector 506. The processor 504 controls the timing of the beam deflection. In addition, the block diagram 500 shows a synchronous translation stage 508 where the substrate 510 or workpiece being exposed to the exposure beams is moved in the X-axis direction (shown in FIG. 5B). In many embodiments, the substrate 510 or workpiece being exposed to the exposure beams is moved in the X-axis direction at a constant velocity. In many embodiments, the processor 504 or data generator includes an output that is electrically connected to a control input of the synchronous translation stage 508. The processor 504 controls the timing of the synchronous stage 508 scanning shown in FIG. 5B. FIG. 5B illustrates a schematic diagram 550 of the operation of the synchronous scanning multi-beam lithography apparatus shown in FIG. 5A. For the purpose of illustrating the invention, the schematic diagram 550 shows a single optical beam 552 being asynchronously modulated (write/no-write) during the lithography process. The schematic diagram 550 also shows synchronous beam deflection 554 where the single optical beam 552 is synchronously deflected by a predetermined distance in the Y-axis direction by the beam deflector 506. In many embodiments, the optical beams are synchronously deflected by the predetermined distance at a constant velocity. In addition, the schematic diagram 550 shows synchronous stage translation or scanning 556 where the substrate 510 or workpiece being exposed to the exposure beams is moved along the X-axis direction. In many embodiments, the substrate 510 is moved along the X-axis direction at a constant velocity. For example, the exposure beams in the source array 502 can be arranged on a 500 μm grid center (i.e. Dbeam=500 μm) and can be simultaneously deflected in the Y-axis direction by 5 μm at a constant velocity with the beam deflector 506. A 5 μm×500 um exposure is performed for all beams after the translation stage 508 has traveled 500 um. This process is repeated by subsequent Y-axis direction stage motions of 5 μm microns, until each pixel in the desired exposure pattern is exposed. The entire area is exposed (written) simultaneously. FIG. 6 is a schematic diagram 600 illustrating the sequencing of a single beam multi-scan writing operation that forms two images in a photosensitive material according to the present invention. The schematic diagram 600 shows the beam modulation 602 in the “on” and in the “off” state. The schematic diagram 600 also shows the relative dose profile 604 from 0-100%. In addition, the schematic diagram 600 shows the resulting single scan image 606 having the beam of spot size SR. In addition, the schematic diagram 600 shows multiple overlapping scans 608. The dose threshold is reached by energizing and de-energizing the beam for a predetermined length of time while the beam is moving during the overlapping scans relative to the substrate or workpiece in one axis. The multiple overlapping scans are produced by translating the beam of spot size SR relative to a substrate in the X and Y directions. The exposures are repeatedly overlapped at increments of the positional resolution GR. The resultant image has an area of the spot size SR which has reached the threshold 100% dose level of the photosensitive material. One skilled in the art can appreciate that any geometry can be created by changing both the beam modulation in one axis and the length of overlap in the other axis. FIG. 7 illustrates a flowchart 700 of a method of performing synchronous multi-beam lithography according to the present invention. In a first step 702, the data for the pattern to be written to the entire substrate or workpiece is obtained. In a second step 704, the data is converted into a desired pixel format. The desired pixel format is one in which the data can be interpreted and used for defining the writing sequence of the multi-beams of exposing radiation. In a third step 706, the substrate is moved into position relative to the multi-beam array. In a fourth step 708, each of the exposure beams is energized or de-energized to write data in the individual pixels. In the fifth step 710, the method is repeated for the next position until the area between the beams of dimension Dbeams is exposed to the energized or de-energized beams. In the sixth step 712, the method is repeated over the entire area defined by the substrate. In the seventh step 714, the entire method is repeated for the next substrate. One skilled in the art will appreciate that the layout of the beams, size of the source array, and relative substrate motions can be optimized for a particular application. FIG. 8 illustrates an embodiment of a multi-electron beam source 800 for performing synchronous raster scanning lithography according to the present invention. The multi-electron beam source 800 includes a field emission array 802 that emits a plurality of electron beams for exposing an electron beam sensitive material. In the embodiment shown, the field emission array 802 is a cluster of one or more tips that can be individually energized to emit the plurality of electrons. A secondary electron multiplication array 804 is positioned in the path of the electrons generated by the field emission array 802. An output of the field emission array 802 is focused by an electrostatic lens array 806, which forms an array of focused beams 808. In some embodiments, the electrostatic lens array 806 also deflects the array of focused beams 808 in at least one of the X-axis and the Y-axis directions at the WP plane as shown in the figure. The array of focused beams 808 is directed onto a substrate 810 having a top surface coated with a photosensitive material 812. In one embodiment, the array of focused beams 808 is a two-dimensional array. Using a two-dimensional array will increase the number of beams in a particular source area and thus, will increase lithographic throughput. One skilled in the art will appreciate that the image forming and scanning technique described herein can be applied to numerous other devices. While the present teachings are described in conjunction with various embodiments and examples, it is not intended that the present teachings be limited to such embodiments. On the contrary, the present teachings encompass various alternatives, modifications and equivalents, as will be appreciated by those of skill in the art, may be made therein without departing from the spirit and scope of the invention as defined by the appended claims. |
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claims | 1. A method of causing fusion between particles, comprising:providing a magnet configured to generate an axial magnetic field;providing a cylindrical chamber located in said magnetic field;providing a target in said cylindrical chamber;providing a discharge electrode located in said cylindrical chamber;providing a voltage between said discharge electrode and an inner surface of said chamber to develop an electric field perpendicular to said magnetic field;supplying hydrogen gas into said cylindrical chamber;applying a voltage between said discharge electrode and an inner surface of said chamber containing said hydrogen gas to thereby create a plasma of ions and neutrals from said hydrogen gas in said chamber;said electric and magnetic fields being configured to cause said plasma of ions and neutrals to rotate within said cylindrical chamber by ion-neutral coupling;providing an electron emitter in said chamber configured to emit a quantity of electrons in said chamber under temperature conditions sufficient to reduce a Coulomb barrier between two approaching nuclei in said plasma by virtue of the collective negative charges of said quantity of electrons;said rotation of ions and neutrals being configured to induce a force that causes hydrogen atoms and protons in said chamber to press against the electron emitter;wherein a fusion reaction is caused to occur between neutral particles in said plasma and said target. 2. The method of claim 1, wherein the electron emitter comprises lanthanum hexaboride (LaB6). 3. The method of claim 1, wherein the electron emitter is connected to ground potential. 4. The method of claim 1, wherein said target is stationary within said chamber. 5. The method of claim 1, wherein said target rotates within said chamber. 6. The method of claim 1, wherein rotation of hydrogen atoms in said chamber is configured to cause neutrals and ions to be pushed to an inner surface of the discharge electrode by rotation of hydrogen atoms in said chamber, whereby densities of said neutrals and ions can be increased in dependence on the rate of rotation, resulting in a correspondingly higher rate of fusion, said rate of fusion being proportional to the product of the densities of reactants. 7. The method of claim 2, wherein:collisions between rotating hydrogen atoms and ions and a stationary LaB6 emitter result in increase in temperature of the target, giving rise to additional emitted electrons, thereby creating a further lowering of the Coulomb barrier and causing addition fusion reactions, which release energetic charged particles of MeV energies; said charged particles dissipating their energies inside the LaB6 target, resulting in a higher temperature. 8. The method of claim 1, wherein:ionization of hydrogen by charged particles from fusion is configured to lower the resistivity of the hydrogen medium, giving rise to a higher current between the discharge electrode and an outer electrode in said chamber, where this higher ion current impinges upon the outer electrode such that the increased population of electrons lower the coulomb barrier, enabling additional low-energy ions and neutrals to fuse by their proximity to each other, through quantum tunneling. 9. The method of claim 1, whereinan axial magnetic field is created by said magnet, which causes rotation of fusion reactants, where it produces elements and isotopes from rotating reactants. 10. The method of claim 1, wherein energetic alpha particles (He4) are produced as a result of said fusion reaction; the method further comprising using said energetic alpha particles for direct conversion to electrical energy. 11. The method of claim 1, wherein energetic alpha particles (He4) are produced as a result of said fusion reaction; the method further comprising slowing down said alpha particles, which are emitted in a same direction as original rotating hydrogen atoms, thereby yielding a charging current to an external power supply. 12. The method of claim 1 wherein said target comprises the element Boron having a dominant isotope of B11. 13. The method of claim 12, wherein the rotation of ions and neutrals in said chamber causes plasma particles to undergo the following fusion reaction as a result of repeated collisions between B11 and hydrogen neutrals or protons:p+B11→3He4+8.7 MeV energy. 14. A method of causing fusion between particles, comprising:providing a vacuum chamber;providing outer and inner electrodes defining a space therebetween within said vacuum chamber;introducing positive and negative ions into said space within said vacuum chamber;providing a series of alternating current (AC) fields on the circumference of said outer and inner electrodes, thereby driving said positive and negative ions introduced into said space within said vacuum chamber in rotation by an oscillating electric field;providing neutrals into said space, whereby said neutrals are driven in rotation around an axis of said electrodes through collisions with said positive and negative ions; andincreasing the frequency of such AC fields, thereby increasing the rotational speed of rotating ions and neutrals around said axis through resonant wave-particle interactions, said rotating ions and neutrals being configured to thereby cause a fusion reaction among said ions and neutrals. |
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abstract | A scintillator panel includes a support and a scintillator layer, wherein the scintillator layer includes scintillator particles, a binder resin, and a void, and the porosity of the scintillator layer is from 14 to 35% by volume. |
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summary | ||
06295329& | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to an apparatus and a method for handling reactor-internal equipments and, more particularly, an apparatus and a method for loading/unloading control rods, fuel supports, and control rod guide tubes into/from a boiling water reactor (referred to as "BWR" hereinafter). 2. Description of the Related Art FIG. 10 is a fragmental sectional view showing a part of a reactor-internal structure of the BWR in an enlarged manner. As shown in FIG. 10, in the BWR, a core plate 3 is provided to a shroud 2 which is constructed integrally with a reactor pressure vessel (reactor vessel) 1. Control rod drive mechanism housings (referred to as "CRD housings" hereinafter) 4 are provided to a bottom portion of the reactor pressure vessel 1 which is located at a lower portion of the core plate 3. A control rod drive mechanism (referred to as "CRD" hereinafter) 5 is installed inside the CRD housing 4. A control rod guide tube (referred to as "CRGT" hereinafter) 6 is provided on the CRD housing 4. The CRGTs 6 are supported to be slightly protruded upwardly from the core plate 3. A fuel support(referred to as "FS" hereinafter) 8 is detachably attached to an upper opening portion of the CRGT 6. The CRGT 6 and the FS 8 are fixed by a positioning pin 11 which is provided on the core plate 3. A control rod (referred to as "CR" hereinafter) 7 with a cruciform lateral sectional shape is provided in the CRGT 6 to be moved vertically. A bottom end of the CR 7 is coupled to a top end of the CRD 5. The CRs 7 can be inserted into clearances between the fuel assemblies 10 from top ends of the CRGTs 6 respectively. A top portion of the FS 8 supports a bottom portion 15 of the fuel assembly 10. Also, a top portion 12 of the fuel assembly 10 is supported by an top guide 13 which is provided to the shroud 2. FIG. 11 is a perspective view, partially omitted, showing an example of the CRGT 6 of the BWR. As shown in FIG. 11, a plurality of CRGT orifices 32 are formed on a top end side wall of the CRGT 6. FIG. 12 is a perspective view showing an example of the FS 8 of the BWR. As shown in FIG. 12, a cruciform through hole 9 into which the CR 7 is inserted is formed in the FS 8. Four fuel assembly sustaining holes 31, which sustain four fuel assemblies 10 respectively, are formed on a top portion of the FS 8. In addition, a plurality of fuel support orifices (referred to as "FS orifices" hereinafter) 33 are formed on a bottom side surface of the FS 8. FIG. 13 is a perspective view showing the fuel support with pin (referred to as "Pin-FS" hereinafter) 22 of the BWR. The Pin-FSs 22 are provided to support mimic fuel assemblies (not shown) arranged in a peripheral portion of the core. This Pin-FS 22 is constructed by adding a pin 53 to the FS 8 shown in FIG. 12. Then, if the CRs 7 are exchanged in the periodical inspection, the CRs 7 must be exchanged after the fuel assemblies 10 and the FSs 8, which are positioned at an upper location than the CRS 7, have been removed. The fuel assemblies 10 are unloaded by a refueling machine (not shown) which is installed on an operation floor (not shown), and then the CRs 7 and the FSs 8 are unloaded to outside the reactor. As an equipment employed to execute unloading operations of the CRS 7 and the FSs 8, a CR & FS grapple (not shown) which can grasp the CR 7 and the FS 8 at the same time has been proposed. After the CRs 7 and the FSs 8 have been unloaded from the reactor by the CR & FS grapple, the exchanging operation of the CRs 7 can be performed. In addition, there is a rare case where the CRGTs 6 must be removed. In this case, such CRGTs 6 can be removed by a CRGT grapple (not shown). However, in the event that the unloading operations of all the CRs 7, the FSs 8, and the CRGTs 6 from the reactor and the installing operations of them into the reactor must be performed in the periodical inspection or in the preventive maintenance, etc., there have been such problems that, if the CR & FS grapple and the CRGT grapple, for example, are employed, a time required for such operations is extremely prolonged and thus an exposure dose of the operator is increased correspondingly. Further, there has been such a problem that, if the CRs 7, the FSs 8, and the CRGTs 6 are unloaded from the reactor separately, it is difficult to secure respective storage spaces for them. SUMMARY OF THE INVENTION Therefore, it is an object of the present invention to provide a reactor-internal equipment handling apparatus and method capable of reducing a term of work which is needed for operations to load/unload control rods, fuel supports, and control rod guide tubes. In order to achieve the object, according to the present invention, there is provided a reactor-internal equipment handling apparatus comprising: control rod holding means for releasably holding a control rod which is loaded in a reactor vessel; fuel support/control rod guide tube holding means for releasably holding both a fuel support, which supports a bottom end of a fuel assembly, and a control rod guide tube, on which the fuel support is placed at top end; and a main body frame to which both the control rod holding means and the fuel support/control rod guide tube holding means are fitted and is adapted to be hung down inside the reactor vessel. Preferably, in the reactor-internal equipment handling apparatus, the fuel support/control rod guide tube holding means includes an orifice engaging member which is adapted to engage edge portions of orifices formed in the fuel support and the control rod guide tube, an orifice engaging member linking mechanism for manipulating the orifice engaging member, and orifice engaging member driving means for driving the orifice engaging member linking mechanism. Preferably, the reactor-internal equipment handling apparatus further comprises a stroke varying mechanism for varying an operating stroke of the orifice engaging member linking mechanism; wherein an amount of motion of the orifice engaging member is adjusted by changing the operating stroke of the orifice engaging member linking mechanism by the stroke varying mechanism such that the orifice engaging member engages only the edge portion of the orifice formed in the fuel support. Preferably, in the reactor-internal equipment handling apparatus, the orifice engaging member has stepped portions which come into contact with the edge portions of the orifices formed in the fuel support and the control rod guide tube, and wherein the orifice engaging member linking mechanism is constructed to disable a motion of the orifice engaging member in a situation that the stepped portion of the orifice engaging member comes into contact with the edge portion of the orifice. Preferably, in the reactor-internal equipment handling apparatus, the control rod holding means has a handle engaging member which is swingable and holds a hoisting handle provided on a top end of the control rod, and handle engaging member driving means for driving the handle engaging member to swing, and wherein the handle engaging member is formed of a hook member, and an own weight of the control rod is applied to hold a engaged state of the hoisting handle by the handle engaging member in a situation that the control rod holding means hoists the control rod via the handle engaging member. Preferably, in the reactor-internal equipment handling apparatus, the control rod holding means and the fuel support/control rod guide tube holding means are fitted to the main body frame such that these means can be relatively displaced mutually along a longitudinal direction of the control rod, and both the fuel support and the control rod guide tube are hoisted after the control rod is slightly hoisted. Preferably, the reactor-internal equipment handling apparatus further comprises a holding state detecting mechanism for detecting both a holding state of the control rod holding means about the control rod and a holding state of the fuel support/control rod guide tube holding means about the fuel support and the control rod guide tube; and a positioning state detecting mechanism for detecting a positioning state of the main body frame in the reactor vessel. Preferably, in the reactor-internal equipment handling apparatus, the holding state detecting mechanism has a holding state confirming indicator lamp whose lighting state is changed depending upon a change in the holding states of the control rod holding means and the fuel support/control rod guide tube holding means, and the positioning state detecting mechanism has a positioning state confirming indicator lamp whose lighting state is changed depending upon a change in the positioning state of the main body frame in the reactor vessel. Preferably, in the reactor-internal equipment handling apparatus, the positioning state detecting mechanism further includes a motion limiting mechanism for limiting a motion of the fuel support/control rod guide tube holding means when the main body frame is not properly placed at a predetermined position in the reactor vessel. Preferably, the reactor-internal equipment handling apparatus further comprises a motion limiting mechanism locking device for making the motion limiting mechanism inoperative temporarily. In order to achieve the object, according to the present invention, there is provided a reactor-internal equipment handling method of loading/unloading equipments, which are placed inside a reactor vessel, into/from the reactor vessel by a reactor-internal equipment handling apparatus, the reactor-internal equipment handling apparatus comprising control rod holding means for releasably holding a control rod which is loaded in the reactor vessel; fuel support/control rod guide tube holding means for releasably holding both a fuel support, which supports a bottom end of a fuel assembly, and a control rod guide tube, on which the fuel support is placed at top end; and a main body frame to which both the control rod holding means and the fuel support/control rod guide tube holding means are fitted and is adapted to be hung down inside the reactor vessel, the method comprising the steps of: holding the control rod by the control rod holding means and also holding both the fuel support and the control rod guide tube by the fuel support/control rod guide tube holding means; and hoisting the main body frame by a refueling machine and then simultaneously loading or unloading all of the control rod, the fuel support, and the control rod guide tube into or from the reactor vessel. Preferably, in the reactor-internal equipment handling method , the fuel support/control rod guide tube holding means includes an orifice engaging member which is adapted to engage edge portions of orifices formed in the fuel support and the control rod guide tube, an orifice engaging member linking mechanism for manipulating the orifice engaging member, and orifice engaging member driving means for driving the orifice engaging member linking mechanism, and the orifice engaging member is simultaneously brought into contact with the edge portions of the orifices formed in the fuel support and the control rod guide tube. |
claims | 1. A system for reducing contaminant deposition on a surface immersed in a fluid, the system comprising:a fluid source including a pump configured to draw the fluid from a cavity in which the apparatus is immersed; andan apparatus configured to discharge a fluid from the fluid source against a surface while immersed in the fluid, wherein the apparatus includes the pump such that the pump is immersed in the fluid. 2. The system of claim 1, wherein the fluid is a coolant water, and wherein the apparatus is further configured to discharge the coolant water to create a flow speed of the coolant water immediately on the surface of at least 2 meters per second. 3. The system of claim 1, further comprising:a filter within at least one of the fluid source and the apparatus, wherein the filter is configured to remove contaminants from the fluid prior to discharge from the apparatus, wherein the filter includes at least one stage of resins to remove dissolved contaminants in the fluid. 4. The system of claim 3, wherein the fluid source includes,a heat exchanger to cool the fluid prior to discharge from the apparatus below a temperature of the fluid in which the surface is immersed; anda chemical injector configured to modify a chemistry of the fluid prior to discharge from the apparatus. 5. The system of claim 1, wherein,the fluid source includes a base having the pump, a first filter, and piping together configured to draw the fluid from the cavity in which the apparatus is immersed in the fluid,the apparatus is a mobile assembly including an induction pump connected to the fluid source, andthe mobile assembly is moveable within the fluid to discharge the fluid on several different positions on the surface. 6. The system of claim 5, wherein,the fluid source further includes a heat exchanger and a chemical injector,the mobile assembly includes an intake and a second filter together configured to draw the fluid from the cavity in which the mobile assembly is immersed in the fluid, andthe mobile assembly is moveable by force from the induction pump discharging the fluid within the cavity. 7. The system of claim 6, wherein the intake and the second filter are connected to the induction pump so that the fluid is actively drawn through the intake and the second filter is entrained with fluid from the fluid source to create the discharged fluid. 8. A mobile apparatus to reduce contaminant depositions by inducing flows in a space flooded with a liquid, the mobile apparatus comprising:a flow intake receiving the liquid from the space; anda multi-stage filter configured to filter particulate and dissolved contaminants from the liquid; andan outlet directing the liquid toward a surface so as to create a flow of the liquid having speed of at least 2 meters per second of the liquid contacting the surface, wherein the flow moves the mobile apparatus within the space. 9. The mobile apparatus of claim 8, further comprising:a pump connected to the flow intake, wherein the pump is configured to create the flow while the mobile apparatus is completely immersed in the liquid in the space. 10. The mobile apparatus of claim 9, further comprising:a liquid connection configured to receive liquid from a source, wherein the pump is an induction pump that drives the liquid from the source and the liquid from the intake together to create the flow. 11. The mobile apparatus of claim 10, wherein the liquid from the source has a different temperature, a different pH, and a different oxidizer content from the liquid from the intake. 12. The mobile apparatus of claim 8, wherein the multi-stage filter includes,a coarse reservoir stage,a fibrous filter stage,a charged bed stage,a metallic filtering bed stage, anda resin bed stage secured between two screens. 13. The mobile apparatus of claim 12, wherein each of the stages are in separate sections of the multi-stage filter, and wherein the separate sections are removably attached to each other through flanges on ends of each of the sections. 14. The mobile apparatus of claim 12, wherein the resin bed includes a resin configured to filter metallic solutes from the liquid that are removed from depositing on the surface. 15. A method of removing contaminant depositions on open-system surfaces immersed in a water bearing contaminants, the method comprising:flowing the water at a speed of about 2 meters per second where the water meets the surface, wherein the water includes an oxidizer and a weak acid when flowed at the surface, wherein the flowing is produced by a mobile assembly completely submerged in the water, wherein the mobile assembly includes an induction pump causing the flowing the water; andfiltering the water for contaminants dissolved in the water as a result of the flowing. |
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abstract | An illumination optical unit includes a collector mirror which produces a polarization distribution that is applied to the first faceted optical element during the operation of the illumination optical unit. There are at least two first facet elements to which radiation having a differing polarization is applied. The first faceted optical element has at least one first state in which the normal vectors of the reflective surfaces of the first facet elements are selected so that a first predetermined polarization distribution results at the location of the object field during the operation of the illumination optical unit. |
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062696297 | abstract | A single energy storage capacitor is used to apply electrical energy to selected propellant modules, each module generating thrust in a desired direction by the high voltage induced vaporization of the copolymer spacer material in a conventional coaxial cable within the selected module. A lightweight, low cost, and EMI free thruster is thus provided. |
description | The present application claims priority from Japanese application JP2005-142095 filed on May 16, 2005, the content of which is hereby incorporated by reference into this application. 1. Field of the Invention The present invention relates to electron-beam length measurement technologies used in fabrication processes for such devices as semiconductor integrated circuits. More particularly, it relates to a high-accuracy electron-beam length measurement system and a calibration method therefor. 2. Description of the Related Art In the conventional calibration method for the electron-beam length measurement system, as disclosed in Patent Document 1 for example, the calibration is performed using a moiré diffraction image which is acquired by scanning an electron beam on a two-dimensional arrangement lattice sample with a certain angle set thereto. On account of high-accuracy implementation of the electron-beam length measurement system, it is increasingly becoming required that the deflection calibration therefor be replaced by the calibration which is based on an absolute dimension. This situation requires that the pitch size of a sample to be used for the calibration be replaced by a pitch size standardized as the absolute dimension. Also, the electron-beam length measurement system is so requested as to implement high-accuracy length measurements in semiconductor fabrication processes. Accordingly, the length measurements on insulating materials and materials which are likely to be easily deformed are absolutely necessary. This situation necessitates implementation of the measurements using a low-current electron beam whose beam irradiation gives rise to less damage. In the conventional two-dimensional arrangement lattice sample, however, there existed none of the technologies for arranging the lattice points with a pitch size which is precise in the transverse and longitudinal directions. Neither did the methods for measuring the pitch size precisely. Also, since the diffraction grating is constituted by an assembly of the lattice points, in the case of the low-current electron beam, secondary-electron signals acquired for each scanning are small in number. This resulted in a drawback that the accuracy becomes worse. In the method disclosed in JP-A-2003-022773, in the case of scanning the electron beam with an angle set to the lattice-point arrangement, locations at which the beam and the lattice points coincide with each other decrease in number. This fact resulted in a decrease in the acquired secondary-electron signals in number as well, which became a cause for degradation in the calibration accuracy. Moreover, when the point lattice is used, signals between the respective lattice points cannot be acquired. This resulted in a problem that the secondary-electron signal intensity acquired is not high enough. In the present invention, in order to allow the deflection be calibrated with a high accuracy in the electron-beam length measurement system, the calibration is performed such that the calibration will be divided into a direction perpendicular to the electron-beam scanning direction and a direction parallel thereto. Also, different methods will be used for the divided calibrations each. Regarding the deflection calibration in the direction parallel to the electron-beam scanning direction, first, a one-dimensional diffraction grating is located such that the direction of the diffraction grating becomes perpendicular to the scanning direction. Then, the pitch size of the diffraction grating is determined from a secondary-electron signal waveform acquired by scanning the electron beam. Next, the deflection is calibrated so that the determined pitch size of the diffraction grating and an actual pitch size of the diffraction grating will coincide with each other. Moreover, the one-dimensional diffraction grating is located such that the direction of the diffraction grating becomes parallel to the electron-beam scanning direction. Next, the electron-beam scanning is horizontally performed while displacing the electron-beam scanning in the perpendicular direction so that the electron-beam scanning displacement quantity will coincide with the pitch size of the diffraction grating. Here, from a secondary-electron signal image acquired, based on the presence or absence of moiré interference fringes, it can be judged whether or not the deflection calibration in the direction perpendicular to the electron-beam scanning direction has been correctly performed. Also, both the one-dimensional diffraction grating and the electron-beam scanning are of linear configurations. As a result, if the one-dimensional diffraction grating and the electron-beam scanning are caused to coincide with each other, the resultant secondary-electron signal intensity acquired proves larger than the one in the case of the lattice points. Accordingly, an excellent secondary-electron signal image can be acquired even in the case of the exceedingly-weak electron beam. Furthermore, the precise pitch size of the one-dimensional diffraction grating can be determined from diffraction-angle measurement on diffraction light diffracted thereby. This fact makes it possible to guarantee the absolute accuracy of the above-described calibration. Also, if the one-dimensional diffraction grating is of the superlattice cross-section structure of a multi-layer crystal composed of compound semiconductors, the precise pitch size thereof can be determined by measuring lattice number of the multi-layer crystal using devices such as transmission electron microscope. This fact makes it possible to guarantee the absolute accuracy of the above-described calibration. Hereinafter, the enumeration will be given below concerning representative configuration examples of the present invention. (1) An electron-beam system used in the present invention includes the following configuration components: An electron-optics system for performing an irradiation with an electron beam, a function for performing an electron-beam scanning on a one-dimensional-diffraction-grating calibrating mark by displacing the electron-beam scanning thereon by a predetermined transfer quantity, and a function for detecting reflected electrons or secondary electrons, and performing, from the detection result, calibration for a deflection quantity or deflection distortion of the electron beam, the reflected electrons or secondary electrons being emitted from the calibrating mark and a proximity thereto by the electron-beam scanning. (2) An electron-beam calibration method according to the present invention includes the following steps: Scanning an electron beam on a pattern of a one-dimensional-diffraction-grating calibrating mark provided on a stage, the electron beam being emitted from an electron source, displacing the electron beam by a constant transfer quantity by using a deflection unit, detecting reflected electrons or secondary electrons emitted from the calibrating mark and a proximity thereto by the electron-beam scanning, and performing, from the detection result, calibration for a deflection quantity or deflection distortion of the electron beam. According to the present invention, it becomes possible to implement the electron-beam system and a graphics-drawing method which allow the deflection distortion in the electron-beam system to be calibrated with an excellent accuracy. In particular, it becomes possible to implement the high-accuracy deflection calibration even in a low-acceleration and low-current electron-beam length measurement system. Other objects, features and advantages of the invention will become apparent from the following description of the embodiments of the invention taken in conjunction with the accompanying drawings. Hereinafter, referring to the drawings, the explanation will be given below concerning embodiments of the present invention. FIG. 1A illustrates the configuration of an electron-beam length measurement system used in the present embodiment. An electron beam 2 emitted from an electron gun (i.e., electron source) 1 is scanned on a sample 7 by using a deflector 4. A deflection calibrating mark 8 exists on a stage 9. Also, using an electron detector 10 for detecting secondary electrons 6 generated by the electron beam, a secondary-electron image is displayed, then making length measurement using the secondary-electron signal waveform. The length measurement is made as follows: In a line-width calculation unit, the pattern line-width is determined from the secondary-electron signal waveform. Next, in a line-width calibration calculation unit, the pattern line-width determined in the line-width calculation unit is corrected, then being displayed and memorized as the length-measurement value. Confirmation of the length-measurement position is performed using the secondary-electron image in an image display unit. Here, the explanation will be given below concerning a beam deflection calibration method with respect to the direction which is perpendicular to an electron-beam scanning direction. First of all, the stage is positioned so that the calibrating mark is directly under the electron beam. Next, the electron beam is scanned on a one-dimensional diffraction grating pattern which is perpendicular to the electron-beam scanning direction, then determining the pitch size from the secondary-electron signal waveform in the line-width calculation unit. Moreover, the line-width calibration calculation unit makes a comparison between the pitch size determined in the line-width calculation unit and a pitch size determined using an optical diffraction measurement method. Furthermore, the calibration calculation unit makes a correction to a beam deflection control unit so that the difference resulting from the comparison will become equal to zero, thereby performing the deflection calibration. After the calibration has been completed, once again, the electron beam is scanned on the one-dimensional diffraction grating pattern, then determining the pitch size from the secondary-electron signal waveform in the size calculation unit. Moreover, the line-width calibration calculation unit makes a comparison between the pitch size determined in the line-width calculation unit and the pitch size determined using the optical diffraction measurement method. Furthermore, the calibration calculation unit makes a correction to the measurement value so that the difference resulting from the comparison will become equal to zero. Next, in accordance with processing steps illustrated in FIG. 2, the explanation will be given below concerning a beam deflection calibration method with respect to the direction which is horizontal to the electron-beam scanning direction. First of all, the stage 9 illustrated in FIG. 1A is displaced, thereby causing the calibrating mark 8 to be positioned directly under the electron beam. The calibrating-mark substrate 8 illustrated in FIG. 1B includes not only the above-explained one-dimensional diffraction grating pattern which is perpendicular to the electron-beam scanning direction, but also a one-dimensional diffraction grating pattern which is horizontal to the electron-beam scanning direction. Next, as illustrated in FIG. 3, the electron beam is scanned on the one-dimensional diffraction grating pattern 12 which is horizontal to the electron-beam scanning direction. Using the beam deflection control unit, the electron beam is horizontally scanned while displacing the electron beam in the direction perpendicular to the electron-beam scanning direction so that the electron-beam scanning displacement quantity coincides with the pitch size of the one-dimensional diffraction grating pattern 12 determined using the optical diffraction measurement method. A secondary-electron image acquired in this way is displayed on the image display unit. As a result, in the secondary-electron image 11, moiré interference fringes 13 are observed in the electron-beam scanning portion on the above-described diffraction grating 12. This phenomenon is caused to occur by a situation where the pitch size of the diffraction grating 12 and the electron-beam scanning displacement quantity do not coincide with each other, i.e., there occurs a shift in the scanning position. Also, it can be confirmed from FIG. 3 that the beam deflection displacement quantity is uniformly larger than the pitch size of the diffraction grating 12. Therefore, as illustrated in FIG. 4, the electron beam is scanned once again on the one-dimensional diffraction grating pattern 14 horizontal to the electron-beam scanning direction, while the beam deflection control unit is adjusted until none of the moiré interference fringes have been observed within a secondary-electron image 15, or until the position shift quantity has become smaller than a predetermined threshold value. When the adjustment has been completed, a correct beam deflection parameter for adjusting the beam deflection displacement quantity is memorized into a calibration parameter memory unit inside a signal calculation unit. Based on this parameter, the beam deflection control unit corrects the beam deflection displacement quantity. Then, the secondary-electron image 15 after being corrected is displayed on a calibration status display unit. The calibration for the beam deflection displacement quantity using the one-dimensional diffraction grating patterns has been performed based on the images acquired. It is also possible, however, to perform the calibration for the beam deflection displacement quantity by executing waveform analysis of the moiré interference fringes in the calculation units. The above-described one-dimensional diffraction gratings are manufactured on the silicon substrate using laser interference exposure method and anisotropic etching. The pitch size of the one-dimensional diffraction gratings has been determined with an accuracy of 1 nm or less with reference to the pitch size of 240 nm by using the optical diffraction measurement method. As a result, the deflection calibration based on absolute values is made executable in the horizontal and perpendicular directions to the electron-beam scanning direction. Not only can the length-measurement value be acquired as the absolute dimension value, but also can the secondary-electron images with a precise magnification be acquired. This fact makes it possible to implement the high-accuracy length measurement, and simultaneously allows the system status to be confirmed in the calibration status display unit at whatever time. Furthermore, as a microscopic one-dimensional diffraction grating, a diffraction grating can be mentioned which uses the superlattice cross-section structure of a multi-layer crystal composed of compound semiconductors. This diffraction grating, which is used as a deep-groove sample as illustrated in FIG. 5, is manufactured as follows: The cross section of a structure, which is formed by repeating 5-nm-thick GaAlAs layers 17 and 5-nm-thick GaAs layers 18 over 40 layers on a GaAs substrate 16, is etched using an acid solution which causes only the GaAlAs layers 17 to be etched. By measuring lattice number of the multi-layer crystal using devices such as transmission electron microscope, the pitch line-width of this one-dimensional diffraction grating can be acquired as being the pitch size of 10 nm as an absolute dimension with the accuracy of 1 nm or less. This fact makes it possible to guarantee the absolute accuracy of the above-described calibration. The use of this one-dimensional diffraction grating allows implementation of the calibration with a higher magnification. FIG. 6 illustrates a deflection status where the beam deflection displacement quantity is locally modulated. The calibration therefor will be performed in accordance with basically the same processing steps as the ones illustrated in FIG. 2. First of all, the stage 9 illustrated in FIG. 1A is displaced, thereby causing the calibrating mark 8 to be positioned directly under the electron beam. At this time, using the beam deflection control unit, the electron beam is scanned on the calibrating mark 8 while displacing the electron beam so that the electron-beam scanning displacement quantity will coincide with the pitch size of a one-dimensional diffraction grating pattern 20 determined using the optical diffraction measurement method. A secondary-electron image acquired by the scanning is displayed on the image display unit. As a result, as illustrated in FIG. 6, moiré interference fringes 21 have been observed on the diffraction grating 20 and the electron-beam scanning portion in the secondary-electron image 19. This phenomenon is caused to occur by a situation where the pitch size of the diffraction grating and the electron-beam scanning displacement quantity do not coincide with each other. In view of this situation, the beam deflection control unit will make a correction to the beam deflection displacement quantity. Namely, for example, the pitch quantity of the electron beam will be locally changed, and the electron beam is scanned once again on the one-dimensional diffraction grating pattern horizontal to the electron-beam scanning direction. Then, the beam deflection control unit will be adjusted until none of the moiré interference fringes have been observed, or until the position shift quantity has become equal to a predetermined threshold value. A calibration parameter which has locally changed the beam deflection displacement quantity is memorized into a control unit 32, or preferably the image memory unit inside the signal calculation unit. Based on this parameter, the beam deflection control unit will correct the beam deflection displacement quantity. Then, this correction result is displayed on the calibration status display unit as the secondary-electron image after being corrected. The above-described processing steps allow the beam deflection displacement quantity to be calibrated into a correct beam deflection displacement quantity with none of the moiré interference fringes appearing as illustrated in FIG. 4, thereby making it possible to display the status on the calibration status display unit. FIG. 7 illustrates a deflection status where the pitch size direction of a diffraction grating and the electron-beam scanning displacement direction do not coincide with each other. Similarly to the deflection status described earlier, the calibration therefor will be performed in accordance with the processing steps illustrated in FIG. 2. First of all, the stage 9 illustrated in FIG. 1A is displaced, thereby causing the calibrating mark 8 to be positioned directly under the electron beam. Next, the electron beam is scanned in the direction which is horizontal to a one-dimensional diffraction grating pattern 23. At this time, using the beam deflection control unit, the electron beam is scanned while displacing the electron beam in the perpendicular direction so that the electron-beam scanning displacement quantity will coincide with the pitch size of the one-dimensional diffraction grating pattern 23 determined using the optical diffraction measurement method. A secondary-electron image 22 acquired in this way is displayed on the image display unit. As a result, moiré interference fringes 24 in an oblique direction as are illustrated in FIG. 7 can be confirmed on the diffraction grating 23 and the electron-beam scanning portion in the secondary-electron image 22. This phenomenon is caused to occur by a situation where the pitch size direction of the diffraction grating and the electron-beam scanning displacement direction do not coincide with each other. In view of this situation, the beam deflection control unit will make a rotation correction to the beam deflection displacement direction, and the electron beam is scanned once again on the one-dimensional diffraction grating pattern horizontal to the electron-beam scanning direction. Then, the beam deflection control unit will be adjusted so that none of the moiré interference fringes will be observed. A calibration parameter which has added the rotation to the electron-beam scanning direction is memorized into the control unit 32, or preferably the image memory unit inside the signal calculation unit. Based on this parameter, the beam deflection control unit will correct the beam deflection displacement quantity. Then, this correction result is displayed on the calibration status display unit as the secondary-electron image after being corrected. The above-described processing steps have allowed the beam deflection displacement quantity to be calibrated into the correct beam deflection displacement quantity with none of the moiré interference fringes appearing as illustrated in FIG. 4, thereby making it possible to display the status on the calibration status display unit. As described above, in the calibration according to the present embodiment, the areas are large where the one-dimensional diffraction grating and the electron beam coincide with each other. This characteristic allows the secondary-electron generation to be acquired in large quantities, thereby making it possible to obtain the secondary-electron signals with a large S/N ratio. In samples to be used in the length measurements, e.g., semiconductor samples, it is required to perform the length measurements using a low-current beam of 10 pA or less and at a low acceleration of several hundreds of volts. This low-current beam is required in order to avoid influences of electrification and beam irradiation damages especially in the case of oxide film samples and resist materials. With respect to the low-current beam of 10 pA or less and at the low acceleration of several hundreds of volts like this, the calibration according to the present embodiment has found it possible to guarantee the accuracy of 1 nm or less. In the conventional calibration method as illustrated in FIG. 8 which scans the electron beam with an angle set to the lattice-point arrangement, the locations at which the beam and the lattice points coincide with each other decrease further down to the one-tenth or less in number as compared with the present embodiment. As a result, the secondary-electron signals acquired also becomes smaller in number, i.e., the one-several-tenth or less as compared therewith. This fact has became a cause for degradation in the calibration accuracy, thus worsening the calibration accuracy to 100 nm or more. In the present embodiment, the explanation has been given regarding the case where the calibrating-mark substrate 8 includes not only the above-explained one-dimensional diffraction grating pattern which is perpendicular to the electron-beam scanning direction, but also the one-dimensional diffraction grating pattern which is horizontal thereto. In addition to this configuration, the following configurations are also allowable: Namely, as illustrated in FIG. 11, two calibrating-mark substrates are prepared on one and the same stage. Then, patterns of these calibrating-mark substrates are located such that the substrates become a calibrating-mark substrate of the one-dimensional diffraction grating pattern perpendicular to the scanning direction, and a calibrating-mark substrate of the one-dimensional diffraction grating pattern horizontal thereto. Also, a rotation mechanism is prepared onto the stage, thereby rotating, by 90°, the calibrating-mark substrate of the one-dimensional diffraction grating pattern perpendicular to the scanning direction. Otherwise, the rotation mechanism is prepared onto only the calibrating-mark substrate of the diffraction grating, thereby changing the one-dimensional diffraction grating pattern perpendicular to the scanning direction to the one-dimensional diffraction grating pattern horizontal thereto. Next, the explanation will be given below concerning another embodiment. The calibration therefor will be performed in accordance with the processing steps illustrated in FIG. 2. First of all, in FIG. 1A, the stage 9 is displaced, thereby causing the calibrating mark 8 to be positioned directly under the electron beam. Next, in FIG. 3, the electron beam is scanned on the one-dimensional diffraction grating pattern 12 which is horizontal to the electron-beam scanning direction. At this time, using the beam deflection control unit, the electron beam is horizontally scanned while displacing the electron beam in the direction perpendicular to the electron-beam scanning direction with a period which is smaller than the pitch size of the one-dimensional diffraction grating pattern 12 determined using the optical diffraction measurement method. Moreover, the detection by the secondary-electron detector 10 illustrated in FIG. 1A is intermittently performed in correspondence with a period during which the electron-beam scanning is positioned on the diffraction grating. The secondary-electron image 11 acquired in this way is displayed on the image display unit. As a result, as illustrated in FIG. 3, the moiré interference fringes 13 have been observed in the electron-beam scanning portion on the above-described diffraction grating 12 in the secondary-electron image 11. This phenomenon is caused to occur by a situation where the pitch size of the diffraction grating and the electron-beam scanning displacement quantity corresponding thereto do not coincide with each other. In view of this situation, the beam deflection control unit will make a correction to the beam deflection displacement quantity. Namely, the electron beam is scanned once again on the one-dimensional diffraction grating pattern horizontal to the electron-beam scanning direction. Then, the beam deflection control unit will be adjusted so that none of the moiré interference fringes will be observed as is illustrated in FIG. 4. As a result, in FIG. 3, it can be recognized that the beam deflection displacement quantity is uniformly larger than the pitch size of the diffraction grating. Accordingly, a calibration parameter for making the beam deflection displacement quantity uniformly smaller is memorized into the signal calculation unit. Based on this parameter, the beam deflection control unit will correct the beam deflection displacement quantity. Then, this correction result is displayed on the calibration status display unit as the secondary-electron image after being corrected. The pitch line-width of the above-described one-dimensional diffraction grating has been determined with the accuracy of 1 nm or less by using the optical diffraction measurement method. As a result, the secondary-electron image with a precise magnification can be acquired. This fact makes it possible to implement the high-accuracy length measurement, and simultaneously allows the system status to be confirmed in the calibration status display unit at whatever time. Next, the explanation will be given below concerning still another embodiment. The calibration therefor will be performed in accordance with the processing steps illustrated in FIG. 2. First of all, in FIG. 1A, the stage 9 is displaced, thereby causing the calibrating mark 8 to be positioned directly under the electron beam. Next, in FIG. 3, the electron beam is scanned on the one-dimensional diffraction grating pattern 12 which is horizontal to the electron-beam scanning direction. At this time, using the beam deflection control unit, the electron beam is horizontally scanned while displacing the electron beam in the direction perpendicular to the electron-beam scanning direction with a period which is smaller than the pitch size of the one-dimensional diffraction grating pattern determined using the optical diffraction measurement method. Moreover, in the signal calculation unit, an addition processing is performed between a secondary-electron image acquired as a result of the above-described electron-beam scanning, and a reference image, e.g., an ideal secondary-electron image and CAD data. Here, the reference image is memorized in the image memory unit, and corresponds to the pitch size of the one-dimensional diffraction grating. Furthermore, a consequence acquired as a result of the above-described addition processing is displayed on the image display unit. As a result, as illustrated in FIG. 9, moiré interference fringes 29 have been observed on the reference image 28 and the electron-beam scanning portion in the secondary-electron image 27. This phenomenon is caused to occur by a situation where the pitch size of the diffraction grating and the electron-beam scanning displacement quantity corresponding thereto do not coincide with each other. In view of this situation, the beam deflection control unit will make a correction to the beam deflection displacement quantity. Namely, the electron beam is scanned once again on the one-dimensional diffraction grating pattern horizontal to the electron-beam scanning direction. Then, the beam deflection control unit will be adjusted so that, as is illustrated in FIG. 10, none of the moiré interference fringes will be observed on the reference image 30 and the electron-beam scanning portion in the secondary-electron image 31. As a result, in FIG. 3, it can be recognized that the beam deflection displacement quantity is uniformly larger than the pitch size of the diffraction grating. Accordingly, a calibration parameter for making the beam deflection displacement quantity uniformly smaller is memorized into the signal calculation unit. Based on this parameter, the beam deflection control unit will correct the beam deflection displacement quantity. Then, this correction result is displayed on the calibration status display unit as the secondary-electron image 31 after being corrected. The pitch line-width of the one-dimensional diffraction grating has been determined with the accuracy of 1 nm or less by using the optical diffraction measurement method. As a result, the secondary-electron image with a precise magnification can be acquired. This fact makes it possible to implement the high-accuracy length measurement, and simultaneously allows the system status to be confirmed in the calibration status display unit at whatever time. As the above-described one-dimensional diffraction grating, the following diffraction gratings have been employed: The diffraction grating manufactured on the silicon substrate using the laser interference exposure method and anisotropic etching, or the diffraction grating using the superlattice cross-section structure of the multi-layer crystal composed of compound semiconductors. As having been explained in detail in the above-described embodiments, according to the present invention, it becomes possible to calibrate the deflection distortion in the electron-beam system with an excellent accuracy. It should be further understood by those skilled in the art that although the foregoing description has been made on embodiments of the invention, the invention is not limited thereto and various changes and modifications may be made without departing from the spirit of the invention and the scope of the appended claims. |
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abstract | An X-ray generator introducing an X-ray to an illumination optical system includes plural plasma light sources, and a reflector, movably arranged among the plural light sources, for switching light sources and for reflecting the X-ray from one of the plural light sources to the illumination optical system, wherein an angle between a plane determined by an optical axis of the X-ray emitted from the reflector and a line that connects the plural light sources to the reflector, and a polarization plane on which an electric field vector oscillates is between 45° and 135°, the polarization plane maximizing a reflectance to the X-ray of the illumination optical system. |
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description | The field of the invention relates to the prediction of, the protection against and the diagnostics of pellet-clad interaction stress-corrosion-cracking failure of nuclear fuel rods. The field more particularly relates to assessment of susceptibility to PCI failure for guidance in the design of nuclear reactors. The field additionally relates to the protection against PCI failure by providing operational information to operators of a nuclear reactor during power maneuvering. Additionally, the field relates to the diagnostics of an event suggesting a possible PCI failure. Pellet-Clad Interaction—Fuel Rod Failure Strong pellet-clad interaction resulting in high stress in the cladding is one of the lead causes of preventable fuel failure in light water reactors. This type of reactor is fueled with multiple assemblies where the basic element of each assembly is a single fuel rod. A fuel rod is made of a long cylindrical tube of a zirconium alloy inside which a stack of uranium oxide pellets occupies the majority of the volume. Each of the fuel pellets is generally cylindrical in shape. Geometrical variations of the pellet structure include dishing the flat ends of the pellet, chamfering of the pellet edges, introducing a vertical hole in the pellet center, and varying the pellet height-to-diameter ratio. Material variations of the fuel pellets include the ceramic density relative to its maximum possible theoretical value, i.e. variation in the pellet material porosity. Pellet material variation includes the isotopic composition of the fissionable elements, i.e. uranium enrichment in the fissile isotope U-235 and the possibility of including plutonium oxide in the ceramic matrix. Another common variation of the pellet material is the addition of a neutron absorber for reactivity control, such as gadolinium or boron. Additives to the fuel pellets may include trace elements for modifying its microstructure such a chromium oxide doping. The fuel tube, henceforth called clad or cladding, may also vary in its dimensions and material composition. The base material is zirconium alloyed with varying percentages of other elements, most common alloys being the well known Zircaloy 2 and Zircaloy 4. The clad structure is generally a thin-walled tube, where the diameter and the wall thickness-to-diameter ratio vary from design to design. In some fuel designs, the inner surface of the cladding tube is made of a so-called liner or barrier which is a thin layer of soft zirconium or a different zirconium-based alloy. The cladding tube, sealed at one end, is loaded with fuel pellets and then pressurized with helium gas before it is sealed at the other end. While the stack of fuel pellets occupies the majority of the fuel rod volume, there is a free volume reserved for the collection of fission gasses mainly in the upper part, the so-called upper plenum, and optionally at the lower end. The upper plenum contains a spring to press the pellet stack down and maintain the fuel stack continuity. The inner diameter of the cladding is slightly larger than the outer diameter of the fuel pellets, leaving a space of the approximate size of 100 microns, henceforth called the gap or pellet-clad gap, which also varies depending on fuel design and manufacturing process. Operational Effects on Fuel Rod During reactor operation, the fuel rods are subjected to extreme conditions which must be understood to protect the integrity of the fuel rods. External to the fuel rods, the coolant is pressurized to approximately 7 MPa in the case of boiling water reactors (BWR) and 14 MPa in the case of pressurized water reactors (PWR). Under neutron irradiation, the nuclei of the heavy fissionable isotopes split into medium weight nuclei and emit additional neutrons to sustain the fission chain reaction while releasing energy. The heat generated in fuel rods per unit length is so high that steep temperature gradients between the center and surface of the fuel pellets develop which results in large internal stress gradients due to thermal expansion of the pellet material, with the effect that the ceramic pellet material is deformed and is often cracked. The multiple effects on the fuel pellets include volume swelling due to the fact that there are two fission product nuclei for each heavy isotope that undergoes fission. There is also the opposite effect of pellet volume reduction due to the collapse of the pores. The net volume change is initially negative, but a net volume increase occurs as the exposure to neutron flux continues and the so-called burnup increases. Some of the fission products are gaseous, such as Xe, and a fraction of these gases escape outside the pellet material microstructure into the free volume inside the fuel rod causing its pressure to increase. There are numerous chemical compounds that form from the fission products, and some of these have corrosive effects on the cladding material. Notable among the corrosive chemical agents is iodine, which forms a series of zirconium iodides, and the most corrosive of these iodides is ZrI4. The irradiation effects on the clad include the microstructure defects that occur when the fast neutrons (energy >1 MeV) collide with the zirconium nuclei. The effect of the cumulative fast neutron flux, the so-called fluence, reduces the ductility of the cladding material, and also promotes the permanent deformation, the so-called irradiation-induced creep, when the material is subjected to mechanical stresses. Clad elongation under fast neutron irradiation also occurs without the influence of external stress, and so-called temperature-induced creep occurs under external stress without the presence of fast neutron irradiation. However, the processes involved are so complicated such that temperature-induced creep rate is also affected by the fast neutron fluence. The mechanical stresses on the clad include the compressive component due to the external coolant pressure, which cause the so-called clad creep-down, which is accelerated under the influence of fast neutron irradiation, with the general effect of reducing the pellet-clad gap. A tensile stress component in the clad occurs when the gap vanishes as the pellet and clad come into direct contact, a condition called pellet-clad-mechanical interaction (PCMI) or pellet-clad interaction (PCI). The tensile stress that occurs due to PCMI is in the tangential direction, also called hoop stress. The magnitude of the tensile stress increases with the processes that cause the pellet volume to increase and decreases (stress relief) as the cladding material creeps out in response to the imposed stress. It should be noted that the zirconium alloys used for cladding material are highly anisotropic where the degree of anisotropy is dependent on the heat treatment of the cladding, i.e. being fully or partially re-crystallized under relatively high temperature as part of the manufacturing process. Large tensile stresses may result when a gap is initially closed before a large power increase causes the pellet temperature to increase and subsequently pellet volume increase due to thermal expansion. The amount of pellet volume increase is proportional to the power increase, and is larger when the pellet thermal conductivity is low, the latter being the case when the pellet temperature is high and when its burnup is high. Stress-Corrosion-Cracking Failure Mode The corrosive chemical interaction of iodine released from the irradiated pellets and zirconium cladding material is greatly accelerated under tensile stress. This process is called stress-corrosion-cracking (SCC). Thus, a strong PCI which generates high tensile stress is considered the direct cause for the corrosive cracking of the cladding material which may result in the breach of the cladding integrity and the release of radioactive material outside the cladding tube into the coolant. The prevention of this so-called PCI failure mode is the subject of this patent. The chemical, mechanical, and thermal processes that occur in the fuel rod are very complicated beyond the brief description given in the above section, and are extensively covered in the open literature such as cited here as references (1) through (5). These processes are commonly modeled using large computer programs for simulating fuel performance, so-called fuel performance codes. One such code is RODEX4 (reference 6), also known as SIERRA code (reference 7) used by AREVA, and there are other codes of the same class which are used by other organizations (for example STAV at Westinghouse Electric Sweden, PRIME at General Electric, FRAPCON at Pacific Northwest National Laboratory, FALCON at Anatech Inc.). The common attribute of all fuel performance codes is their high level of sophistication in modeling the fuel performance by solving a large number of equations describing numerous interacting local parameters. Fuel performance codes are used routinely for fuel rod design activities and also for post-failure analysis of the causes that lead to clad failure. The application of fuel performance codes prior to a power maneuver in a reactor to predict whether large stresses will develop is an obvious extension to their field of applicability, however, the intensive computer resources required for such application is prohibitive particularly in the light of the large number of fuel rods in a reactor core. The technique commonly used for achieving a measure of protection against PCI failure is the online application of empirical rules, also called maneuvering guidelines, to restrict the so-called linear heat generation rate in representative fuel rods and limit the time rate of its increase during a reactor power maneuver. The linear heat generation rate (LHGR) is also referred to as linear power or fuel duty, and is defined as the thermal power generated in a fuel rod per unit length measured in units of kW/ft or equivalently kW/m. Some maneuvering guidelines, relate the allowable LHGR to a reference LHGR value, interchangeably called the conditioned LHGR or the conditioned state LHGR or the conditioning state. The conditioned LHGR is qualitatively defined as the asymptotic operational LHGR where operation for sufficiently long time leads to mechanical equilibrium where the pellet-clad contact pressure presumably settles to a safe low value. By analogy to an athlete's training or conditioning, a fuel rod is capable of operating at high power because it has been trained or conditioned to operate at high power, and the reverse is true where a fuel gradually loses the capacity to operate at high power after operating for a period of time at low power. The time constants for the dynamic tracking of the conditioned state are generally obtained by empirical means using empirical formulas. According to the prior art maneuvering guides, reactor power increase using control rod motion is restricted such that the operating LHGR for a monitored rod at a given elevation does not exceed its conditioned state within a prescribed tolerance. Thus, the linear power, P, at a given elevation of a monitored fuel rod is restricted such thatP≦PCS+ΔP where PCS is the conditioned state, and ΔP is an empirically prescribed tolerance.For slow power increase maneuvers such as achieved by increasing the recirculation flow in a boiling water reactor, a ramp rate upper limit is prescribed where the ramp rate is defined as the time rate of increase of LHGR commonly measured in units of kW/ft/hr. Although the above described empirical power maneuvering is applicable to both Boiling and Pressurized Water Reactors, local implementation (at the level of single fuel assembly segment or a single fuel rod segment) is common in Boiling Water Reactors. Pressurized Water Reactor maneuvering is commonly based on a global variable representing the entire core thermal power, taking advantage of the relative simplicity and invariance of power distribution in its core. Maneuvering guidelines are largely based on experience derived from experiments with power ramping of sample rods in test reactors, trends of the results of fuel performance codes, and actual PCI failures in power reactors. Due to their empirical nature, a balance must be considered between the level of protection that can be achieved and the operational flexibility a reactor operator needs, with the result that only a limited success is possible and PCI failures may continue to occur at a reduced frequency. PCI failures occur from time to time despite the occasional adjustment of the maneuvering guides following such events. The occurrence of fuel failures of any type requires power suppression and occasionally a mid-cycle outage, both are associated with environmental and economic cost. The space permitted for reactor power maneuvering evolved to minimal flexibility with each revision that introduces new conservatism to the maneuvering guidelines and power ascension is being restricted to slow rates that result in reducing the plant capacity factor and profitability. A novel method is needed to reliably prevent PCI failures while maintaining reactor power maneuvering flexibility. 1. Donald R. Olander, “Fundamental Aspects of Nuclear Reactor Fuel Elements,” Published by Technical Information Center, Office of Public Affairs, Energy Research and Development Administration, 1976, ISBN 0-87079-031-5 (v. 1) 2. John Gittus, ed., “Water Reactor Fuel Element Performance Computer Modelling,” Applied Science Publishers LTD 1983, ISBN-0-85334-217-2 3. Proceedings of the International Topical Meeting on LWR Fuel Performance, Avingnon France, Apr. 21-24, 1991 4. Proceedings of the International Topical Meeting on Light Water Reactor Fuel Performance, West Palm Beach, Fla., Apr. 17-21, 1994 5. Proceedings of the International Topical Meeting on Light Water Reactor Fuel Performance, Portland Oreg., Mar. 2-6, 1997 6. M. R. BILLAUX, “Modeling Pellet-Cladding Mechanical Interaction and Application to BWR Maneuvering,” Proceedings of the International Meeting on LWR Fuel Performance, Paper 1047, Orlando, Fla., Sep. 19-22, 2004 7. M. R. Billaux, S.-H. Shann, L. F. van Swam, F. Sontheimer, and H. Landskron, “SIERRA: A Code to Predict the Mechanical Behavior of LWR Fuel Rods,” 14th International Conference on Structural Mechanics in Reactor Technology, Lyon France, Aug. 17-22, 1997 Patents and patent applications broadly related to the art of the invention herein are found as follows: Patent Application 2006165210A1 to Karve et al; U.S. Pat. No. 6,400,786 to Mourlevat et al; U.S. Pat. No. 6,430,247 to Mourlevat et al; U.S. Pat. No. 6,535,568 to Reese: Patent Application 2006/0146973 to Yeager et al. The patents referred to herein are provided herewith in an Information Disclosure Statement in accordance with 37 CFR 1.97. The publications referred to herein will be provided, in accordance with 37 CFR 1.97 and 37 CFR 1.98 within the time allowed by said regulations. The PCI related fuel failure problem associated with nuclear reactor power maneuvering is solved by this invention with introducing a new model that is: 1. A reduced order model, henceforth called XEDOR, in the sense that the number of equations and parameters are far fewer than existing fuel performance codes. 2. By virtue of its reduced order, its algorithmic representation in a computer code is so fast that it permits online application in the sense that predictive calculations are performed and the results taken into account prior to executing any power maneuver in a reactor. 3. It is physically based, unlike empirical maneuvering guides of the prior art, and therefore provides for a qualitatively higher degree of accuracy and reliability. 4. Capable of predicting the stress level and the associated LHGR margin to the maximum allowable limits, and doing so using definite representation of each fuel design parameters and material properties. 5. Of definite structure as to its formulas and internal variables and input-output compatibility with reactor monitoring software.The model formulation is fundamentally based on first order differential equations describing the time evolution of the two main state variables, pellet-clad gap size and clad stress, and driven by the operating parameters obtained from the reactor instruments and monitoring software as represented by the linear heat generation rate and fast neutron flux for each fuel rod at several axial locations. Reactor operator is assisted in performing power maneuvers by being provided with timely information regarding the stress levels predicted by XEDOR for that maneuver, whereby the maneuver is allowed only provided that a stress threshold specific to the subject fuel type is not exceeded. This application discloses a reduced order model, called XEDOR, for calculating stress and pellet-clad gap size in at least one fuel rod segment to provide information for core loading pattern design, operator power maneuvering and/or diagnostics of suspected PCI cladding failure comprising: a. calculating a gap size evolution from its initial as-manufactured value by ⅆ δ ⅆ t = f ( σ ) - a 0 P - a 1 ⅆ P ⅆ t ( 4 ) and, b. calculating an evolution of clad hoop stress with time upon gap size reaching zero by 1 E ⅆ σ ⅆ t + f ( σ ) = a 0 P + a 1 ⅆ P ⅆ t ( 3 ) where t is time, δ is the pellet-clad gap size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is the clad inelastic strain rate as function of hoop stress while being irradiated with a given fast neutron flux level and having accumulated a fast neutron fluence, specific for a given clad material and heat treatment, P is the linear heat generation rate in pellets of the fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power and, c. displaying the calculated parameters, σ and/or δ, for use in core loading pattern design, operator power maneuvering or diagnostics of suspected PCI cladding failure. The calculated parameters, σ and/or δ, are displayed for use in core loading pattern design, operator power maneuvering or diagnostics of suspected PCI cladding failure and said calculated parameters are utilized in guiding the design of a core loading pattern of a nuclear reactor and used by an operator for power maneuvering and in the diagnosing of a suspected PCI cladding failure in a nuclear reactor all related to nuclear reactor operations. A computer program named XEDOR is disclosed herein, which is inputted into a computer implementing the algorithms for calculating the gap size evolution from its initial as-manufactured value by ⅆ δ ⅆ t = f ( σ ) - a 0 P - a 1 ⅆ P ⅆ t ( 4 ) and, b. calculating the evolution of clad hoop stress with time upon gap size reaching zero by 1 E ⅆ σ ⅆ t + f ( σ ) = a 0 P + a 1 ⅆ P ⅆ t ( 3 ) where t is time, δ is the pellet-clad size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is the clad inelastic strain rate as function of hoop stress for given fast neutron flux level and cumulative fast neutron fluence, specific for a given clad material and heat treatment, P is the linear heat generation rate in pellets of the fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power. Regarding power maneuvering, the computer program XEDOR, provides the following: a. tracking the maximum allowable linear power compared with the operating linear power in order to ensure that a safety margin to PCI fuel failure exists; projecting a safe power trajectory in order to plan maneuvers such as start up and control rod sequence exchange; b. dividing the projected safe power trajectory into a safe power jump followed by a continuous safe power ramp; c. predicting a safe power jump by determining the gap and hoop stress when a safe power trajectory projection is requested; calculating the safe power jump as the power corresponding to thermal expansion that consumes the present gap plus clad elastic strain corresponding to a given safe stress setpoint; d. predicting an instantaneous power ramp rate where the stress increase due to pellet thermal expansion is balanced by the stress relief due to creep in the cladding material, the power ramp rate being limited toR=(ƒ(σ*)−a0P)/a1 (12) e. predicting a continuous power ramp [0040] by limiting the power increase rate such that the stress increase due to pellet thermal expansion is balanced by the stress relief due to creep in the cladding material, wherewith the linear power is calculated from a 1 ⅆ P ⅆ t + a 0 P = f ( σ * ) ( 11 ) which is integrated numerically by the reduced order model XEDOR whenever a safe power projection over a period of time is requested by the reactor operator; f. displaying the predicted safe power jump and power ramp for nuclear reactor operations and for operator consideration in power maneuvering for nuclear reactor operators for safe power jump and power ramp and in guiding the design of a core loading pattern of a nuclear reactor. Regarding power maneuvering, the disclosed computer program XEDOR, is utilized as follows: a. inputting into a computer a reduced order computer code XEDOR wherein algorithmic representations, in a computer code, receive data from core instrumentation and monitoring software programs thereby permitting online application for predictive calculations which are performed and yielding calculations prior to executing any power maneuver in a reactor; b. displaying said calculations in physical operator displays for operator consideration where calculations are physically based, unlike empirical maneuvering guides, and providing a qualitatively higher degree of accuracy and reliability for power maneuvering safety; c. demonstrating, via the operator display, calculations which are predictive of the stress level and the associated linear heat generation rate margin to the maximum allowable limits using definite representation of each fuel design parameters and material properties. Again, regarding power maneuvering, the program XEDOR, with XEDOR algorithms, is loaded into a computer having a computer-readable storage medium having computer program logic stored thereon for enabling a processor to execute the XEDOR algorithm, which, upon receipt of operational data from a reactor core monitoring system, calculates margin to PCI failure with said calculations displayed and used for reactor operator and reactor operations use for executing a safe power maneuver. The foregoing and other features and advantages of the present invention will become more readily appreciated as the same become better understood by reference to the following detailed description of the preferred embodiment of the invention when taken in conjunction with the accompanying drawings, wherein: A model is developed to calculate hoop stress in the clad when the pellet-clad gap is closed, at which condition the pellet-clad contact pressure contributes a tensile component. When the gap is open, hoop stress is a result of the external coolant pressure contributing a compressive component, which results in a nearly constant stress when the variation of inner clad gas pressure is recognized as insignificant. The model therefore needs to calculate the gap size and transition to a dynamic stress calculation when the gap is closed. A conceptual model is outlined below. For the condition of a closed gap, the cladding relative dimensional change, also known as strain, is driven by the pellet's relative dimensional change. The clad strain rate balance is written asElastic strain rate+Inelastic strain rate=Pellet-expansion-imposed strain rate.Thus, the stress equation is formulated as 1 E ⅆ σ ⅆ t + f ( σ ) = g ( t ) ( 1 ) where the ratio between the clad stress, σ, and its elastic strain is the so-called Young's modulus, E. The function, ƒ(σ), signifies the clad inelastic strain rate, also known as creep rate. The function, g(t), signifies the rate of the pellet relative dimensional change as function of time.The strain imposed by the pellet as a given time, G(t), is the sum of two components: the first is due to the cumulative burnup which is proportional to the power integral, and the second due to thermal expansion which is proportional to the power itself. Thus,G(t)=a0∫P(t)dt+a1P(t) (2)where, P is the linear heat generation rate in the pellet, the coefficient a0 is the pellet strain rate per unit change in linear heat generation rate, and the coefficient a1 is the pellet strain resulting from a unit increase in linear heat generation rate.With g(t)=dG(t)/dt=a0P+a1dP/dt, Equation (1) becomes, 1 E ⅆ σ ⅆ t + f ( σ ) = a 0 P + a 1 ⅆ P ⅆ t ( 3 ) For the alternative condition of an open gap, the clad hoop stress is negative and is assumed to be a constant dominated by the external coolant pressure. Clad creep down under compressive hoop stress in combination with pellet volume eventual increase under neutron irradiation will gradually close the gap. The calculation of gap size is important in modeling stress because, for power ramps starting with an open gap and terminating with a closed gap, the peak stress level depends on the initial gap size, where an initially large gap size results in a reduced stress peak and vice versa. Also, transitioning to Eq. (3) for calculating stress is triggered by reaching a zero gap size, the determination of which requires a calculation to track the evolution of the gap size. The evolution of the gap size, δ, measured in strain units (relative to the inner clad radius) is obtained from the difference of the dimensional change of the two surfaces (inner clad surface and outer pellet surface) defining the gap. Therefore, the rate of relative gap size change is given by dδ/dt=ƒ(σ)−g(t), which is equivalent to, ⅆ δ ⅆ t = f ( σ ) - a 0 P - a 1 ⅆ P ⅆ t ( 4 ) The initial gap size, δ0, needed to solve Eq. (4) is required as a model input, which is provided by fuel manufacturer as a fuel rod characteristic.The differential equations (3) and (4), for tracking the main state variables identified as clad stress and pellet-clad gap size respectively, constitute the main structure of the XEDOR system. All the physics in the XEDOR model is contained in a0, a1, and ƒ; these are fitted to quantitatively approximate the behavior of a detailed fuel performance code. The present reduced order model is fitted to AREVA's fuel performance code RODEX4 and is applicable for use with other fuel performance codes. The basic features of the parameters are outlined below. The coefficient a0 is generally dependent on burnup. Its functional form allows the model to represent the fuel swelling process as well as the short-term densification resulting from the reduction of unstable pores that exist in the fuel pellet ceramic material. The long-term densification process at the expense of stable pores, also known as accommodation, is also represented. An example of the parameter a0 is given below:a0=2.6×10−8−10−5pue−B/2.3−10−6pse−B/25+6×10−9[exp(P/50)−1] (5)where B is the burnup in MWd/kgU, pu is the fraction of unstable pores in the ceramic pellet material, and ps is the fraction of stable pores in the same material. The coefficient, a1, which is related to the thermal expansion of the fuel pellet, is generally dependent on burnup and the pellet temperature, which in turn depends on the linear power level, P. The value of the coefficient a1 increases with both burnup and power in order to account for the degradation of fuel thermal conductivity. Thus, the temperature response to a given increase in power, and consequently the thermal expansion response, is larger for higher burnup and higher initial linear power level. An example of the parameter a1 is given below:a1=[1.3×10−5+5.7×10−5P+1.6B]eP/50+3×10−5 (6) The stress relaxation in the clad is directly linked to the inelastic strain rate given by the function ƒ(σ). Cladding inelastic strain rate is a strongly nonlinear function of stress, and is induced by the high energy portion of the neutron flux, also called fast flux, Φ, and increases with temperature. A so-called thermally induced creep component occurs under stress without requiring a fast neutron field, which is distinguishable from the so-called irradiation-induced creep that requires both stress and fast neutron flux. The form of the clad creep function allows for the so-called primary creep stage where the creep rate is initially large, followed by secondary creep stage which occurs at a reduced, almost constant, rate under the same stress. This transition from primary to secondary creep is attributed to the so-called strain hardening, where the thermally induced creep produces an effect on the material microstructure retarding further creep. An example for the clad creep function is given below asƒ(σ)=cT sin h(sσ)+cIRRσvΦw (7)where the first term accounts for the thermal-induced creep and the second term accounts for the irradiation-induced creep. The thermal creep coefficient, cT, generally increases with clad temperature and decreases with the accumulated exposure to fast neutrons (fast fluence) and decreases with the accumulated strain. The irradiation induced component of the creep rate increases with stress and fast neutron flux. The set of coefficients, (cT, s, cIRR, v, w) determine the magnitude of the creep rate function. The creep rate function, ƒ(σ), is written explicitly as a function of clad stress because the latter is a major state variable in the model, the definition of the creep rate function implies its dependence on fast neutron flux. It also accounts for cladding material anisotropy and is specific to a particular type of zirconium alloy and heat treatment. The model parameters, or set of coefficients, defining a creep rate function are therefore fitted for application to a given clad material and heat treatment, and the resulting creep function is applicable to any irradiation environment including boiling and pressurized water reactors. Alternative methods for providing the coefficients and material property functions, a0, a1, and ƒ(σ), include tabulated data obtained from a combination of experimental measurements and fuel performance code results for specific fuel rod materials. Such tabulated data can be used directly or in a numerically fitted form. The representation of stress as the true physical parameter responsible for PCI fuel failure is not the only method of providing such information to the reactor operator. An alternative representation of the margin that exists before PCI failure is based on calculating the maximum allowable linear power at which the maximum allowable stress is reached. The margin parameter based on the difference, or ratio, between the maximum allowable and the present linear powers can be a preferable one as it given in power units and therefore directly linked to the power maneuver. However, it must be mentioned that while a maximum allowable stress is a single value for each fuel type, the maximum allowable linear power varies with time for each rod segment according to its operational history. The stress margin and the linear power margin are therefore equivalent in providing the operational limits on power maneuvering. Fuel Conditioning and De-Conditioning The concept of conditioning is deeply embedded in the paradigm of fuel behavior and its response to power change. A benefit of this reduced order model is to provide a rigorous interpretation of the conditioning process leading to a well-defined conditioning state. This new conditioning state is beneficial as it provides for conceptual continuity of understanding bridging the gap between the empirical prior art and the physical modeling of this invention as embodied in XEDOR. The new conditioning state can be also used to define more precise maneuvering rules superior to the empirical rules of the prior art, but not better than a full implementation of XEDOR, thus serving as an intermediate step that is easy to implement with minimal change to existing reactor monitoring software and also easy for reactor operator as minimal, if any, training is needed. Conditioning a segment of a fuel rod by operating at this conditioning level LHGR for a sufficiently long time implies that a state of mechanical equilibrium is assumed to exist under these conditions where hoop stress reaches an asymptotic value, and the inelastic strain (creep) rate becomes constant. Mathematically, the properties of the new time-dependent conditioned state, Π(t), satisfy the following conditions: 1. The conditioned state, Π(t), is attracted to the linear power, P(t). 2. The conditioning state, Π(t), is a continuous function of time, in contrast with P(t), which is not necessarily continuous. 3. Operating under linear power below the conditioning state is always safe. Given that a stress threshold, σ*, in the approximate range of 200-400 MPa, exists for causing fuel failure, the conditioning state satisfies the inequalityσ(t)<σ* for P(t)<Π(t) (8) The property given in Equation (8) is behind the main interest in conditioning state. Specifically, given that a conservative stress threshold is known or implied from experience and testing of a particular cladding material, then a maximum allowable linear power can be obtained from the conditioning state plus an overshoot or a tolerance to allow the stress to increase from the relatively low value associated with the conditioning state to the maximum safe stress threshold. Thus, the relationship governing the restriction of the local linear power, P, is given byP≦Π(t)+ΔP (9)which differs from the prior art in the important aspect that the empirical conditioned state, PCS, is replaced by the new conditioning state, Π, where the latter is defined rigorously as presented below. A conditioning state satisfying the properties mentioned above is introduced in the context of the new reduced order model of this invention, XEDOR, by replacing P(t) with Π(t) in Eq. (3) while dropping the stress derivative term because the stress presumably reached a constant signifying mechanical equilibrium. Thus, a 1 ⅆ Π ⅆ t + a 0 Π = f ( σ ) ( 10 ) where the source term is the clad inelastic strain rate driven by the time-dependent stress, σ(t), which is obtained by integrating the XEDOR system of equations. For the conditioning state equation to be integrated, an initial conditioning state is required, which is given an appropriate value depending on fuel type in the approximate range of 35-45 kW/m. A precise value of the initial conditioning is not important as its effect will decay in time. Inspecting Eq. (10), we note that a linear power increase results in increasing stress and subsequently increasing the strain rate and leads to a continuous increase of Π. This increase of the conditioning function to catch up with an increase in linear power is precisely the XEDOR-specific meaning of “conditioning,” while the opposite direction is termed “deconditioning.” It should be noted that the equation for calculating the new conditioning state, Π, applies only when the pellet-clad gap is closed. For the condition when the gap is open, the conditioning state is defined as the linear power required to cause pellet thermal expansion sufficient to close the gap and apply pressure on the clad to reach a specific given conditioning stress, the later being of the approximate range of 30-100 MPa. The open gap conditioning state is not a limiting factor in restricting operation, but it serves as an initial condition for the critical stage in fuel operation when the gap becomes closed. It should be noted that the mechanical equilibrium that was assumed is not strictly attainable under constant linear power operation, due to the change in the material properties mainly due to the hardening that occurs with accumulated creep. This illustrates why the application of the conditioning state, Π, albeit obviously superior to the prior art as basis for maneuvering, a less accurate method compared with direct use of XEDOR calculated stress as basis for PCI protection during reactor power maneuvering. The XEDOR model differential equations (3) and (4) are used directly to evaluate power maneuvers and predict safe power trajectory without the need of any conditioning concept. The conditioning parameter, as rigorously defined and contained in the variable Π, is a novel construct to be utilized for providing continuity for the human operator of the nuclear power plant upon transition from an empirical maneuvering guide to the physically based reduced order model. Modes of XEDOR Applications The flow chart of FIG. 1 shows the embodiment of the invention as coupled to the reactor core instrumentation and monitoring system to provide input to the XEDOR system where gap size and stresses are calculated as the main variables, and shows the XEDOR output displays for the reactor operator who in turn affects the reactor power maneuver through the reactor controls within the margins displayed by XEDOR. The flow chart of FIG. 2 shows the embodiment of the invention as coupled to the reactor core instrumentation and monitoring system to provide input to the XEDOR system where the conditioning state is calculated as the main variable in addition to gap size and stress, and shows the XEDOR output displays for the reactor operator who in turn affects the reactor power maneuver through the reactor controls within the margins displayed by XEDOR. The flow chart of FIG. 3 shows the application of the invention as a computer code XEDOR provided with input representing operational data that was previously stored, e.g. by the reactor monitoring software. The result of the calculations performed by XEDOR as a diagnostic tool is used by an analyst for the purpose of identifying the fuel assemblies or rods where the PCI margins represented as stress or linear power may have been violated. This is particularly helpful in speeding up the process of identifying and removing a failed fuel assembly to restore reactor operation to normal as soon as possible. The flow chart of FIG. 4 shows the application of the invention as a computer code XEDOR provided with input representing anticipated operational data for a reactor core loading in the design stage. The result of the calculations performed by XEDOR as a design optimization tool is used by an analyst for the purpose of identifying the fuel assemblies or rods where the PCI margins represented as stress or linear power may become violated. If such violation, or unacceptably small margins, were found, the analyst is enabled to revise the fuel loading pattern until an acceptable pattern is obtained with sufficient PCI margin during anticipated operation. An example of a hypothetical control rod withdrawal error is presented here to demonstrate the XEDOR model in comparison with RODEX4 and to validate the accuracy of the novel reduced order model compared with large fuel performance code. FIG. 5 depicts the evolution of the size of the pellet-clad gap at a selected elevation of a particular fuel rod. The calculation demonstrates good comparison between the two methods, and shows the important features of an initially increasing gap due to pellet densification followed by gradual closure of the gap. Gap size variations due to pellet thermal expansion following power changes can be discerned. FIG. 6 shows the hoop stress calculated by RODEX4 and XEDOR, where a large spike is calculated due to the sudden power increase resulting from a postulated control rod withdrawal error. FIG. 7 is a zoom of FIG. 6 around the power (and stress) spike time to show the details of the stress evolution calculated by the two methods. The high stress (>500 MPa for RODEX4 and >600 MPa for XEDOR) constitute a severe violation of the safe operation threshold and fuel failure would be expected. This is demonstrated in FIG. 8 where the operating linear power is shown to exceed the maximum allowable linear power (corresponding to a maximum allowable hoop stress) as calculated by both methods. Monitoring a BWR with XEDOR The example shown above demonstrates one aspect of the monitoring function that can be performed with XEDOR, mainly tracking the maximum allowable linear power compared with the operating linear power in order to ensure that a safety margin exists. The other important function is projecting a safe power trajectory in order to plan maneuvers such as start up and control rod sequence exchange. The projected power trajectory is divided into a power jump followed by a continuous power ramp. These are calculated as follows. Safe Power Jump The XEDOR state variable set is available up to the point where a safe power trajectory projection is requested, i.e. the gap and hoop stress are known. The safe power jump is then calculated as the power corresponding to thermal expansion that consumes the present gap (if any exists) plus clad elastic strain corresponding to a given safe stress setpoint, also called threshold stress or maximum allowable stress, σ*. In this way, a large power jump would be predicted if the initial state is defined by an open gap, while a smaller power jump would be predicted if the initial state is defined by a closed gap, and even smaller if a large hoop stress is initially present. The magnitude of the predicted power jump is also dependent on the initial power and burnup, as high power (i.e. high temperature) and burnup result in pellet thermal conductivity degradation and thus increasing the thermal expansion strain response to power increase. Safe Power Ramp The pellet-clad gap is always closed following a power jump to a predetermined safe stress level, σ*. The subsequent allowed safe power increase must be limited to a ramp rate where the stress increase due to pellet thermal expansion is balanced by the stress relief due to creep in the cladding material. This is calculated by reversing Eq. (3) in the sense of using power as the independent variable, while specifying stress as a constant. With the stress derivative vanishing, Eq. (3) is rewritten as a 1 ⅆ P ⅆ t + a 0 P = f ( σ * ) ( 11 ) The safe ramp rate, R=dP/dt, can thus be calculated fromR=(ƒ(σ*)−a0P)/a1 (12) Equation (11) is integrated numerically by the XEDOR computer code whenever a safe power projection over a period of time is requested by the reactor operator. Equation (12) provides an instantaneous value of the maximum allowed ramp rate which is applicable over a short period of time. Following a series of power increase steps following the ramp rate provided by equation (12) results in linear power increase after each step which reduced the allowed ramp rate for the subsequent step; this procedure is equivalent to numerical integration of equation (11) to obtain an projected linear power curve bounding allowed power maneuvers. An example of a power jump followed by a projected safe power ramp is given in FIG. 9. The plot shows the power jump and the subsequent ramp function, in addition to the maximum allowable power which corresponds to a higher stress than the projected power in order to keep a safety margin. The conditioning power is also shown to rise in response to the attracting operating linear power function. It is important to notice that the projected safe linear power runs nearly parallel to the conditioning state power, illustrating the applicability of the maneuvering method according to equation (9) where the projected safe linear power is larger than the conditioning state power by a nearly fixed quantity, ΔP. Structure of Reactor Monitoring Software for PCI Protection The structure of the power maneuvering function using XEDOR as coupled to a core monitoring system includes the following elements as a preferred embodiment: 1. A computer based core monitoring system capable of reading reactor control state such as position of control rods, flow rate, and total core thermal power. 2. A software module capable of simulating the detailed core state such as three-dimensional distributions of neutron flux and power. The software is capable of calculating such parameters as fast neutron flux and linear heat generation rate for every fuel rod and at different elevations. The software is capable of integrating fast neutron flux to get the fast neutron fluence, and integrating power density to get burnup, for each rod at several elevations. 3. A computer module to embody the algorithmic representation of XEDOR, namely equations (3) and (4), capable of integrating same equations to obtain the state variables representing stress and gap size for each rod at different elevations. For this state variable integration, the initial conditions are the gap size representative of the fuel rod as manufactured, and the stress level imposed by the external coolant pressure. The driving parameters are linear heat generation rate and fast flux provided by the host core monitoring system. The integrated parameters of the fast fluence and burnup are also provided by the host core monitoring software. 4. A display of the stress level distribution in the core, and the associated margin to maximum allowable LHGR. The reactor operator utilizes this margin information to decide on whether a planned power maneuver is safe. 5. A display of the maximum allowable rate of increasing LHGR to guide the reactor operator action regarding power ascension by flow control or control rod motion. The functions performed by the XEDOR system include: 1. Using the data provided online by a nuclear plant monitoring and instrumentation system, calculate the stress distribution in many or all fuel rods in the reactor core at several axial locations. 2. Calculate a parameter, for each fuel rod segment representing an axial location, to identify the margin to reaching an operator provided stress limit. Such parameter can be the linear power required to reach the stress limit, or the linear power increase above the existing level at the time required to reach the stress limit, or the ratio between the maximum linear power associated with the stress limit and the existing linear power at the time, or other convenient parameter to provide the operator with information regarding the operation margin to reaching the stress limit. 3. Project in a predictive manner the time evolution of the linear power at each rod segment in the core designated for such calculation conditional on stress remaining at or below an operator provided value. 4. Perform one or more of the above functions listed in items 1 through 3 online, i.e. while directly connected to the reactor instrumentation and monitoring software. 5. Perform the stress survey and related functions listed in items 1 through 3 offline using stored data obtained from reactor instrumentation and monitoring systems. 6. Perform the function of item 5 for the purpose of identifying a failed fuel assembly and further identifying the location of a failed fuel rod in the fuel assembly. 7. Perform the stress survey and related functions using design data for the purpose of optimizing the loading pattern of different fuel assemblies in a reactor core, and the associated planned power maneuvers. The latter include control rod sequences for boiling water reactors, and reactivity management with soluble boron and control rod motion for pressurized water reactors.A Summary Disclosed here is a preferred embodiment of a reduced order model for calculating stress and pellet-clad gap size in at least one fuel rod segment to provide information for core loading pattern design, operator power maneuvering and or diagnostics of suspected PCI cladding failure comprising and calculating a gap size evolution from its initial as-manufactured value by ⅆ δ ⅆ t = f ( σ ) - a 0 P - a 1 ⅆ P ⅆ t ( 4 ) and, and calculating an evolution of clad hoop stress with time upon gap size reaching zero by 1 E ⅆ σ ⅆ t + f ( σ ) = a 0 P + a 1 ⅆ P ⅆ t ( 3 ) where t is time, δ is the pellet-clad gap size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is the clad inelastic strain rate as function of hoop stress while being irradiated with a given fast neutron flux level and having accumulated a fast neutron fluence, specific for a given clad material and heat treatment, P is the linear heat generation rate in pellets of the fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power and, displaying the calculated parameters, σ and/or δ, for operator use in power maneuvering. The reduced order is further explained by noting that the parameter a0 is determined bya0=2.6×10−8−10−5pue−B/2.3−10−6pse−B/25+6×10−9[exp(P/50)−1], (5),where B is the burnup in MWd/kgU, P is the linear power in kW/m, pu is the fraction of unstable pores in the ceramic pellet material, and ps is the fraction of stable pores in the same material. Further, the parameter a1 is determined bya1=[1.3×10−5+5.7×10−5P+1.6B]eP/50+3×10−5, (6)and is generally dependent on burnup and linear power level and additionally where the clad creep rate function ƒ(σ) is determined byƒ(σ)=cT sin h(sσ)+cIRRσvΦw, (7)and is generally dependent on stress and fast neutron flux, Φ, with the coefficient set cT, s, cIRR, v, w being specific to clad material and heat treatment. Additionally it is seen that the reduced order model may further comprise the parameters a0 and a1 and function ƒ(σ) being in tabulated form or fitted by curve fitting means to a combination of experimental measurements of pellet and clad material properties and the results of large scale fuel performance codes. Seen in this disclosure are a number of expressions of the invention. Seen is a method further comprising tracking the maximum allowable linear power compared with the operating linear power in order to ensure that a safety margin to PCI fuel failure exists; projecting a safe power trajectory in order to plan maneuvers such as start up and control rod sequence exchange. Further is seen the step of dividing the projected safe power trajectory into a safe power jump followed by a continuous safe power ramp and predicting a safe power jump by determining the gap and hoop stress when a safe power trajectory projection is requested; calculating the safe power jump as the power corresponding to thermal expansion that consumes the present gap plus clad elastic strain corresponding to a given safe stress setpoint. An additional step in the method is predicting an instantaneous power ramp rate where the stress increase due to pellet thermal expansion is balanced by the stress relief due to creep in the cladding material, the power ramp rate being limited toR=(ƒ(σ*)−a0P)/a1, (12),and then predicting a continuous power ramp [0040] by limiting the power increase rate such that the stress increase due to pellet thermal expansion is balanced by the stress relief due to creep in the cladding material, wherewith the linear power is calculated from a 1 ⅆ P ⅆ t + a 0 P = f ( σ * ) , ( 11 ) which is integrated numerically by the XEDOR computer code whenever a safe power projection over a period of time is requested by the reactor operator. Thereafter is the step of displaying the predicted safe power jump and power ramp for operator consideration in power maneuvering. As additional disclosure of this invention we see the reduced order model of the preferred embodiment where said calculated parameters are for operator use in power maneuvering in either a Boiling Water Reactor or a Pressurized Water Reactor and said calculated parameters, σ and δ, are displayed for operator use in power maneuvering. An alternative approach to the invention is seen in the method of using reduced order models for fast calculation of cladding stress in the majority of the fuel rods at different elevations of a reactor core for the purpose of calculating margin to and protecting against PCI failures comprising with a step of inputting into a computer a computer program implementing the algorithms for calculating the gap size evolution from its initial as-manufactured value by ⅆ δ ⅆ t = f ( σ ) - a 0 P - a 1 ⅆ P ⅆ t , ( 4 ) and, calculating the evolution of clad hoop stress with time upon gap size reaching zero by 1 E ⅆ σ ⅆ t + f ( σ ) = a 0 P + a 1 ⅆ P ⅆ t , ( 3 ) where t is time, δ is the pellet-clad size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is the clad inelastic strain rate as function of hoop stress for given fast neutron flux level and cumulative fast neutron fluence, specific for a given clad material and heat treatment, P is the linear heat generation rate in pellets of the fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power. Additionally, inputting, into the computer, core instrumentation measurements, using said measurements to provide input to core monitoring software, accessing the monitoring software generated operational data of linear heat generation rate, fast flux, burnup, fast neutron fluence, for all or the majority of the fuel rods at several axial locations and calculating, by the computer program, the gap size and clad stress when the gap is closed. Thereafter, outputting, to an operators display, the stress response to a planned power increase, as operational guidance for power maneuvering. Additionally there may be seen the steps of calculating the linear heat generation rate maximum allowable limit based on the calculated stress and outputting same for operational guidance, calculating the margin to reaching the maximum allowable linear heat generation rate and outputting same for operational guidance and calculating the margin in terms of total reactor thermal power to any rod segment reaching the maximum allowable linear heat generation rate and outputting same for operational guidance. Additionally seen, as further consideration of an alternative embodiment, is that the core instrumentation measurements and monitoring software may be from a Boiling Water Reactor or from a Pressurized Water Reactor. An additional view, seen in an alternative embodiment, is a reduced order model to reduce PCI failure in power maneuvering comprising the steps of inputting into a computer a reduced order model XEDOR wherein algorithmic representations, in a computer code, receive data from core instrumentation and monitoring software programs thereby permitting online application for predictive calculations which are performed and yielding calculations prior to executing any power maneuver in a reactor. Further will be seen the steps of displaying said calculations in physical operator displays for operator consideration where calculations are physically based, unlike empirical maneuvering guides, and providing a qualitatively higher degree of accuracy and reliability for power maneuvering safety. And additionally the steps of demonstrating, via the operator display, calculations which are predictive of the stress level and the associated linear heat generation rate margin to the maximum allowable limits using definite representation of each fuel design parameters and material properties. This additional alternative embodiment of the reduced order model of may further comprise the algorithms presented in the computer code and inputted into a computer are 1 E ⅆ σ ⅆ t + f ( σ ) = a 0 P + a 1 ⅆ P ⅆ t , and ( 3 ) ⅆ δ ⅆ t = f ( σ ) - a 0 P - a 1 ⅆ P ⅆ t , ( 4 ) where,t is time, δ is the pellet-clad gap size relative to clad inner radius, σ is clad hoop stress, E is clad material Young's modulus, the function ƒ is the clad inelastic strain rate as function of hoop stress while being irradiated with a given fast neutron flux level and having accumulated a fast neutron fluence, specific for a given clad material and heat treatment, P is the linear heat generation rate in pellets of the fuel rod as provided by reactor instrumentation and monitoring software, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power and, where the initial gap size, δ0, is a manufacturing parameter, the coefficients a0 and a1 are not necessarily constant, and the creep rate function, ƒ, is not solely dependent on stress but also on the creep itself and the fast neutron fluence. As additional consideration for such an alternative embodiment we see that the parameter a0 is determined bya0=2.6×10−8−10−5pue−B/2.3−10−6pse−B/25+6×10−9[exp(P/50)−1], (5)where B is the burnup in MWd/kgU, P is the linear power in kW/m, pu is the fraction of unstable pores in the ceramic pellet material, and ps is the fraction of stable pores in the same material and where the coefficient, a1 is determined bya1=[1.3×10−5+5.7×10−5P+1.6B]eP/50+3×10−5 (6)and is generally dependent on burnup and linear power level. Additionally, the stress relaxation in the clad is directly linked to the inelastic strain rate given by the functionƒ(σ)=cT sin h(sσ)+cIRRσvΦw (7)and is generally dependent on stress, σ, and fast flux, Φ, with the coefficient set cT, s, cIRR, v, w being specific to clad material and heat treatment. This alternative embodiment may further receive core instrumentation measurements and monitoring software from a Boiling Water Reactor or from a Pressurized Water Reactor. Another alternative embodiment of the invention is described as a core monitoring system where XEDOR algorithms are encoded in a computer program for the purpose of calculating stress in fuel rod segments prior to performing reactor power maneuvers comprising the steps of inputting a computer program encoding XEDOR algorithms into a computer, of measuring, via core instrumentation, core operational data required for core monitoring software to generate detailed operational parameters for each fuel rod in the core at a plurality of axial locations or segments in the form of computer data arrays; the XEDOR algorithms accepting as input the data arrays representing the fuel rod segment data and of receiving calculations, from the XEDOR manipulation of the input data, and outputting the calculations to a display for operator consideration in power maneuvering. Yet another alternative embodiment of the invention is expressed as a computer having a computer program comprising a computer-readable storage medium having computer program logic stored thereon for enabling a processor to execute a XEDOR algorithm, upon receipt of operational data from a reactor core monitoring system, calculate margin to PCI failure, said calculations displayed for reactor operator use for executing a safe power maneuver. An additional alternative embodiment is noted as a computer with the XEDOR computer program, performing the steps of: receiving operational data from a reactor core monitoring system; calculating stress level in cladding of at least one fuel rod and the associated linear power margin to the maximum allowable limits, using definite representation of each fuel type design parameters and material properties; displaying said calculations for operator use to execute a safe PCI failure free power maneuver such that the said limits are not exceeded. Yet another alternative embodiment is noted as the method of power maneuvering using XEDOR as coupled to a core monitoring system comprising using the data provided online by a nuclear plant monitoring and instrumentation system; calculating the stress distribution in many or all fuel rods in the reactor core at several axial locations, calculating a margin parameter, for each fuel rod segment representing an axial location, to identify the margin to reaching an operator provided stress limit where such parameter can be the linear power required to reach the stress limit, or the linear power increase above the existing level at the time required to reach the stress limit, or the ratio between the maximum linear power associated with the stress limit and the existing linear power at the time, or other parameters which provide the operator with information regarding the operation margin to reaching the stress limit, projecting, in a predictive manner, the time evolution of the linear power at each rod segment in the core designated for such calculation conditional on stress remaining at or below an operator provided value, performing one or more of the above functions online while directly connected to the reactor instrumentation and monitoring software and displaying the margin parameters to reaching an operator provided stress limit for operator consideration in power maneuvering. This alternative embodiment may further comprise performing a stress survey and related functions offline using stored data obtained from reactor instrumentation and monitoring systems, performing the stress survey and related functions for the purpose of identifying a failed fuel assembly and further identifying the location of a failed fuel rod in the fuel assembly, performing the stress survey and related functions using design data for the purpose of optimizing the loading pattern of different fuel assemblies in a reactor core, and the associated planned control rod sequences for boiling water reactors, and reactivity management with soluble boron and control rod motion for pressurized water reactors, and displaying the results of the stress survey for designers consideration and benefit to identify and distinguish design patterns with respect to their respective propensity to PCI related fuel rod failures. Another alternative embodiment is seen as a method for power maneuvering guidance of nuclear reactors comprising calculating a fuel conditioning state, Π, as function of time, t, by a 1 ⅆ Π ⅆ t + a 0 Π = f ( σ ) , ( 10 ) where the function ƒ is the clad inelastic strain rate driven by the time-dependent hoop stress, σ, while being irradiated with a given fast neutron flux level and having accumulated a fast neutron fluence, specific for a given clad material and heat treatment, a0 is pellet strain rate per unit linear power, and a1 is pellet strain per unit linear power and, allowing the linear power at a fuel rod segment, P, to vary during a reactor power maneuver with the constraints that it does not exceed the corresponding conditioning state plus a prescribed margin, ΔP, such thatP≦Π+ΔP, (9) This alternative embodiment may further comprise obtaining the stress driving the conditioning state calculation from the solution of the XEDOR system equations. This alternative embodiment may further comprise entering the algorithms for calculating the conditioning state in a computer program connected to the reactor instrumentation and monitoring software and displaying the linear power margin, Π+ΔP−P, calculated by the program for guiding power maneuvering of a boiling water reactor. Additionally, this alternative embodiment may comprise displaying the core thermal power required for at least one fuel rod segment reaching the maximum linear power, Pmax=Π+ΔP, for guiding the power maneuvering of a pressurized water reactor. While a preferred embodiment of the present invention has been shown and described, it will be apparent to those skilled in the art that many changes and modifications may be made without departing from the invention in its broader aspects. The appended claims are therefore intended to cover all such changes and modifications as fall within the true spirit and scope of the invention. |
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description | This application claims priority to U.S. Provisional Application No. 62/825,506, filed Mar. 28, 2019, the entire disclosure of which is incorporated by reference herein. The presently-disclosed invention relates generally to systems for raising and lowering control rods in nuclear reactors and, more specifically, to control rod drive mechanisms that include drive motors that function independently of the latch coils. As is known in the art, various types of nuclear reactors utilize control rods to regulate reactivity levels and, therefore, the operation of the reactors by raising and lowering the control rods within the respective reactor cores, as required by power demands. Control rod drive mechanisms are used to raise and lower the control rods. An example control rod drive mechanism may include a pair of segment arms that are pivotably attached to a corresponding torque tube so that the upper ends of the segment arms are disposed within a central bore of a stator of a control rod drive motor, and the lower ends of the segment arms each include a roller nut for selectively engaging a threaded lead screw from which one or more control rods are suspended. During power operations, energizing the stator causes the upper ends of the segment arms to pivot radially-outwardly due to an induced magnetic field in the stator so that the bottom ends of the segment arms pivot radially-inwardly, causing the roller nuts to engage, or latch, the lead screw. Subsequent rotation of the segment arms is achieved by stepping the drive motor, i.e., changing the phase or the motor's stator. Whether the control rods are raised or lowered is dependent upon the direction of rotation of the induced magnetic field, as determined by the stator. The control rods can be rapidly dropped, or scrammed, into the reactor core in order to shut down the reactor core. This is achieved by securing power to the stator, thereby preventing the stator from producing the induced magnet field. Although the above described control rod drive mechanisms have been found to operate reliably for extended periods of time, there are various disadvantages that may exist. Specifically, typical control rod drive motors include a single stator coil that performs latch, hold, and drive operations. As such, the single stator coil, which has a single power source, provides both the induced rotating magnetic field for rotating the segment arms, as well as the roller nut clamping force which pulls the upper portions of the segment arms radially-outwardly so that the roller nuts engage the lead screw during latch and hold operations. As such, the motor's drive torque and roller nut clamping force are dependent variables, as shown in FIG. 6. As shown, peak rotational torque may not occur at the same rotor lag angle as the peak roller nut clamping torque, which means the roller nuts may partially open at peak motor torque if the clamping force is not adequate. Additionally, referring now to FIG. 7, in known control rod drive motors 20 the pressure boundary tube 26 located between the segment arms 24 and the stator 28 decreases the motor's efficiency due to leakage flux. The reduced efficiency requires higher current to the stator, resulting in increased temperatures within the stator's core and increased copper losses due to increased resistance. The increased temperatures and high copper losses create significant issues with regard to operating in high temperature environments when not providing external cooling for known control rod drive motors having a single stator with a single power source. There at least remains a need, therefore, for a system for controlling drive rods in a nuclear reactor that allows maximization of the drive torque independently of the latching force. One embodiment of the present disclosure provides a control rod drive mechanism for use in a nuclear reactor including a reactor core disposed in a pressure vessel, including a control rod configured for insertion into the reactor core, a lead screw having a top end and a bottom end, the control rod being secured to the bottom end of the lead screw, a drive mechanism including a torque tube having a top end and a bottom end, a pair of segment arms, each segment arm including a top end and a bottom end and being pivotably mounted to the torque tube, a pair of roller nuts, each roller nut being rotatably secured to the bottom end of a respective segment arm, and a drive motor including a stator and a rotor secured to the top end of the torque tube that includes a plurality of permanent magnets embedded therein, wherein the stator defines a central bore and the rotor is disposed within the central bore of the stator, and a latch coil assembly including a latch coil, wherein the latch coil assembly defines a central bore and the top ends of the segment arms are disposed within the central bore of the latch coil assembly radially-inwardly of the latch coil. An alternate embodiment of the present disclosure provides a control rod drive mechanism for use in a nuclear reactor including a reactor core disposed in a pressure vessel, including a control rod configured for insertion into the reactor core, a lead screw having a top end and a bottom end, the control rod being secured to the bottom end of the lead screw, and a drive mechanism including a torque tube having a top end and a bottom end, a pair of segment arms, each segment arm including a top end and a bottom end and being pivotably mounted to the torque tube, a pair of roller nuts, each roller nut being rotatably secured to the bottom end of a respective segment arm, and a drive motor including a stator and a rotor secured to the top end of the torque tube that includes a plurality of permanent magnets embedded therein, wherein the stator defines a central bore and the rotor is disposed within the central bore of the stator. Yet another embodiment of the present disclosure provides a control rod drive mechanism for use in a nuclear reactor including a reactor core disposed in a pressure vessel, including a control rod configured for insertion into the reactor core, a lead screw having a top end and a bottom end, the control rod being secured to the bottom end of the lead screw, a drive mechanism including a torque tube having a top end and a bottom end, a pair of segment arms, each segment arm including a top end and a bottom end and being pivotably mounted to the torque tube, a pair of roller nuts, each roller nut being rotatably secured to the bottom end of a respective segment arm, and a drive motor including a stator and a rotor that is secured to the top end of the torque tube, wherein the stator defines a central bore and the rotor is disposed within the central bore of the stator, and a latch coil assembly including a latch coil and a hold coil, wherein the latch coil assembly defines a central bore and the top ends of the segment arms are disposed within the central bore of the latch coil assembly radially-inwardly of the latch coil and the hold coil. The accompanying drawings, which are incorporated in and constitute a part of this specification, illustrate one or more embodiments of the invention and, together with the description, serve to explain the principles of the invention. Repeat use of reference characters in the present specification and drawings is intended to represent same or analogous features or elements of the invention according to the disclosure. The invention now will be described more fully hereinafter with reference to the accompanying drawings, in which some, but not all, embodiments of the invention are shown. Indeed, this invention may be embodied in many different forms and should not be construed as limited to the embodiments set forth herein; rather, these embodiments are provided so that this disclosure will satisfy applicable legal requirements. As used in the specification, and in the appended claims, the singular forms “a”, “an”, “the”, include plural referents unless the context clearly dictates otherwise. Referring now to the figures, as shown in FIG. 1, an embodiment of a control rod drive mechanism 100 in accordance with the present invention, a cylindrical torque tube 122, a drive mechanism 110 (FIGS. 2A and 2B) including a pair of segmented arms 132 that are pivotally secured to torque tube 122, a drive motor 112 including a stator 114 and a rotor 118 affixed to torque tube 122, wherein rotor 118 is disposed within a central bore 116 defined by stator 114, and a latch coil assembly 152 disposed radially-outwardly of the top ends of segment arms 132, latch coil assembly 152 including a hold coil 154 and a latch coil 156 disposed within a magnetic sleeve 164. Referring additionally to FIGS. 2A and 2B, each segment arm 132 includes a pair of threaded roller nuts 140 rotatably secured to its bottom end so that roller nuts 140 may be selectively engaged and disengaged from a corresponding lead screw 142 that is received within a bore 130 defined by a body 128 of torque tube 122, as discussed in greater detail below. In alternate embodiments, more than two segment arms 132 may be used and each segment arm 132 may include more than two roller nuts 140. Note, although not shown in FIG. 1, torque tube 122, rotor 118 of driver motor 112, segment arms 132, and roller nuts 140 are disposed within the interior of a pressure tube 140 that forms a portion of the pressure boundary of the corresponding nuclear reactor. The electric components noted above, those being stator 114 of drive motor 112 and latch coil assembly 152, are disposed on the outer surface of pressure tube 150, meaning they are not subject to the internal conditions of the nuclear reactor. Referring again to FIGS. 2A and 2B, the present embodiment of control rod drive mechanism 100 includes an independent stator for each of the latch, hold and drive operations. Specifically, stator 114 of drive motor is used to create a rotating magnetic field which induces rotor 118 to rotate, thereby causing torque tube 122 to rotate in the desired direction, latch coil 156 is energized during latching operations, thereby urging top ends 134 of segment arms 132 radially-outwardly about their pivot points 138 so that roller nuts 140 are urged radially-inwardly to engage lead screw 142, and hold coil 154 maintains top ends 134 of segment arms in the radially outward position (FIG. 2B) during normal reactor operations. Note, it may be preferable during latching operations to energize both latch coil 156 and hold coil 154 and, upon verifying proper engagement of roller nuts 140 with lead screw 142, securing power to latch coil 156 once normal reactor operations commence. Securing power to latch coil 156 with the exception of during latching operations helps reduce the internal temperatures of the coils, thereby leading to greater efficiency as compared to known control rod drive mechanisms. As shown, hold coil 154 is disposed above latch coil 156, radially-outwardly of the uppermost portions of segment arms 132 in order to maximize the amount of leverage and, therefore, clamping force that hold coil 154 exerts on roller nuts 140 by way of segment arms 132 pivoting about the corresponding pivot points 138. Preferably, hold coil 154 and latch coil 156 include horizontal windings, although they may be wound vertically as well. Referring now to FIG. 3, hold coil 154 and latch coil 156 are housed within magnetic sleeve 164. As shown, magnetic sleeve 164 includes an upper flux rib 158, a middle flux rib 160 and a lower flux rib 162. Upper flux rib 158 is adjacent to the upper end of hold coil 154, lower flux rib 162 is disposed adjacent the lower end of latch coil 156, and middle flux rib 160 is disposed between the bottom end of hold coil 154 and the upper end of latch coil 156. Flux ribs 158, 160 and 162 facilitate the flow of magnetic flux about hold coil 154 and latch coil 156 without over-saturating since the ribs conduct flux from the outer surface of magnetic sleeve 164 to the top ends 134 of segment arms 132. Referring now to FIG. 2A, FIG. 2B and FIG. 4, drive motor 112 of control rod mechanism 110 includes a slotted stator 114 that includes vertical windings 115, preferably of ceramic insulated magnetic wire, and rotor 118 that includes a plurality of permanent magnets 120 such as, but not limited to, samarium cobalt (Sm2Co17) magnets 120 embedded therein about is outer perimeter. Magnets 120 on rotor 118 improve the performance of drive motor 112 as compared to existing drive motors (FIG. 7) in that magnets 120 provide a permanent magnetic field on rotor 118, meaning stator 114 does not have to induce a magnetic field in the rotor during drive operations. As such, power requirements for drive motor 112 are reduced as compared to similar drive torque requirements in existing drive motors. Reduced power loads means reduced internal temperatures, few copper losses and, therefore, increased efficiency. As such, the presently discussed embodiment of control rod drive mechanism 100 may either handle similar loads as existing control rod drive mechanisms while operating more efficiently at lower internal operating temperatures of stator 114, or the present embodiment may operate at similar internal operating conditions yet handle higher torque loads due to the greater operating efficiency of drive motor 112. Note, in alternate embodiments, the rotor of the drive motor may be formed by embedding the permanent magnets 120 directly to the top ends of the corresponding segment arms 132. In this type of embodiment, the hold coil 154 functions in the dual role of being the stator of the drive motor. As noted above, control rod drive mechanism 100 utilizes separate coils for each of the latch, hold and drive operations. As such, drive torque of drive motor 112 and roller nut clamping force of latch coil assembly 152 are not dependent upon each other, and the design of each coil can be optimized for its intended function without adversely affecting the operation of the other coils for their intended functions. As shown in FIG. 5, the clamping force of hold coil 154 and/or latch coil 156 is constant and not affected by the rotor angle of drive motor 112. Therefore, the drive motor 112 may be operated at all rotor lag angles without adversely affecting the roller nut clamping force provided by hold coil 154 and/or latch 156. Note also, unlike existing control rod drive mechanisms, the ability to secure power to latch coil 156 during normal reactor operations further increases the efficiency of control drive mechanism 100 as internal temperatures are reduced. Furthermore, the power to the hold coil 154 can be varied to accommodate the drive torque required. These and other modifications and variations to the invention may be practiced by those of ordinary skill in the art without departing from the spirit and scope of the invention, which is more particularly set forth in the appended claims. In addition, it should be understood that aspects of the various embodiments may be interchanged in whole or in part. Furthermore, those of ordinary skill in the art will appreciate that the foregoing description is by way of example only, and it is not intended to limit the invention as further described in such appended claims. Therefore, the spirit and scope of the appended claims should not be limited to the exemplary description of the versions contained herein. |
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040381382 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The fuel element 1 shown in FIG. 1 comprises uranium carbide enclosed within a stainless steel sheath. The fuel element has a wire wrap 2 which serves to space the fuel element from adjacent fuel elements 1 in a bundle of fuel elements enclosed by a peripheral wrapper (not shown). The wire wrap 2 comprises a single wire helically wound about its longitudinal axis in the manner of an open coil helical spring. The helically wound wire is wound in a helical path about the fuel element so that it makes contact along a helical path in a series of regularly spaced intermittent contacts. The diameter of the wire is approximately 50 percent of the coil diameter. A wire wrap of this form enables coolant to flow between the wire wrap and the fuel element between the points of contact. Such a wrap also has sufficient transverse stiffness to maintain constant the spacing of fuel elements in the bundle under normal operation. The outside diameter of the fuel element is 9.14 mm, the outside diameter of the coils of wire is 2.5 mm and the diameter of the wire 1 mm. In the construction of fuel element 1 shown in FIG. 2, a wire wrap 3 comprises two wires of equal diameter interwound in the manner of twin lead flex used for electric lighting installations. The combination of interwound wires is wound helically about the fuel element to contact it in a series of regularly spaced intermittent contacts. The twin wire wrap is wound in a clockwise direction about the fuel element and the wires are wound clockwise about each other. Alternatively, the twin wire wrap may be wound in an anticlockwise direction about the fuel element and the wires wound anticlockwise about each other. By winding the wires about each other and the wound combination about the fuel element in the same direction, nesting of fuel elements one with another in the bundle is improved and the spacing of fuel elements is more uniform. In a third construction of fuel element (which is generally similar to the combination shown in FIG. 2) the wire wrap comprises a relatively large diameter single wire having a smaller diameter wire helically wound about it. The combination of wires is wound helically about the fuel element so that the coils of the small wire make contact with the fuel element along a helical path in a series of regularly spaced intermittent contacts. Wires having diameters in the ratio 2:1 are suitable, for example, in a wire wrap for effecting 0.1 inch spacing between fuel elements the larger wire is approximately 0.050 inch diameter whilst the smaller wire is approximately 0.025 inch diameter. |
051788250 | summary | This invention relates to fuel bundles for boiling water nuclear reactors utilizing the prior art suggested concept of tapered fuel rods having relatively large diameter bottom tubes and relatively smaller diameter top tubes placed in the fuel rod array within nuclear fuel bundles. More particularly, a nuclear fuel bundle and improved upper fuel rod spacer for use within the fuel bundle is disclosed which enables tapered fuel rod removal from the top and replacement from the top of the fuel bundle. SUMMARY OF THE PRIOR ART Ueda Japanese Kokai Patent No. Sho 52-50498 of Apr. 22, 1977 discloses a fuel bundle having partial length fuel rods. These partial length fuel rods are arrayed is side-by-side relation so as to define large central voids in the center of fuel bundles within a boiling water reactor. The large central void defined by the partial length rods is in the upper two phase region of the fuel bundle and has the shape of a cone. In a first embodiment of Ueda, the cone in the upper portion of the fuel bundle is filled with a correspondingly large central conical water rod. In a second--and apparently preferred embodiment of Ueda--the large central conical void is vacant. This large central conical void permits the discharge of steam within the fuel bundle. The purpose of the design of Ueda is to provide a kind of fuel assembly that can suppress "output peak", that is the tendency of portions of the fuel rods in the upper two phase region adjacent the control rod to heat to levels approaching the thermal limit of the fuel rods. The function of the above mentioned disclosed designs "is characterized by the feature that the volumetric ratio of themmoderator (water) to fuel is made to change in the axial direction in the channel box, so that the above-mentioned peak caused by the distribution of axial voids can be inhibited." The Ueda disclosure continues. It points out that in addition to the conical voids, other expedients can be used. Specifically the suggestion is made as follows: "In addition, similar effects as above can be achieved by using fuel rods with changed diameter in the axial direction, i.e., larger outer diameter of the fuel rods in the lower portion of the fuel assembly, and small diameter of the fuel rods in the upper portion of the fuel assembly." No practical design of the suggested fuel rods is disclosed. SUMMARY OF THE INVENTION In a nuclear fuel bundle for a boiling water reactor, the prior art suggested concept of tapered fuel rods forming the bundle array is disclosed in a practical design. This design includes a fuel bundle and an improved associated spacer. A tapered fuel rod is utilized having a large diameter bottom tube for holding corresponding large diameter fuel pellets at the fuel rod bottom, a smaller diameter top tube for holding correspondingly smaller diameter pellets at the top of the fuel rod, and a bell reducer situated between the large diameter lower fuel tube and the smaller diameter upper fuel tube. This bell reducer tapers from the large outside diameter of the lower fuel tube to the smaller outside fuel diameter of the upper fuel tube and forms a smooth transition between the large and small diameter sections of the fuel rod. Current spacer designs can easily be adapted to the upper small diameter portion of the fuel rods, and to the lower large diameter portions of the fuel rods, resulting in one spacer design for the upper portion of the fuel bundle and a different spacer design for the lower portion of the fuel bundle. However, a major problem arises with such a fuel bundle; the large diameter portion of the fuel rods cannot pass through the upper spacers. Current practice in fuel bundle assembly is to insert fuel rods into the bundle from the top. Insertion of fuel rods from the bottom of the bundle would greatly complicate the fuel bundle assembly. In addition, it is sometimes necessary to replace fuel rods in fuel bundle after operation in a nuclear reactor. Current practice is to remove and insert from the top of the fuel bundle. Insertion from the bottom would require major changes in reactor servicing equipment, would be a more complex procedure, and would require more time. Improved spacer designs are disclosed which allow insertion and removal of tapered fuel rods. An improved fuel rod spacer is disclosed for use with the upper and smaller tube portions of the fuel rods. This upper spacer defines a matrix of rod holding positions having sufficient dimension for enabling both the upper smaller diameter tube and the lower large diameter tube to pass through the cells defined for each fuel rod. A spring having a relatively wide flexure range protrudes into each cell. When the fuel bundle is conventionally assembled, the spring in each cell biases the upper smaller diameter tubes in their deigned side-by-side relation against stops defined at each cell. At the same time, the spring is provided with sufficient flexibility to allow the lower large diameter tube and the transition bell reducer to pass through the individual cells of the upper spacer. With the improved upper spacer, a fuel rod can be lifted vertically upward and out of the fuel bundle or replaced vertically downward into the fuel bundle. In the upper spacers the stops are smaller than in a conventional spacer. The reduced stop dimension provides additional space in the cell, allowing the larger diameter portion of the fuel rod to pass through. In the assembled condition with the small diameter portion of the fuel rod in the cell, and with the fuel rod biased against the stops by the spring, the fuel rod is off-center with respect to the cell. All of the cells and stops in the spacer have the same orientation, so that a uniform array of fuel rods results. Two embodiments are disclosed for the lower spacers. In the preferred embodiment, the cells are identical to those of the upper spacers. When the large diameter portion of the fuel rod is biased against these stops, the fuel rod is off-center with respect to the cell, in the direction opposite to that of the upper portion of the fuel rod in the upper spacers. All of the cells and stops of the lower spacers have the same orientation, resulting in a uniform array of fuel rods. In an alternate embodiment, the stop dimensions are different from those of the upper spacers, and are chosen to center the fuel rods in the cells. There results a fuel bundle which has a uniform distribution of fuel and coolant with resultant uniform critical power response, no regions of non uniform vapor flow, and improved flow area in the upper two phase region of the bundle. OTHER OBJECTS, FEATURES AND ADVANTAGES An object of this invention is to disclose a practical design for a fuel bundle utilizing an array of vertically upstanding tapered fuel rods. The prior art suggested concept of a tapered fuel rod is provided with an upper smaller diameter portion, a lower larger diameter portion, and a reducer providing a smooth transition there between. An improved spacer is disclosed. The spacer--like all prior art spacers--defines a plurality of individual cells with stops for the bias of the fuel rods to a designed limit of side-by-side spacing. The spacer--unlike all prior art spacers--has special characteristics allowing the removal and replacement of the tapered fuel rods from the top of the fuel bundle. Accordingly, each cell of the spacers in the upper portion of the fuel bundle is of sufficient dimension to permit withdrawal of the lower large diameter portion of the fuel rods. At the same time, springs within each of the cells of the upper spacers have two discrete design functions. In the first design function, the springs act to bias the upper smaller diameter portions of the fuel rods to their designed side-by-side spacing. In a second design function, the springs deflect sufficiently to permit serial passage of the reducer and large diameter portion of the fuel rod. An advantage of the disclosed fuel bundle and improved spacer is that the fuel bundle thus permits conventional removal and replacement of the individual fuel rods from the top of the fuel bundle without requiring inversion of the fuel bundle. A further object of the invention is to disclose a cell arrangement which preserves a uniform rod to rod spacing while each fuel rod center is offset from the center of the cell surrounding it. In this arrangement of cells the orientation of spring and stops is the same in all cells. The direction and magnitude of the fuel rod center displacement relative to the cell center is the same for all cells. In this arrangement each spring can act on only one fuel rod. The prior art for ferrule spacers where one spring acts on two adjacent fuel rods cannot be used. A further object of this invention is to disclose a spacer format for maintaining both the upper and lower portions of the tapered rods in perfect verticality. According to this aspect of the invention, the orientation of the cell stops and springs is the same in the upper and in the lower spacers. In the upper spacers all the fuel rods are displaced relative to the cells by a constant amount, and are displaced in the same direction. There is a uniform array of cell centers and a uniform array of rod centers, and the rod array is offset relative to the cell array. In one embodiment of the lower spacers, the cell center array is coincident with the array of fuel rod centers. In another embodiment, the arrays are offset. In current practice the spacer is located in the bundle channel, and is centered in the channel. If this practice were followed with the spacers disclosed here, the fuel rod centers in the upper spacers would be offset relative to the fuel rod centers in the lower spacers. In the invention disclosed here, the upper and lower spacers are offset from their centered positions so the array of fuel rod centers is centered in the channel. In either embodiment, the cells on the lower spacers are offset with respect to the cells of the upper spacers so that the centers of the upper portions of the fuel rods lie directly above the centers of the lower portions of the fuel rods. There results a spacer system that maintains top to bottom designed verticality of the tapered rods within the fuel bundle. An advantage of this aspect of the invention is that the disclosed spacer displacement scheme can be utilized for all spacer designs having cell stops identically arrayed. Accordingly, the disclosed scheme will work with either spacers having Zircaloy ferrules or lighter Inconel grids. |
description | This invention relates to radioactive shields used with particle accelerators, and more particularly, to a mold for forming a radioactive shield component and for shielding radioactivity. Positron Emission Tomography (PET) is a procedure used for imaging and measuring physiologic processes within the human body. As part of the procedure, radioisotopes are injected into a patient to assist in diagnosing and assessing a disease. A radioisotope, such as Fluorine-18, may be produced through the use of a particle accelerator. In particular, the particle accelerator produces radioisotopes by accelerating a particle beam and bombarding a target material, housed in a target system, with the particle beam. Referring to FIGS. 1a and 1b, a general configuration for a particle accelerator 10 is shown. The particle accelerator 10 generates a particle beam 12 within movable concrete shields 14 and 16 and stationary concrete shields 18 and 20. The particle beam 12 then bombards target material 11 inside target enclosure 22 to produce a radioactive isotope. The resulting decay of the isotope as well as other interactions generate gamma rays and neutron particles 13 that are absorbed by the concrete shields 14, 16, 18,20 to protect any person located in areas outside of the concrete shields 14, 16, 18,20. High density materials are used to shield a particle accelerator in order to reduce the energy from accelerated particles in the shortest distance possible. The effectiveness of a material in reducing energy or slowing down particles varies with different types of particles and their energy level. Concrete is typically used for shielding many types of radiation including gamma rays and neutron particles and is used extensively in the nuclear industry. In particular, the concrete shields 14, 16, 18,20 are made of concrete formed through use of a steel mold that is supported by reinforcement ribs. Additional shielding materials may also be mounted to the concrete shields 14, 16, 18,20 to improve the performance of the concrete shields. Reinforced steel and wood forms are typically used to mold the concrete used in stationary shield components such as walls, floors, and ceilings. In use, the shield components are typically covered with commercial materials such as sheet rock and paint to make their appearance more cosmetically acceptable. However, such commercial materials are labor intensive to use and lack durability. In low energy cyclotrons and linear particle accelerators moveable concrete shields may be used in a self-shielded configuration. In these applications individual concrete pieces are also formed using steel molds. However, the molding process produces parts with poor surface finishes that require a great deal of manual labor to repair in order to produce a cosmetically acceptable surface. In addition, the resulting surface is brittle and is easily damaged when the steel mold is removed from the concrete and during the manufacturing, assembly, shipping and installation process. This requires constant repair of the damaged surfaces with the use of common filler materials, sheet rock mud and other repair materials in order to maintain the cosmetic appearance of the surfaces. However, such repair materials do not provide the same strength and shielding properties as concrete. Concrete also has other structural limitations. Concrete has excellent compressive strength but is very weak in tension. These tensile strength limitations require that all concrete structural members be reinforced with steel reinforcement members 38 such as steel rebar in areas where the concrete structural members experience tensile loads. Use of steel to provide reinforcement is an undesirable feature in concrete radiation shields. The steel reinforcement members 38 become radioactive when acted upon by neutron particles 13 produced from the interaction of particle accelerator beam 12 on target material 11 inside target enclosure 22 whereas the surrounding concrete material does not. Thus, the reinforcement members 38 require additional shielding in the concrete shield. In addition, the steel reinforcement members 38, when radioactive, pose increased disposal cost to the customer when the particle accelerator is decommissioned at its end of life. Further, the steel molds are subjected to frequent flexing, bending and movement during the fabrication of the concrete shields which quickly fatigues and damages the steel molds. As a result, the steel molds require frequent maintenance which increases costs. Moreover, the maintenance is labor intensive which further increases costs. The manufacturing process for the concrete shields 14, 16, 18,20 takes place at the manufacturers' facility. The concrete shields 14, 16, 18,20 are then shipped to a customer location. The steel molds, on the other hand, are kept at the manufacturers' facility because their size and weight make them cost prohibitive to ship. Therefore, new techniques are needed for casting shield components that reduce costs, are less labor intensive and result in shields that are cosmetically acceptable, provide enhanced structural characteristics and are durable and functional. A shield for absorbing radiation emitted during generation of a radioisotope is disclosed. The shield includes an inner portion fabricated from a first type of shielding material. The shield also includes an outer portion fabricated from a second type of shielding material wherein the outer portion serves as a mold for forming the inner portion. The inner portion may be fabricated from a material which shields against gamma rays such as concrete. The outer portion may be fabricated from a material which moderates neutrons such as high density polyethylene. In the case of concrete shields, the material is structurally reinforced using a structural member that does not become radioactive in radiation fields. Before any embodiments of the invention are explained in detail, it is to be understood that the invention is not limited in its application to the details of construction and the arrangement of components set forth in the following description or illustrated in the following drawings. The invention is capable of other embodiments and of being practiced or of being carried out in various ways. Also, it is to be understood that the phraseology and terminology used herein is for the purpose of description and should not be regarded as limiting. The use of “including,” “comprising,” or “having” and variations thereof herein is meant to encompass the items listed thereafter and equivalents thereof as well as additional items. Unless specified or limited otherwise, the terms “mounted,” “connected,” “supported,” and “coupled” and variations thereof are used broadly and encompass direct and indirect mountings, connections, supports, and couplings. Further, “connected” and “coupled” are not restricted to physical or mechanical connections or couplings. In the description below, like reference numerals and labels are used to describe the same, similar or corresponding parts in the several views of FIGS. 1-2. Referring to FIG. 2, a portion of a shield 24 configured in accordance with the present invention is shown. The shield 24 includes a polyethylene portion 26 located on an outer periphery of an inner concrete portion 28. Polyethylene is a good neutron moderator and reflector and thus the polyethylene portion 26 serves as a shielding component along with the concrete portion 28 in the shield 24. In accordance with the present invention, the polyethylene portion 26 also serves as a mold for forming the concrete portion 28 of the shield 24. The concrete portion 28 includes a curved surface 32 for accommodating a portion of a particle accelerator. The particle accelerator is used to generate radioisotopes by accelerating a particle beam and bombarding a target material, housed in a target system, with the particle beam. The shield 24 may be configured for use in a self-shielding particle accelerator arrangement that uses a low energy particle accelerator. In this arrangement, the shield 24 may be configured to be moveable. The polyethylene portion 26 may be formed by an injection molding or a blow molding process. In a preferred embodiment, the polyethylene portion 26 is formed by rotation molding (i.e. “rotomolding”). The injection molding, blow molding and rotomolding processes all provide continuous walls in the polyethylene portion 26 that are suitable for molding the concrete portion 28. In addition, the thickness of polyethylene portion 26 may be varied as desired by using either process. Polyethylene is an inexpensive material with many desirable properties suitable for molding concrete. In particular, polyethylene is cosmetic, durable, repairable, readily colored to meet customer requirements, has good strength properties in addition to being a neutron moderator and reflector. In one embodiment, ribs 36 are used to provide support for the polyethylene portion 26. The polyethylene portion 26 may also be blended with other materials suitable for increasing the material shielding properties without significantly affecting the structural properties of the material. In addition, shielding components 30 and 40 may be embedded in the concrete portion 28 to improve the structural or attenuation properties of the shield 24. In a preferred embodiment, the shielding components 30 may be fabricated from lead or polyethylene and structural components 40 may be fabricated from fiberglass. Alternatively, the shielding components 30 and structural components 40 may be fabricated from steel. Use of polyethylene in the concrete portion 28 and fiberglass structural components 40 have the added benefit that the amount of radioactive shielding that has to be disposed is reduced when the system is decommissioned. Additional shielding elements 34 may be located adjacent to surface 32 to provide further shielding capability. The polyethylene portion 26 is much lighter than conventional steel molds. This enables shipment of the polyethylene portion 26 to the customer. The concrete portion 28 is then fabricated at or near the installation site thus substantially reducing costs as compared to shipping a conventional concrete shield. In addition, costs for post processing to repair damaged surfaces are also substantially reduced. Moreover, the supplier is then able to provide customized mold arrangements for fabricating customized shields 24 suitable for the customer. In a preferred embodiment the polyethylene portion 26 is fabricated from polyethylene such as high density polyethylene (HDPE), low-density polyethylene (LPDE) and linear low-density polyethylene (LLPDE). However, polyvinyl chloride (PVC), fluorocarbons, polypropylene, nylon and polycarbonate or other blends or combinations thereof may also be used. The polyethylene portion 26 may be repaired using unprocessed HDPE powder and simple heating tools such as soldering irons and butane torches. While the invention has been described in conjunction with specific embodiments, it is evident that many alternatives, modifications, permutations and variations will become apparent to those skilled in the art in light of the foregoing description. Accordingly, it is intended that the present invention embrace all such alternatives, modifications and variations. |
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abstract | An X-ray generator for generating plasma and X-ray emitted from the plasma includes a unit for generating the plasma, and plural reflection optical systems for introducing the X-ray through different optical paths. |
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summary | ||
claims | 1. A method of converting fusion product energies into electric power, comprising the steps of injecting ions along a helical path within a tapered cylindrical cavity formed by first and second tapered hemi-cylindrical electrodes in spaced relation with first and second elongate gaps formed there between, and converting at least a portion of the ion energy into electrical energy. 2. The method of claim 1 , further comprising the step of applying an oscillating potential to the first and second electrodes. claim 1 3. The method of claim 2 , further comprising the step of creating an azimuthal electric field across the first and second gaps. claim 2 4. The method of claim 3 , further comprising the step of decelerating the ions. claim 3 5. The method of claim 4 , wherein the injecting step includes converting substantially all of the ions axial energy to rotational energy. claim 4 6. The method of claim 5 , wherein the ions are injected in the form of an annular beam. claim 5 7. The method of claim 6 , further comprising the step of directing the annular beam through a magnetic cusp. claim 6 8. The method of claim 7 , further comprising the step of collecting charge neutralizing electrons from the annular beam as the electrons follow magnetic field lines of the magnetic cusp. claim 7 9. The method of claim 8 further comprising the step of collecting the ions once a substantial portion of their energy is converted to electric energy. claim 8 10. The method of claim 9 further comprising the step of conditioning the electric energy converted from the ion energy to match existing power grids. claim 9 11. The method of claim 7 further comprising the step of creating the magnetic cusp. claim 7 12. The method of claim 11 further comprising the steps of creating first and second magnetic fields within the cavity, wherein field lines of the first and second magnetic fields extend in opposing directions, and joining the first and second magnetic fields. claim 11 13. A method of converting fusion product energies into electric power, comprising the steps of injecting ions along a helical path within a tapered cylindrical cavity formed by first and second tapered hemi-cylindrical electrodes in spaced relation with first and second elongate gaps formed there between, converting substantially all of the ions axial energy to rotational energy, and converting at least a portion of the ion energy into electrical energy. 14. The method of claim 13 , further comprising the step of applying an oscillating potential to the first and second electrodes. claim 13 15. The method of claim 14 , further comprising the step of creating an azimuthal electric field across the first and second gaps. claim 14 16. The method of claim 15 , further comprising the step of decelerating the ions. claim 15 17. The method of claim 13 further comprising the step of creating a magnetic cusp. claim 13 18. The method of claim 17 further comprising the steps of creating first and second magnetic fields within the cavity, wherein field lines of the first and second magnetic fields extend in opposing directions, and joining the field lines of the first and second magnetic fields. claim 17 19. The method of claim 18 , wherein the ions are injected in the form of an annular beam. claim 18 20. The method of claim 19 , further comprising the step of directing the annular beam through a magnetic cusp. claim 19 21. The method of claim 20 , further comprising the step of collecting charge neutralizing electrons from the annular beam as the electrons follow magnetic field lines of the magnetic cusp. claim 20 22. The method of claim 16 further comprising the step of collecting the ions once a substantial portion of their energy is converted to electric energy. claim 16 23. The method of claim 16 further comprising the step of conditioning the electric energy converted from the ion energy to match existing power grids. claim 16 24. A method of converting fusion product energies into electric power, comprising the steps of creating first and second magnetic fields within a cavity formed in part by first and second tapered hemi-cylindrical electrodes in spaced relation with first and second elongate gaps formed there between, wherein field lines of the first and second magnetic fields extend in opposing directions, joining the field lines of the first and second magnetic fields to form a magnetic cusp, injecting ions in the form of an annular beam along a helical path within the cavity, directing the annular beam through the magnetic cusp, and converting at least a portion of the ion energy into electrical energy. 25. The method of claim 24 , further comprising the step of applying an oscillating potential to the first and second electrodes. claim 24 26. The method of claim 24 , further comprising the step of creating an azimuthal electric field across the first and second gaps. claim 24 27. The method of claim 26 , further comprising the step of decelerating the ions. claim 26 28. The method of claim 24 , wherein the injecting step includes converting substantially all of the ions axial energy to rotational energy. claim 24 29. The method of claim 24 , further comprising the step of collecting charge neutralizing electrons from the annular beam as the electrons follow magnetic field lines of the magnetic cusp. claim 24 30. The method of claim 27 further comprising the step of collecting the ions once a substantial portion of their energy is converted to electric energy. claim 27 31. The method of claim 27 further comprising the step of conditioning the electric energy converted from the ion energy to match existing power grids. claim 27 |
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abstract | A probe is provided for an SPM (Scanning Probe Microscope), and a method is provided for fabricating the probe in which a double side alignment process is not required to simplify the fabricating. The probe includes a cantilever; a body supporting the cantilever; and a tip formed at an end of the cantilever, wherein the cantilever, the body and the tip are made of silicon, and boron is diffused into the cantilever and a predetermined area of the body. The method includes steps of: forming a first mask layer on an area of a silicon substrate to be formed with the body and the tip; etching the silicon substrate in a predetermined depth using the first mask layer to form the tip; removing the first mask and forming a second mask layer on an area of the silicon substrate except for an area to be formed with the body and the cantilever; forming a boron-diffused layer by diffusing boron into an area to be formed with the cantilever and a predetermined area of the body using the second mask; removing the second mask layer and forming a third mask layer on the boron-diffused layer; and etching the silicon substrate using the third mask layer to form the body and the cantilever. |
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summary | ||
description | 1. Field of the Invention This invention pertains generally to control rod drive mechanism seismic supports for nuclear power plants and more particularly to a quick disconnect seismic support tie rod system. 2. Description of the Related Art In conventional reactors, the head package includes the pressure vessel head which seals the reactor vessel, control rod drive mechanisms which are used to raise and lower control rods in the core of the reactor, a seismic platform adjacent the upper ends of the control rod drive mechanisms, which laterally restrains the drive mechanisms, and various cables for operation of the control rod drive mechanisms. A missile shield, which conventionally was formed of a concrete slab, is positioned above the head package to protect the containment housing and associated equipment from penetration by any of the control rod drive mechanisms in the event of a major pipe break. The problems associated with such conventional head packages are more fully described in U.S. Pat. No. 4,678,623, issued Jul. 7, 1987, and assigned to the assignee of this invention. In such conventional plants, the large concrete slabs installed above the reactor vessel to act as a missile shield must be removed and stored prior to head disassembly and refueling of the reactor, and then must be replaced after the refueling and head reassembly. Such operations affect overall refueling time and radiation exposure and require space in the containment area for placement of the missile shield slabs when removed from the position above the reactor vessel. In order to reduce the refueling time, personal exposure and space requirements, an improved system, designated as an integrated head package was developed which incorporates an integral missile shield and head lift rig. The missile shield is in the form of a perforated circular plate which is directly attached to a head lift rig. Such an integral head package system is described in U.S. Pat. No. 4,830,814, issued May 16, 1989 and assigned to the assignee of this invention. As described therein, and illustrated in FIG. 1 of the present drawings, an integral head package 10 includes a three-legged head lifting rig 12 that is pin connected at 14 by lift lugs 16, to a missile shield assembly 18. The perforated circular plate 20 that forms the missile shield 18 acts as a spreader for the head lift load, and as a seismic support for the tops of the control rod drive mechanisms 22, with rod travel housings extensions 24 of the control rod drive mechanisms protruding through apertures 26 in the circular plate 20. The missile shield 18 interfaces with the tops of the control rod travel housings 22 which limits the overall vertical travel (and impact force) of a missile before it impacts the shield. The impact load of the missile against the underside of the perforated plate 20 is transmitted to head lift rods 28, through vessel head lift lugs 30 secured to the vessel head 32, and closure studs 34 to the vessel head 32, and ultimately to the vessel supports. A cooling shroud 36 surrounds the control rod drive mechanisms 22, while electric cabling 38 is routed from the top of the control rod drive mechanisms 22 to a connector plate 40 and then along a cable tray 42 to respective cable terminations. Cooling fans 44 circulate air within the shroud 36 to transfer waste heat from the control rod drive mechanisms 22. Hoist supports 46, and trolleys 48 on hoist assemblies 50 are used to position stud tensioner tools and stud removal tools during refueling operations. The integrated head package and variants of the design which have since evolved, were a marked improvement over conventional head package designs, and are adaptable for retrofitting existing reactors or for incorporation into new reactor designs as will be described hereafter. However, there is still room for improvement in reducing the number of steps that have to be performed in the critical path of a refueling outage. For example, many nuclear power plants have control rod drive mechanism seismic support tie rods. Typically, there are five to six tie rods which are pinned at the reactor head assembly attachment, and pinned at a refueling cavity wall mounted anchor. During plant refueling, the tie rods need to be removed in order to move the vessel head assembly to the head storage stand. The ends of the tie rods that are attached to the wall mounted anchors (and in most cases the head assembly mounted ends also) are disconnected and reconnected by operators in a man basket supported by the overhead polar crane. Because these activities utilize the polar crane, which is also required for numerous refueling activities, they are considered to be in the critical path of the refueling outage. Any reduction in critical path time results in significant savings in the form of the refueling schedule and electric utility dollars. Accordingly, a new tie rod support system is desired that can reduce the number of steps required to disconnect the tie rod wall anchors from the vessel head so that the vessel head can be removed. Furthermore, a new tie rod support system is desired that can remove the disconnection of the tie rods from the refueling outage critical path. Additionally, such a system is desired that can enable the vessel head assembly to be removed from the vessel with the tie rods attached to the head assembly. These and other objects are achieved by the control rod drive mechanism seismic support tie rod system of this invention that has the tie rod ends that engage anchors on the vessel cavity walls and are connected to the anchors with a locking mechanism that is directly operable from a location remote from the anchor to lock or unlock the tie rod from engagement with the anchor. In one preferred form, the locking mechanism is latchable in a locked position to lock the tie rod into engagement with the anchor. In a second position, the locking mechanism is preferably latchable in an unlocked position to maintain the locking member in an open state so that the tie rod can be removed. In one embodiment, the end of the tie rod that engages the anchor includes a lateral extension that extends from the tie rod end in a first direction and is engaged by a pivotable hook on the locking mechanism when the locking mechanism is in a locked position. In the foregoing embodiment, the pivotable hook has a distal end that is spaced from a pivot coupling on the locking mechanism. The distal end is pivotably connected to an actuation arm which is operable from the location remote from the anchor to lock or unlock the end of the tie rod. Preferably, in the foregoing embodiment, a stationary arm, over which the actuation arm rides, has a plurality of holes along a length thereof, at least one of which mates with a corresponding hole in the actuation arm when the actuation arm moves the locking mechanism into the locked position and into the unlocked position. In still another embodiment, the end of the tie rod that engages the wall anchor includes a second lateral extension that extends from the tie rod end in a second direction that is opposite the first direction and is engaged by a clevis on the locking mechanism. Preferably, the nuclear containment facility employing this invention includes an operating deck within the vicinity of the vessel cavity. A generally vertical oriented wall extends down from the operating deck into the cavity, opposed from at least a portion of the reactor vessel, on which the tie rod anchors are mounted. The anchors are preferably secured at or between a foot to two feet (30.5-61 cm.) below the operating deck. Preferably, one end of the tie rods are pivotably connected to the reactor vessel so that the tie rods can pivot up into a generally vertical position when they are released from the anchors. Desirably, a winching system is located on the reactor vessel head assembly for raising and lowering the tie rods. In still another embodiment wherein the anchor includes a plate that extends generally toward the reactor vessel, substantially in line with the corresponding tie rod that it connects with, the anchor has a first dowel portion extending laterally from one side of the plate and a second dowel portion extending from an opposite side of the plate. The adjacent end of the tie rod is formed to slip over and engage the first and second dowel portions from above when the tie rods are pivoted down at the reactor vessel head assembly. The plate has a locking bar that is operable from a remote location on an operating deck of the containment to move over the tie rod end when the tie rod end fully engages the first and second dowel portions to lock the tie rod end to the anchor. Preferably, the tie rod end is configured as a fork with two tines spaced to receive the plate therebetween. Each of the tines has a downward facing clevis opening that receives a corresponding dowel portion as the tie rod is pivoted downward over the plate. Desirably, a locking bar is operable from the remote location on the operating deck to rotate between an open position wherein the locking bar lies over the top of the plate, clear of the tines of the tie rod end and a closed position, approximately 90 degrees from the open position, where the locking bar lies over each of the tines. The locking bar is preferably rotated by a long-handled tool that has a forked lower end that grips an upwardly extending wall on the locking bar. Desirably, the locking bar can be latched in either or both the open or closed position. Preferably, the clevis opening in the tie rod end has a taper at a lower end of at least one wall of the clevis opening slot to facilitate alignment over the dowel pin portions. In still another embodiment, the anchor includes an engagement interface which is angularly adjustable to align with the corresponding tie rod. Preferably, the angular adjustment is in the vertical orientation. FIG. 2 shows a perspective view of the portion of the operating deck 52 which surrounds the reactor vessel cavity 54 and a portion of the integrated head package 10 showing a missile shield 18 seated over the control rod travel housing extensions 24 and coupled to the lifting rig 12 that was previously described with respect to FIG. 1. The integrated head package also includes a seismic ring 56 that is secured around the control rod travel housing extensions 24 and is secured against lateral movement by the tie rods 58 and 60 which are connected between the seismic ring 56 and the vessel cavity wall 62. As can be seen from plan view in FIG. 3 of this prior art arrangement, six tie rods, four radial tie rods 60 and two tangential tie rods 58 secure the control rod travel housing extensions 24 from lateral movement. The tie rods are connected at one end 72 to the seismic ring 56 on the integrated head package 10 through a pinned coupling 70 and are connected at another end 74 to the anchor 64 secured to the vessel cavity wall 62 through a similar connection 66. As previously mentioned, during plant refueling, these tie rods 58 and 60 have to be removed in order to move the vessel head assembly 10 to the head storage stand. The end of the tie rods 74 that is attached to the wall mounted anchor 64 (and in some cases the head assembly mounted end 72 also) is disconnected and reconnected by operators in a man basket supported by the overhead polar crane. The quick disconnect control rod drive mechanism seismic support tie rod system of this invention eliminates the need to use the overhead polar crane for this purpose and thus takes this task outside of the critical path. One embodiment of the quick disconnect control rod drive mechanism seismic support tie rod system of this invention is illustrated in FIG. 4 and includes a winching system 76 attached to the head assembly 10 which is used to raise and lower the tie rods 58 and 60 without the use of a man basket. The tie rods 58 and 60 are secured to the head assembly structure through the pivot connection 70 and remain with the head assembly 10 for the move to the head storage stand. In this embodiment, a wall anchor attachment 64 includes a slotted clevis 90 which is designed to receive a laterally extending dowel 92 on the forward end 74 of the tie rod 58, 60. The dowel 92 spans between two spaced circular brackets 94 that are supported at the end of the tie rod 58, 60 with the circular brackets 94 fitting on either side of the clevis 90 when the dowel 92 is seated in the slot of the clevis. A locking mechanism (more clearly shown in FIGS. 7, 8 and 9), is actuated from the operating deck 52 and is designed to prevent the tie rods 58, 60 from becoming displaced from the clevis 90 during a seismic or pipe break activity. Referring more specifically to FIG. 4, it can be seen that the winch system 76 is supported on a post 86 that is mounted on the seismic ring 56. A winch crank 88 is mounted on the post 86 and has a cable 78 that extends from the post 86 to a pulley system 80 which is connected to the forward end 70 of the tie rod 58, 60 through a forward linkage 82. The winch cable 78 extends from the post 86 to the pulley system 80 and back and around guide wheels 84 on the post 86 with the end of the cable connected to the crank 88 so that when the crank is turned in a direction to draw in the cable 78 the tie rod 58, 60 is raised toward the vertical axis lifting the dowel out of the clevis, assuming the locking mechanism 96 is in an open position. Conversely, when the crank 88 is turned in a direction to let out the winch cable 78, the tie rods 58, 60 is lowered to be received in the slotted clevis 90 when the plant is in a cold start-up condition. A winching system 76 is provided for each tie rod 58, 60. Preferably, the winch system 76 includes a lock, such as on the crank 88 that will lock the winch cable 78 in position when the tie rods 58, 60 are in their fully withdrawn position so the tie rods can be removed with the reactor vessel head to the head stand. The design of the locking mechanism 96 of this invention is more fully illustrated in FIGS. 6-9. To appreciate the improvement of this invention, it is helpful to first understand the prior art coupling between the end of the tie rods 74 and the anchor 64 illustrated in FIG. 5. The anchor plate 64 of the prior art has two spaced, parallel plates that extend orthogonally from a base plate which is affixed to the reactor cavity wall 62. The spaced parallel plates 100 have aligned holes through which a dowel 92 passes. The design of the end 74 of the tie rod is very similar to that of the current invention shown in FIGS. 6 and 7 in which the end is formed from a split yoke that is designed to receive at least one of the parallel plates therebetween. The split yoke 102 of the prior art, shown in FIG. 5, has an enlarged circular rounded end 94 on each of the fork tine terminations of the split yoke. The rounded ends 94 have a central opening through which the dowel 92 passes and secures the split yoke 102 to the spaced parallel plate 100. The dowel is affixed on one side of the parallel plates 100 with an enlarged end and on the other side with a cotter pin 98. One side of the anchor plate 64 of the embodiment of this invention described above, is illustrated in the side view shown in FIG. 6. The anchor plate 64 which is attached to the reactor cavity wall 62 has a slotted clevis 90 that extends orthogonally into the reactor well. The slotted opening 104 in the clevis 90 is designed to receive the dowel 92 on the tie rod end 74 between the two rounded ends 94 of the split yoke 102 of the tie rod end 74. A better view of the locking mechanism 96 of the anchor assembly 64 of the foregoing embodiment is shown in FIG. 7. As previously stated, the tie rod end 74 is substantially similar to the prior art tie rod end except the dowel 92 extends laterally from the rounded end 94 of the split yoke 102 in a direction away from the slotted opening 104 in the clevis 90 and is captured by the locking mechanism 96 in the closed position as will be more fully explained hereafter. The dowel 92 in this embodiment can be permanently affixed to both rounded ends 94 of the split yoke tie rod end 102. Alternately, one of the tines of the split fork could be removed and the dowel 92 could extend out of one or both sides of the rounded end 94 so long as the dowel 92 was captured within the clevis slot 104 and the locking mechanism 96 as explained hereafter. The locking mechanism 96 includes a pivotable hook 106 that is attached to a base spacer member 108 at a pivot point 110. The base spacer member 108 is desirably connected to both the base anchor plate 64 and the slotted clevis 90 and is sized to capture the rounded end 94 of at least one tine of the split yoke 102 of the tie rod end 74 between the pivotable hook 106 and the clevis 90. The distal end 112 of the pivotable hook 106 is connected to the end of an actuation arm 114 through a second pivot point 116. The actuation arm 114 extends from the pivot point 116 vertically to a height above the operating deck 52 where the actuation arm 114 terminates in a horizontal handle 118 that extends over the operating deck 52. The operating deck is typically one and one-half feet to two feet (45.72-60.96 cm) above the wall mounted anchor 64. The actuation arm 114 rides over a stationary arm 120 that extends along the anchor wall plate 64, along side the slotted clevis 90. The stationary arm has a locking pin hole 122 which mates with corresponding holes 122 in the actuation arm 114 to receive a locking pin to lock the actuation arm 114 in position when in either the open or, closed orientation. Alternately, the stationary arm can have two holes that will mate with a single hole in the actuation arm. Thus, when the actuation arm 114 is pulled up in the vertical direction, the pivotable hook 106 rotates around the pivot 116 to an open position as shown in FIG. 9. Similarly, when the actuation arm 118 is pushed down in the vertical direction the pivoted hook 106 rotates about the pivot 110 and second pivot point 116 to the closed position illustrated in FIG. 8, locking the dowel 92 in the slot 104 of the clevis 90. Thus, the locking mechanism 96 pins the tie rod end 74 to the wall mounted anchor 64 and can be locked in place in either the open or closed position from the operating deck elevation 52 using locking pins inserted through the holes 122. The position locking pins utilize a lanyard to prevent them from being dropped into the reactor cavity 54 or misplaced. FIG. 10 is a perspective view of another embodiment 124 of the tie rod 58, 60 employed by this invention. The coupling 70 to the reactor head is the same as that previously illustrated in FIG. 2. The short length of threaded piping 126 and the long threaded rod 130 on either side of the turn buckle 128 that are employed to adjust the length of the tie rod has been reversed from the embodiments illustrated in FIGS. 2 and 4. The hex jam nuts 132 are provided to lock in the adjusted length of the pipe. A set screw 134 on the open end clevis 136 on the distal end of the tie rod prevents the clevis from rotating. The open end clevis 136 has two fork tines 138 and 140 are separated by a distance that will accommodate the anchor plate fitting there between, as will be appreciated from the description to follow. The open end clevis 136 includes a downwardly facing slot 142 that is sized to accommodate a dowel pin that it will fit over. The lower portions 144 and 146 of the vertical walls of the slot 142 are angled to guide the open end clevis 136 over the dowel pins. FIG. 11 is a perspective view of the coupling between the anchor 64 and the distal ends 74 of the tie rod 58, 60, in accordance with the embodiment 124. The hex jam nut 132 and the set screw 134 has been omitted for convenience. Like reference characters are used for the corresponding components among the several figures. The anchor 64 includes an embedment plate 148 and a laterally extending lug 150 which is reinforced by the gussets 152. The lug 150 is connected to the open end clevis 136 through an extension plate 154 that is pivotably connected to the lug 150 by way of the extension plate pin 156. The extension plate pin 156 enables the extension plate 154 to rotate in a vertical plane to align the clevis pin dowel 158 with the downwardly facing clevis pin slot 142 in the clevis tines 138 and 140. A U-shaped bracket of an angle adjustment arm assembly 160 is connected to each side of the extension plate 154 and spans over the lug 150. Set screw 162 and lock nut 164 adjusts the height of the U-shaped bracket of the angle adjustment arm assembly 160 over the lug 150 and thus raises or lowers the angle of the extension plate 154 until it is in alignment with the tie rod 58, 60. A tie rod locking plate assembly 168 is rotationally connected on top of the extension plate 154. The tie rod locking plate assembly 168 has a locking bar 170, shown in FIG. 11 in the locked position where it sits over the tines 138 and 140, preventing the open end clevis 136 from being dislodged from the clevis pin 158. The locking bar 170 can be rotated 90 degrees over the top 172 of the extension plate 154 to uncover the tines 138 and 140 for removal of the tie rod end 74 from the clevis pin 158. FIG. 12 shows the extension plate 154 in more detail. The clevis pin 158 passes through an opening in the extension plate 154 and extends on both sides of the extension plate. As can better be appreciated from FIG. 13, a keeper plate 174 sits within a notch in the clevis pin 158 and is secured by screws to the extension plate to firmly hold the clevis pin 158 in place. The hold down bar 170 is connected to the top of the extension plate 172 by a shoulder screw 176 which is loosely tightened to enable the hold down bar 170 to rotate about the screw 176. As can be seen from FIG. 13, spring plungers 178 which fit in indentations in the underside of a hold down bar 170 engage the hold down bar in the closed position to resist rotation. A similar set of plungers can be provided to engage the underside of the hold down bar 170 when it is rotated 90 degrees to the open position to avoid interference with removal of the tines 138 and 140 from the clevis pin 158 as the tie rod 58, 60 is lifted. Rotation of the hold down bar 170 between closed and open positions can be achieved through the use of the long-handled hold down tool 182 illustrated in FIG. 14. The hold down tool has a forked end 184 for engaging the recess 180 in the locking bar 170. The forked end 184 is connected through an elongated shaft 188 to a handle 186 which can be turned from the operating deck 52 previously illustrated in FIG. 2. Accordingly, in addition to eliminating the need for use of a polar crane for this refueling activity, the quick disconnect control rod drive mechanism seismic support tie rod system of this invention eliminates the need for unsafe ladders; eliminates the need for laydown space for the tie rods; and eliminates the potential for the dropping of loose parts. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. For example, other attachment configurations between the anchor plates and the tie rod ends that can be engaged and disengaged remotely can be employed without departing from the scope of this invention. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention, which is to be given the full breath of the appended claims and any and all equivalents thereof. |
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042228220 | claims | 1. A method of operating a pressurized water nuclear reactor steam generation system having a reactive core including fissile material with an axial dimension, means for controlling the axial power distribution of the core, a coolant medium and moderator for respectively removing heat from the core and slowing down neutrons in the core for fissioning, the moderator having a negative reactivity temperature coefficient, a primary coolant loop for conveying the coolant medium to a heat exchanger steam generator and a secondary loop for conveying the steam generated to a steam utilization device, including the steps of: maintaining a substantially symmetric power distribution axially throughout the core during normal reactor power operation including load follow; operating the steam generator completely below the superheated steam region throughout normal power operation, including load follow; and increasing the reactivity within the core and the reactor's power output without substantially altering the axial power distribution of the core, upon a given demand, by reducing the average temperature within the core, during normal reactor power operation including load follow, while maintaining a substantially constant primary coolant flow rate. 2. The method of claim 1 wherein the coolant medium is circulated into and out of the core and through a heat exchanger in a closed recycling primary system wherein the heat exchanger places the coolant medium in heat exchange relationship with water to form steam which is conveyed through a throttling valve to an exhaust and wherein the temperature of the core is reduced by increasing the amount of heat removed from the coolant medium through the heat exchanger to produce steam by increasing the amount of steam exhausted by opening the throttling valve. 3. The method of claim 2 including the step of preventing the temperature of the core from being reduced below a preselected value by limiting the opening of the throttling valve. 4. The method of claim 3 wherein the preselected value is twenty degrees Fahrenheit below the normal operating temperature of the coolant. 5. The method of claim 1 wherein the means for controlling the axial power distribution of the core includes control rods extending axially for at least the axial length of the core and movable axially into and out of the core and wherein the axial power distribution is represented by the difference in flux generated in the upper and lower halves of the core, including the step of maintaining the difference in flux generated in the upper and lower halves of the core substantially constant throughout reactor power operation by the axial positioning of the control rods. 6. The method of claim 5 including the step of preventing the control rods from being withdrawn from the core when the flux difference substantially exceeds a pre-established value. 7. The method of claim 1 wherein the coolant medium includes a neutron absorbing element wherein the reactivity increasing step further includes diluting the neutron absorbing element within the coolant medium. |
description | This patent application claims the benefit of priority under 35 U.S.C. §119 from Korean Patent Application No. 10-2011-0026909 filed on Mar. 25, 2011, the contents of which are incorporated herein by reference. 1. Field of the Invention The present disclosure relates to an apparatus for recovering residual salt from the reduced uranium metal. 2. Description of the Related Art The Korea Atomic Energy Research Institute (KAERI) researches and develops pyroprocessing process which is a technology that recovers valuable resources such as uranium from the spent fuel. The recovered resources are recycled into fuel for next-generation sodium fast reactors, resulting in increased uranium usage efficiency and a marked decrease in radiotoxicity and the amount of radioactive waste generated, thereby increasing the safety and economic efficiency. The KAERI pyroprocessing starts from an electrolytic reduction process in which the spent oxide fuel is reduced into a metal with an electrochemical method. The metal formed in the electrolytic reduction process is used as a feed of an electrolytic refining process as the next process, to thereby selectively recover high purity uranium. The electrolytic reduction process and the electrolytic refining process, which are electrochemical processes, use molten salt as an electrolyte medium. The electrolytic reduction process uses molten lithium chloride (LiCl) salt with small amount of Li2O, and the electrolytic refining process uses eutectic salt of LiCl and potassium chloride (KCl). Thus, in order to increase the level of connectivity between the electrolytic reduction process and the electrolytic refining process, it is important to remove the residual LiCl from the metal reduced in the electrolytic reduction process to maintain the composition of LiCl—KCl salt in the electrolytic refining process. To this end, the present inventors disclose an apparatus that quickly cools LiCl vapor below its melting point to form powder, thereby recovering pure LiCl from mixed molten salt or a uranium metal reduced in the electrolytic reduction process. Embodiments of the present invention are directed to provide an apparatus for recovering residual salt from the reduced uranium metal. According to an aspect of the present invention, there is provided an apparatus for recovering residual salt from the reduced uranium metal, the apparatus including: an evaporating chamber accommodating mixed molten salt or a reduced uranium metal; a heating furnace surrounding the evaporating chamber to heat the mixed molten salt in the evaporating chamber; an insulator disposed over the evaporating chamber to block heat generated from the evaporating chamber, and including an evaporating pipe in a center thereof to move vapor generated from the evaporating chamber; a receiver disposed over the insulator to collect powder formed by condensing and solidifying vapor passing through the evaporating pipe; and a condenser disposed over the receiver to prevent the vapor passing through the evaporating pipe from leaking out of the apparatus. The mixed molten salt or the reduced uranium metal includes lithium chloride (LiCl). The evaporating chamber may include a thermocouple to measure a temperature of the evaporating chamber. The evaporating chamber may include a container containing the mixed molten salt or the reduced uranium metal. The receiver may include a thermocouple to measure a temperature of the receiver. The condenser may be provided with a pipe to connect a vacuum pump to the condenser, and air-tightly seal the apparatus. The vacuum pump may be provided with a valve to control a movement of vapor by a predetermined degree of vacuum (pressure). The pipe may include a pressure gauge and a filter. The apparatus may further include cone-shaped baffles that is disposed in an upper portion of the condenser to sufficiently cool vapor, prevents powder from being formed on an upper end of the condenser, and sufficiently cools vapor generated from the evaporating chamber to prevent the vapor from arriving at a vacuum pump. The apparatus may further include a glove box to recover powder collected within the receiver in inert atmosphere. Features and advantages of the present invention will be more clearly understood by the following detailed description of the present preferred embodiments by reference to the accompanying drawings. It is first noted that terms or words used herein should be construed as meanings or concepts corresponding with the technical sprit of the present invention, based on the principle that the inventor can appropriately define the concepts of the terms to best describe his own invention. Like reference numerals refer to like elements throughout. Also, it should be understood that detailed descriptions of well-known functions and structures related to the present invention will be omitted so as not to unnecessarily obscure the important point of the present invention. Hereinafter, specific embodiments of the present invention will be described in detail with reference to the accompanying drawings. A spent oxide fuel is reduced into a metal in an electrolytic reduction process using molten salt of LiCl as an electrolyte medium, and the metal includes about 20% residual molten salt of LiCl. Powder of LiCl recovered by an apparatus for recovering residual salt from the reduced uranium metal according to an embodiment of the present invention can be reused in an electrolytic reduction process. Since LiCl has a high melting point of about 605° C. and a low vapor pressure, high temperature and low pressure are needed to distill LiCl. FIG. 1 is a schematic view illustrating an apparatus for recovering residual salt from the reduced uranium metal according to an embodiment of the present invention. Referring to FIG. 1, the apparatus includes an evaporating chamber 1, a heating furnace 2, an insulator 3, a receiver 5, and a condenser 6. In detail, the evaporating chamber 1 accommodates mixed molten salt or a uranium metal reduced from a spent nuclear fuel. The mixed molten salt or the reduced uranium metal includes lithium chloride (LiCl). The heating furnace 2 surrounds the evaporating chamber 1 to heat the mixed molten salt or the reduced uranium metal in the evaporating chamber 1. The insulator 3 is disposed over the evaporating chamber 1 to block heat generated from the evaporating chamber 1, and includes an evaporating pipe 4 in the center thereof to move vapor generated from the evaporating chamber 1. The insulator 3 makes a large temperature difference between the evaporating chamber 1 and the condenser 6 to increase supersaturation of LiCl vapor, so that fine white powder of LiCl can be collected. The receiver 5 is disposed over the insulator 3 to collect powder formed by condensing and solidifying vapor passing through the evaporating pipe 4. The condenser 6 is disposed over the receiver 5 to prevent vapor passing through the evaporating pipe 4 from leaking out of the apparatus. The evaporating chamber 1 and the receiver 5 may include thermocouples 7 and 9, respectively, to measure temperature of the evaporating chamber 1 and the receiver 5. The evaporating chamber 1 may include a container 8 containing the mixed molten salt. A part contacting the mixed molten salt may be formed of an Inconel material (an Fe—Ni—Cr-based cast alloy) having excellent corrosion resistance. The container 8 contains the mixed molten salt. The heating furnace 2 heats the mixed molten salt, and a vacuum pump 10 connected to the condenser 6 depressurizes the mixed molten salt, thereby evaporating LiCl from the mixed molten salt. The condenser 6 may be provided with a pipe 13 to connect the vacuum pump 10 to the condenser 6, may air-tightly seal the apparatus, and be depressurized to evaporate LiCl. The pipe 13 may be provided with a pressure gauge 11 to measure and control a pressure variation and a depressurization condition. The vacuum pump 10 is provided with a valve 16, so that vapor generated from the evaporating chamber 1 is moved to the condenser 6 by a predetermined degree of vacuum (pressure), and the moving of the vapor is controlled by the valve 16. The pipe 13 may be provided with a filter 12 to prevent LiCl vapor from being introduced to the vacuum pump 10. Baffles 14 having a cone shape may be disposed in the upper portion of the condenser 6. The baffles 14 sufficiently cools LiCl vapor, prevents LiCl powder from being formed on the upper end of the condenser 6, and sufficiently cools vapor generated from the evaporating chamber 1 to prevent the vapor from arriving at the vacuum pump 10. The apparatus may include a glove box 15 to recover powder collected within the receiver 5 in inert atmosphere, thereby safely collecting LiCl having deliquescence. LiCl having deliquescence absorbs moisture from air, and dissolves in the absorbed water to form a solution that is difficult to handle. A material contacting the solution is susceptible to contamination and corrosion. Thus, the glove box 15 prevents contamination and corrosion in inert atmosphere such as argon atmosphere, and recovers pure LiCl powder. However, the present invention is not limited thereto, provided that inert atmosphere is maintained in the apparatus, and pure LiCl powder is recovered. A process using the apparatus to recover LiCl powder from the reduced uranium metal will now be described. Mixed molten salt of LiCl or a uranium metal reduced from a spent nuclear fuel is accommodated in the evaporating chamber 1. Then, the receiver 5 and the condenser 6 are installed on the evaporating chamber 1. Then, the condenser 6 is connected to the vacuum pump 10 through the pipe 13. It is checked using the vacuum pump 10 and the pressure gauge 11 whether the apparatus is air-tightly sealed. Then, the evaporating chamber 1 is heated using the heating furnace 2 surrounding the evaporating chamber 1 until an inner temperature of the evaporating chamber 1 is about 850° C. or higher, so as to melt the mixed molten salt of LiCl, or LiCl included in the reduced uranium metal. When the inner temperature of the evaporating chamber 1 is constant at the predetermined temperature, the vacuum pump 10 is operated, and the valve 16 of the vacuum pump 10 is opened to evaporate LiCl in the evaporating chamber 1. LiCl vapor is moved upward to the condenser 6 through the evaporating pipe 4 to form LiCl powder that is collected in the receiver 5. At this point, the temperature of the receiver 5 is increased by the heat of sublimation generated when the LiCl vapor is transformed to the LiCl powder. A time taken for evaporating the mixed molten salt of LiCl is proportional to the amount of LiCl included in the mixed molten salt or the reduced uranium metal. When the increased temperature of the receiver 5 begins to decrease, the evaporation of the mixed molten salt is completed. Then, power supplied to the heating furnace 2 is cut off, and the apparatus is cooled down to room temperature. After that, the condenser 6 and the receiver 5 are removed to recover the LiCl powder collected in the receiver 5. FIGS. 2A to 2E are images illustrating an apparatus for an apparatus for recovering residual salt from the reduced uranium metal according to an embodiment of the present invention. A heating furnace is illustrated in FIG. 2A. The apparatus including a condenser in the upper portion thereof is illustrated in FIG. 2B. A receiver provided with an evaporating pipe and an insulator is illustrated in FIG. 2C. An evaporating chamber in which the container is accommodated is illustrated in FIG. 2D. The apparatus including a glove box is illustrated in FIG. 2E. FIG. 3 is a graph illustrating temperature variations with time in a process using an apparatus for recovering residual salt from the reduced uranium metal according to an embodiment of the present invention. Referring to FIG. 3, an initial temperature of a heating furnace of the apparatus is about 1000° C., and an initial temperature of a receiver is about 200° C. LiCl melted at the temperature of about 1000° C. is evaporated by a vacuum pump to form LiCl vapor. While the LiCl vapor is moved upward through an evaporating pipe, a temperature of the receiver increases. When the evaporating of LiCl is completed, the temperature of the receiver decreases. FIG. 4 is a graph illustrating temperature variations of a receiver with time in a process using an apparatus for recovering residual salt from the reduced uranium metal according to an embodiment of the present invention. Referring to FIG. 4, as a heating temperature for the residual molten salt increases, a temperature of a receiver increases. In addition, as the heating temperature for the residual molten salt decreases, a time taken for evaporating the residual molten salt is increased. FIG. 5A is an image illustrating LiCl powder recovered from an apparatus for recovering residual salt from the reduced uranium metal according to an embodiment of the present invention. FIG. 5B is a graph illustrating a result of an X-ray diffraction (XRD) analysis of the recovered LiCl powder of FIG. 5A. Referring to FIG. 5B, the result of the XRD analysis shows that LiCl is recovered using the apparatus. Recovery rates of LiCl powder recovered using an apparatus for recovering residual salt from the reduced uranium metal according to an embodiment of the present invention, and states of the recovered LiCl powder are shown in Tables 1 and 2. TABLE 1RepeatingNumber ofEvaporation RateRecovery RateProcess(%)(%)110092210098310010241001005100100 As shown in Table 1, as the number of times of repeating a recovery process using the apparatus increases, the recovery rate increases. The recovery rate greater than 100% means that powder left in a previous process is recovered. TABLE 2EvaporationTemperaturePowderMelt Crystals(° C.)(%)(%)8509010900653510005545 As shown in Table 2, 90% of LiCl evaporated at 850° C. is powder, and 10% thereof is melt crystals. As an evaporation temperature increases, the amount of powder decreases, and the amount of melt crystals increases. According to the embodiments, mixed molten salt including LiCl is distilled at high temperature and low pressure to form LiCl vapor, and the LiCl vapor is quickly cooled below the melting point thereof to transform pure LiCl powder. Accordingly, molten salt of LiCl can be removed from a uranium metal reduced from a spent nuclear fuel. Although the preferred embodiments of the present invention have been disclosed for illustrative purposes, those skilled in the art will appreciate that various modifications, additions and substitutions are possible, without departing from the scope and spirit of the invention as disclosed in the accompanying claims. |
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047626640 | 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 moved from a fuel assembly and through a transition funnel by movable grippers at opposite ends of the funnel. One movable gripper reciprocates between gripping and release positions in a gap between the fuel assembly and the transition funnel. A stationary gripper can be located in the gap at the entry side of the funnel to hold the fuel rods while the movable gripper returns from a release position to the gripping position. Both grippers include members which can be pressed into frictional engagement with the spaced apart array of fuel rods. All of the fuel rods are withdrawn concurrently and are merged toward one another into a tighter array within the transition funnel and emerge as a bundle. A movable and a stationary bundle gripper are provided between the funnel and the storage container to advance the bundle of fuel rods into the container. |
description | This application claims the benefit of U.S. Provisional Application No. 61/625,491 filed Apr. 17, 2012. U.S. Provisional Application No. 61/625,491 filed Apr. 17, 2012 is hereby incorporated by reference in its entirety. The following relates to the nuclear power arts, nuclear reaction control arts, and related arts. Pressurized water reactor (PWR) designs have certain benefits over other light water reactor designs such as boiling water reactor (BWR) designs. For example, a PWR does not include complex steam separation hardware in the upper portion of the reactor vessel. In conventional PWR designs, this has allowed the control rod assemblies (CRAs) to penetrate the reactor through the top of the pressure vessel with the control rod drive mechanism (CRDM) units mounted externally above the pressure vessel. In this arrangement, a vessel penetration passing a connecting rod occurs at the top of the pressure vessel, whereas in a BWR design such penetrations are at the bottom of the pressure vessel, and in some design even below the reactor core itself. In some planned small modular reactor (SMR) designs, it is contemplated to improve upon the conventional reactor design by locating the entirety of the CRDM units inside the pressure vessel. See Stambaugh et al., U.S. Pub. No. 2010/0316177 A1; Desantis, U.S. Pub. No. 2011/0222640 A1; Stambaugh et al., WO 2010/144563 A1; and Thome et al., U.S. Pub. No. 2010/0316181 A1; all of which are incorporated herein by reference in their entireties. This arrangement eliminates mechanical vessel penetrations entirely, replacing them with electrical and/or hydraulic penetrations to drive the internal CRDM units. In some such SMR designs, it is further contemplated to employ a so-called “integral” PWR configuration in which the steam generators are also located inside the pressure vessel. See Thome et al., U.S. Pub. No. 2010/0316181 A1 which is incorporated herein by reference in its entirety. An integral PWR with internal CRDM units places a substantial number of components inside the pressure vessel. The internal steam generators need to have sufficient height to controllably convert inlet feedwater to steam. The control rod system must also have substantial height: The motors of the CRDM units must be located at least the full length of the control rods above the reactor core to allow full rod retraction, and a similar clearance must be provided above the motors to accommodate the connecting rods when the control rods are fully retracted. Thome et al., U.S. Pub. No. 2010/0316181 A1, which is incorporated herein by reference in its entirety, leverages the conventional arrangement of a central riser surrounded by a downcomer annulus to achieve a compact integral PWR configuration including internal CRDM units. The steam generators are located in the downcomer annulus, while the CRDM units are located inside the central riser. This enables substantial vertical overlap between the steam generators and the CRDM units, providing a more compact SMR design. In one aspect of the disclosure, an apparatus comprises: a cylindrical pressure vessel including an upper vessel section and a lower vessel section joined by a mid-flange; a cylindrical central riser disposed concentrically inside the cylindrical pressure vessel and including an upper riser section disposed in the upper vessel section and a lower riser section disposed in the lower vessel section; a reactor core comprising fissile material disposed inside the cylindrical pressure vessel in the lower vessel section; and control rod drive mechanism (CRDM) units controlling control rod insertion into the reactor core. The CRDM units are disposed inside the cylindrical pressure vessel above the reactor core and in the lower vessel section with no vertical overlap between the upper vessel section and the CRDM units. In another aspect of the disclosure, an apparatus comprises: a cylindrical pressure vessel including an upper vessel section and a lower vessel section joined by a mid-flange; a cylindrical central riser disposed concentrically inside the cylindrical pressure vessel and including an upper riser section disposed in the upper vessel section and a lower riser section disposed in the lower vessel section; steam generators disposed inside the cylindrical pressure vessel in the upper vessel section; a reactor core comprising fissile material disposed inside the cylindrical pressure vessel in the lower vessel section; and control rod drive mechanism (CRDM) units disposed inside the cylindrical pressure vessel above the reactor core and in the lower vessel section. There is no vertical overlap between the steam generators and the CRDM units. In another aspect of the disclosure, a method is disclosed, which is suitably performed in conjunction with the apparatus of either one of the two immediately preceding paragraphs. The method comprises disconnecting the upper vessel section from the mid-flange, and removing the upper vessel section without lifting the upper vessel section over any portion of the CRDM units. The method may further include disconnecting the mid-flange from the lower vessel section, and removing the mid-flange and the CRDM units from the lower vessel section with the CRDM units hanging suspended from the mid-flange. Disclosed herein are small modular reactor (SMR) designs employing an integral pressurized water reactor (integral PWR) configuration with internal control rod drive mechanism (CRDM) units. The approaches disclosed herein provide substantial operational benefits. The disclosed integral PWR designs do not include any vertical overlap between the steam generators or the upper vessel, on the one hand, and the CRDM units located in the lower vessel on the other hand. The disclosed designs establish a complete separation of components into: (1) an upper vessel section that includes the steam generators but not the CRDM units; and (2) a lower vessel section that includes the CRDM units but not the steam generators. The steam generators do not extend downward into the lower vessel, and conversely the CRDM units do not extend upward into the upper vessel. This disclosed configuration has substantial benefits. During an opening of the reactor vessel for refueling, maintenance, inspection, or other purposes, the upper vessel section including the steam generators can be lifted off and moved laterally without raising it above an overlapping portion of the CRDM units. This allows the height of the containment structure to be reduced. Further, by placing the CRDM units entirely in the lower vessel, more of the total vessel height can be accommodated by the lower vessel as compared with the upper vessel. Said another way, for a given total pressure vessel height, the disclosed designs enable the lower vessel to be made relatively taller and the upper vessel relatively shorter. The reduced upper vessel height translates into reduced weight as well, which further facilitates upper vessel removal. During refueling, the upper internals (including the internal CRDM units and control rod guide frames) must be removed to gain access to the reactor core located at or near the bottom of the lower vessel section. The disclosed configuration suspends the upper internals from a “mid-flange” which is located above the upper internals and at the juncture between the upper and lower vessel sections. With this configuration, and after removal of the upper vessel section, a crane or other robotic lifting apparatus can connect with the mid-flange and lift the upper internals out of the lower vessel as a single-piece unit suspended from the mid-flange in order to provide access to the reactor core from above. Because the upper internals are suspended from and hang below the mid-flange which is providing the anchor point for lifting out the upper internals, there is no tendency for the upper internals to tilt. In some embodiments, all power and control lines for the CRDM units are brought in through the mid-flange. In this case, internal connections running from the mid-flange to the CRDMs can be left connected when the upper internals are lifted out of the lower vessel. Indeed, if a dedicated robotic assembly is available then the upper internals can be lifted out and moved to a storage/maintenance bay inside containment, and can be inspected and any indicated maintenance performed while the upper internals remain connected to the robotic assembly used in their removal. With reference to FIG. 1, an illustrative integral PWR includes a cylindrical pressure vessel including an upper vessel section 6 and a lower vessel section 8 joined by a mid-flange 10. A cylindrical central riser is disposed concentrically inside the cylindrical pressure vessel and includes an upper riser section 12 disposed in the upper vessel section 6, a lower riser section 14 disposed in the lower vessel section 8, and a riser transition section 16 connecting the upper riser section 12 and the lower riser section 14. To distinguish the components, the upper and lower riser sections 12, 14 are shown in dashed lines in FIG. 1 while the riser transition section 16 is shown in solid lines in FIG. 1. The riser transition section 16 is connected to the mid-flange 10, in the illustrative embodiment by gussets 20 having first ends welded to the mid-flange 10 and second ends welded to the riser transition section 16. In the illustrative embodiment, the gussets 20 are angled downward such that the riser transition section 16 is disposed below the mid-flange 10. As used herein, the modifier “cylindrical” is intended to denote a generally cylindrical element, and allows for some deviation from a mathematically perfect cylinder. For example, the illustrative pressure vessel of FIG. 1 has a generally narrower upper vessel section 6 as compared with the lower vessel section 8, with the upper vessel section 6 expanding near its connection with the mid-flange 10. Similarly the illustrative upper riser section 12 is smaller-diameter than the lower riser section 14, and the riser transition section 16 includes a frusto-conical portion to accommodate this diameter difference. Moreover, the cylindrical pressure vessel includes various vessel penetrations and other deviations from perfect cylinder. As used herein, the term “vertical” and references to “above” and “below” assume that the pressure vessel is positioned “vertically” with the upper vessel section 6 above the lower vessel section 8. Such phraseology is intended to encompass some tilt or deviation from a perfectly vertical orientation. With continuing reference to FIG. 1, the upper vessel section 6 contains steam generators 24 disposed inside the cylindrical pressure vessel. In the illustrative example, the steam generators 24 are entirely in the upper vessel section 6, and do not extend into the mid-flange 10 or lower vessel section 8. The steam generators 24 are not illustrated in detail, but can have various configurations, such as a helical configuration (see Thome et al., U.S. Pub. No. 2010/0316181 A1 which is incorporated herein by reference in its entirety), a once-through straight-tube steam generator (OTSG) configuration optionally including a lower internal economizer portion, or so forth. The steam generators 24 are internal to the pressure vessel, and the PWR is referred to as an integral PWR. In the illustrative embodiment, the steam generators 24 are secured to the upper vessel section 6 such that the upper vessel section 6 and the steam generators 24 can be lifted as a unit (once the bolts or other fasteners securing the upper vessel section 6 to the mid-flange 10 are detached). Because the steam generators 24 are entirely in the upper vessel section 6, and do not extend into the mid-flange 10 or lower vessel section 8, the upper vessel/steam generators unit does not need to be raised to any particular height to lift the steam generators out of the lower vessel before moving the lifted assembly laterally. The illustrative upper vessel section 6 also includes an uppermost portion 26 that defines an internal pressurizer and may include pressure control components such as spargers, resistive heaters, or so forth. Alternatively, an external pressurizer may be provided and operatively connected with the pressure vessel via suitable piping in order to control pressure inside the pressure vessel. It is to be understood that the pressure vessel of the integral PWR contains a reactor core 30 comprising fissile material such as 235U disposed inside the cylindrical pressure vessel in the lower vessel section 8, and preferably at or near the bottom of the lower vessel section 8. It is to be further understood that the pressure vessel of the integral PWR contains primary coolant water in a subcooled state that fills the interior volume of the pressure vessel, except for a steam bubble in the internal pressurizer volume 26 (this steam bubble is also primary coolant water, but in a steam state). The primary coolant water flows in a circuit including a “hot leg” that flows upward from the reactor core 30, rises through the central riser, discharges from the upper end of the upper riser section 12; the discharged primary coolant water then flows through a return “cold leg” through a downcomer annulus defined between the central riser and the inner cylindrical surface of the pressure vessel. In the illustrative integral PWR, the cold leg includes the steam generators 24, and the primary coolant flows downward through steam generator tubes (i.e., flows “tube-side”) or alternatively flows downward along the outsides of the tubes (i.e., flows “shell-side”). The primary coolant in the cold leg flows downward to the bottom of the central riser where it again reverses direction and re-enters the reactor core 30 to complete the primary coolant flow circuit. In the illustrative example of FIG. 1, feedwater is flowed upward through the steam generators 24, either shell-side (if the primary coolant flows tube-side) or tube-side (if the primary coolant flows shell-side). Heat from the primary coolant in the steam generators 24 heats the feedwater converting it to steam. Feedwater inlets and steam outlets and optional feedwater and steam plenums (features not shown) are provided to flow feedwater into the steam generators 24 and to extract the steam from the steam generators 24. Primary coolant flow circulating through the hot and cold legs is via natural convection driven by heat output by the reactor core 30. Optionally, reactor coolant pumps (RCPs) are provided to assist or drive the primary coolant flow. In illustrative FIG. 1, RCPs 32 are disposed near the top of the upper vessel section 6. The illustrative RCP 32 has a pump motor 34 located externally from the pressure vessel, and a driveshaft 36 passes through a vessel penetration in the upper vessel section 6 to operatively connect the motor 34 with an impeller 38 engaging the primary coolant inside the pressure vessel. The RCPs 32 may optionally include or be associated with further components (not shown), such as a pump housing surrounding the impeller 38 to define the pump flow, a pump plate separating the suction and discharge sides of the impeller 38, and so forth. The illustrative RCPs 32 are secured to the upper vessel section 6 such that the upper vessel section 6, the RCPs 32, and (as mentioned before) the steam generators 24 can be lifted as a unit (again, after bolts or other fasteners securing the upper vessel section 6 to the mid-flange 10 are detached). The illustrative RCPs 32 of FIG. 1 are an example. More generally, RCPs can be mounted elsewhere. See, for example, FIGS. 6 and 7 (discussed elsewhere herein) which illustrate RCPs located proximate to the mid-flange 10. With continuing reference to FIG. 1 and with further reference to FIGS. 2-5, an “upper internals” section of the nuclear reactor is further described. The upper internals include control rod drive mechanism (CRDM) units 40 controlling control rod insertion into the reactor core 30, and guide frames 42 guiding control rods into the reactor core 30. The guide frames 42 are located directly above the reactor core 30 in order to guide the control rods vertically into or out of the reactor core 30; accordingly, the guide frames 42 are located inside the central riser. The illustrative integral PWR employs internal CRDM units 40 which are located inside the pressure vessel, and are also located inside the central riser to engage the control rod assemblies. The illustrative CRDM units 40 are disposed inside the cylindrical pressure vessel above the reactor core 30 and in the lower vessel section 8. There is no vertical overlap between the upper vessel section 6 and the CRDM units 40, and there is no vertical overlap between the steam generators 24 and the CRDM units 40. With this configuration the CRDM units 40 are entirely inside the lower pressure vessel 8, and the CRDM units 40 therefore do not impede lift off of the upper vessel section 6 with its secured steam generators 24. In other words, the upper vessel section 6 and the steam generators 24 do not need to be lifted vertically to clear the CRDM units 40. Moreover, with particular reference to FIGS. 2-5, the upper internals are designed to be suspended from the mid-flange 10. In this way, the mid-flange 10 can be lifted out of the lower vessel section 8 (after removal of the upper vessel section 6) with the riser transition section 16 connected and with the upper internals suspended from the mid-flange 10 (see particularly FIGS. 4 and 5). This assembly can be lifted as a unit out of the lower vessel section 8, and moved laterally to an upper internals bay located inside containment for inspection or maintenance. The overhead suspension of the upper internals from the mid-flange 10 provides stable support that is not susceptible to tilting. Lifting out the upper internals in this way also provides unimpeded access from above to the reactor core 30 for refueling, fuel inspection or maintenance. Toward this end, the riser transition section 16 is connected to the mid-flange 10 in a secure fashion to provide an anchor for the suspended upper internals. In the illustrative embodiment, the riser transition section 16 is welded to the mid-flange 10 by welding the upper ends of the gussets 20 to the inner diameter of the mid-flange 10 and welding the lower ends of the gussets 20 to the riser transition section 16. A CRDM support plate 44 and a lower hanger plate 46 are suspended from the riser transition section 16 by tie rods 48, 50. In the illustrated configuration, tie rods 48 have lower ends connected with the CRDM support plate 44 and upper ends connected with the riser transition section 16 so that the CRDM support plate 44 is suspended indirectly from the mid-flange 10 via the riser transition section 16 and the gussets 20. Lower tie rods 50 then suspend the lower hanger plate 46 from the CRDM support plate 44. In this approach the riser transition section 16 is a weight-bearing member that supports the CRDM support plate 44 and the lower hanger plate 46. Alternatively, the upper ends of the tie rods 48 can be welded or otherwise connected directly to the gussets 20, with the tie rods passing through slots or holes in the riser transition section 16. To prevent mixing of primary coolant from the hot and cold legs, the slots or holes in the riser transition section 16 are suitably closed, for example by weldments that also secure the riser transition section 16 to the gussets 20. In this approach the riser transition section 16 is not a weight-bearing member. In another variation, the tie rods 48 and lower tie rods 50 can be constructed as continuous single-piece tie rods having their lower ends connected with the lower hanger plate 46 and their upper ends connected with the riser transition section 16 or directly with the gussets 20. In this variation, the CRDM support plate 44 connects with the continuous single-piece tie rods at a designated location along the length of the tie rods, i.e. at a designated location between the upper and lower ends of the tie rods. In yet another contemplated variation, the tie rods 48 (or alternatively the continuous single-piece tie rods as previously described) have their upper ends connected directly with the mid-flange 10. In this case the tie rods are at a slight angle respective to the vertical in order to accommodate the larger inner diameter of the mid-flange 10 compared with the inner diameter of the central riser. In this embodiment the riser transition section would not be a weight-bearing member, but would have slots or holes for the tie rods to pass through. With brief reference to FIG. 2, the CRDM units 40 each include a CRDM motor 52, a standoff 54, and a latch mechanism 56. In FIG. 2, a single CRDM unit 40ex is shown in an “exploded” view, that is, removed from the overall assembly, in order to show the CRDM motor 52, standoff 54, and latch mechanism 56. The CRDM standoff 56 mounts to the CRDM support plate 44 to support the CRDM unit with the CRDM motor 52 at a designated height. The CRDM motor 52 operates directly or indirectly to raise or lower a connecting rod 58 that is attached to a set of control rods via a spider (control rods and spider not shown). The assembly of the connecting rod, spider, and set of control rods is sometimes referred to as a control rod assembly (CRA). In normal operation, the CRDM motor 52 raises or lowers the CRA to insert or withdraw the constituent control rods into or from the reactor core 30. In the event of a SCRAM, the latch mechanism 56 releases the connecting rod 58 to allow the CRA to fall under force of gravity toward the reactor core 30 so as to rapidly insert the control rods fully into the reactor core 30. The CRDM unit 40ex of FIG. 2 conforms with embodiments disclosed in Stambaugh et al., U.S. Pub. No. 2010/0316177 A1; Desantis, U.S. Pub. No. 2011/0222640 A1; and Stambaugh et al., WO 2010/144563 A1, each of which is incorporated herein by reference in its entirety. In these embodiments, the CRDM motor 52 operates on a lead screw and the upper end of the control rod is electrically or hydraulically latched with the lead screw by the latch mechanism 56. In a SCRAM the latch mechanism 56 releases to SCRAM the CRA including the connecting rod 58, but not including the lead screw. The CRDM unit 40ex is optionally replaced by a more conventional CRDM unit in which the latch is omitted and SCRAM is initiated by opening a separable roller nut of the CRDM motor that connects with the lead screw, such that the CRA and the lead screw SCRAM together, or by a CRDM unit employing a ratchet mechanism to raise/lower the CRA. In any of these cases, it will be appreciated that when the control rods are fully withdrawn the connecting rod will extend well above the CRDM motor, and the CRDM unit should have sufficient height above the CRDM motor to accommodate this. In the disclosed approaches, the CRDM support plate 44 is suspended below the mid-flange 10 by a distance D (labeled only in FIG. 3) that is sufficient that the tops of the CRDM units 40 including the latch mechanism 56 (or, in other CRDM unit embodiments, including the upper housing or support for the connecting rod/lead screw) are located below the mid-flange 10, that is, so that the CRDM units 40 are disposed entirely in the lower vessel section 8. The CRDM units 40 have no vertical overlap with the upper vessel section 6 and have no vertical overlap with the steam generators 24. The gussets 20 support the weight of the upper internals (except in the contemplated variation in which the tie rods connect directly with the mid-flange). The gussets 20 are also substantially rigid components that locate the riser transition section 16 inside the pressure vessel. During temperature ramp-up as the reactor is started, and subsequent temperature ramp-down prior to a vessel opening, the various components undergo substantial temperature transients and, if different materials are used, some differential thermal expansion may be experienced. The illustrative gussets 20 are advantageously shaped to help accommodate differential thermal expansion by being angled downward such that the riser transition section 16 is disposed below the mid-flange 10. This allows the accommodation of any differential thermal expansion to be distributed in both vertical and lateral directions. As already discussed, the disclosed configurations advantageously enable the upper internals to be removed as a unit suspended below the mid-flange 10. To further facilitate this modularity and ease of removal of the upper internals, in some embodiments power delivery cabling is also integrated into this assembly. Toward this end, the power delivery cabling may include electrical feedthroughs 60 passing through the mid-flange 10, and mineral insulated (MI) cables 62 extending from the electrical feedthroughs 60 to the CRDM units 40 to conduct electrical power to the CRDM units 40. Additionally or alternatively, some or all of the feedthroughs 60 may be hydraulic feedthroughs passing through the mid-flange 10, and the cables 62 may correspondingly be hydraulic cables extending from the hydraulic feedthroughs 60 to the CRDM units 40 to conduct hydraulic power to the CRDM units 40. (Note, the power delivery cabling components 60, 62 are labeled only in FIGS. 2 and 3). Optionally, additional cabling 64 (labeled only in FIG. 3) may extend to the reactor core 30 to connect with in-core sensors (not shown). The MI or hydraulic cables 62 are relatively rigid and are suitably secured to structural elements of the mid-flange/upper internals assembly in order that the cables 62 can be lifted out with the assembly without disconnecting the cables from either the CRDM units 40 or the mid-flange feedthroughs 60. For example, the cables may be secured to the riser transition section 16 and/or to the tie rods 48 and/or to the CRDM support plate 44. In some embodiments, portions of the MI and/or hydraulic cables 62 that run along the CRDM support plate 44 are embedded in or secured to the CRDM support plate 44 such that the CRDM support plate 44 is a power distribution plate for laterally distributing electrical and/or hydraulic power to the CRDM units 40 supported by the plate 44. It is contemplated to similarly use the lower hanger plate 46 to laterally distribute MI cables connecting with in-core sensors of the array of fuel assemblies making up the reactor core 30. With reference to FIGS. 4 and 5, removal of the upper internals is illustrated. FIG. 4 shows the lower vessel section 8 and the mid-flange 10 which remain after the upper vessel section 6 has been removed. FIG. 5 shows the mid-flange 10 raised vertically to lift the suspended upper internals (i.e., the CRDM units 48 and guide frames 50 supported respectively by the CRDM support plate 44 and lower hanger plate 46) out of the lower vessel section 8. The lifted assembly can then be moved laterally (not illustrated) to provide overhead access to the reactor core 30 via a large opening 66 exposed at the top of the lower vessel section 8. With reference to FIGS. 6 and 7, an alternative embodiment is shown, in which the RCP's 32 mounted on the upper vessel section 6 proximate to the internal pressurizer 26 are replaced by RCPs 70 that are disposed entirely inside the cylindrical pressure vessel. The RCPs 70 are mounted on an annular pump plate 72 that has an outer circular perimeter connecting with the mid-flange 10 and an inner circular perimeter connected with the riser transition section 16. In some embodiments the outer and inner circular perimeters of the annular pump plate 72 are welded to the mid-flange 10 and the riser transition section 16, respectively. In such embodiments, the annular pump plate 72 can replace the structural function of the gussets 20, that is, the annular pump plate 72 supports the upper internals. (Alternatively, the gussets 20 can be retained, in which case the annular pump plate supports the RCPs 70 but not the upper internals). The annular pump plate 72 supports the RCPs 70 and also provides separation between the suction and discharge sides of the RCPs 70. In FIG. 6, a single RCP 70ex is shown in an “exploded” view, that is, removed from the overall assembly, in order to show it in more detail. The RCP 70ex has a suction side 80 and a discharge side 82. The RCP 70ex can be substantially any type of canned pump. The preferred embodiments have been illustrated and described. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof. |
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047675934 | claims | 1. A method of fabricating a pressure vessel comprising the steps of attaching a first inner presure vessel having means defining inlet and outlet openings to a top flange, placing a second inner pressure vessel, having means defining inlet and outlet opening, concentric with and spaced about said first inner pressure vessel and attaching said second inner pressure vessel to said top flange, placing an outer pressure vessel, having inlet and outlet openings, concentric with and spaced apart about said second inner pressure vessel and attaching said outer pressure vessel to said top flange, attaching a generally cylindrical inner inlet conduit and a generally cylindrical inner outlet conduit respectively to said inlet and outlet openings in said first inner pressure vessel, attaching a generally cylindrical outer inlet conduit and a generally cylindrical outer outlet conduit respectively to said inlet and outlet opening in said second inner pressure vessel, heating the assembled pressure vessel to a temperature above the melting point of a material selected from the group, lead, tin, antimony, bismuth, potassium, sodium, boron and mixtures thereof, filling the space between said first inner pressure vessel and said second inner pressure vessel with material selected from said group, filling the space between said second inner pressure vessel and said outer pressure vessel with material selected from said group, and pressurizing said material filling said spaces between said pressure vessels to a predetermined pressure, said step comprising pressurizing said spaces to a pressure whereby the wall of said first inner pressure vessel is maintained in compression during steady state operation of said pressure vessel. heating said outer pressure vessel to a temperature higher than the temperature of said first inner pressure vessel to define a temperature gradient across said first inner, second inner and outer pressure vesels, the temperature between said first inner and said second inner pressure vessels being higher than the melting point of a first filler material selected from said group, filling the space between said first inner and said second inner pressure vessels with said first molten material selected from said group and then pressurizing said molten material to pre-stress said first inner pressure vessel, lowering said temperature gradient between said first inner and second inner pressure vessels to below the melting point of said first filler material in said space between said first inner and said second inner pressure vessels, with the temperature gradient between said second inner and said outer pressure vessels above the melting point of said first filler material, filling the space between said second inner and said outer pressure vessels filled with said first molten material selected from said group and the pressurizing said molten material to pre-stress said second inner pressure vessel, lowering the temperature gradient between said second inner and said outer pressure vessels to below the melting point of said first filler material in said space between said second inner and said outer pressure vessels. heating said first inner, second inner and outer pressure vessels to a uniform temperature above the melting point of a first filler material selected from said group of filler materials, filling the space between said first inner and said second inner pressure vessels with said first molten material selected from said group and pressurizing said molten material to pre-stress said first inner pressure vessel, lowering said temperature of said first inner, second inner and said outer pressure vessels to below the melting point of said first filler material in said space between said first inner and second inner pressure vessels, and above the melting point of a second filler material selected from said group, filling the space between said second inner pressure vessel and said outer pressure vessel with said second molten material selected from said group and pressurizing said molten material to pre-stress said second inner pressure vessel, lowering the temperature of said first inner, second inner and outer pressure vessels to below the melting point of said second material in said space between said second inner and said outer pressure vessels. a top flange, a pressure vessel head adapted to engage said top flange, be connected thereto and maintain a seal therebetween, an inner pressure vessel attached to said top flange, an outer pressure vessel disposed concentric about said inner pressure vessel, spaced apart therefrom and attached to said top flange, a reactor coolant inlet port comprising a generally cylindrical inner inlet conduit attached to said inner pressure vessel and in fluid communication therewith, a generally cylindrical outer inlet conduit attached to said outer pressure vessel and in fluid communication with the space between said inner pressure vessel and said outer pressure vessel, a reactor coolant outlet port comprising a generally cylindrical inner outlet conduit attached to said inner pressure vessel and in fluid communication therewith, a generally cylindrical outer outlet conduit attached to said outer pressure vessel and in fluid communication with the space between said inner pressure vessel and said outer pressure vessel, a low melting point, high boiling point material selected from the group, lead, tin, antimony, bismuth, sodium, potassium, boron and mixtures thereof disposed in the space between said inner pressure vessel and said outer pressure vessel, and means for maintaining said low melting point, high boiling point material at a pressure sufficient to maintain the tensile stresses in the wall of said inner pressure vessel below the yield point of the wall material during transient pressure rises within said pressure vessel, and means for maintaining said low melting point, high boiling point material at a constant, predetermined multiple of the pressure inside the inner pressure vessel. said means for maintaining said low melting point, high boiling point material at a constant, predetermined multiple of the pressure inside the inner pressure vessel comprises a first generally cylindrical bellows closed at one end and having its open end attached to said inner pressure vessel with the interior of said first bellows in fluid communication with the space between said inner pressure vessel and said outer pressure vessel, said first bellows projecting into said inner pressure vessel, a second generally cylindrical bellows closed at one end disposed concentric about said first bellows and spaced apart therefrom and having its open end attached to the inside surface of said inner pressure vessel, the closed end of said second bellows being in contact with the closed end of said first bellows, compressible fluid disposed in the space between said first bellows and said second bellows, and means for maintaining said compressible fluid at a predetermined pressure. a top flange, a pressure vessel head adapted to engage said top flange, be connected thereto and maintain a seal therebetween, an inner pressure vessel attached to said top flange, an outer pressure vessel disposed concentric about said inner pressure vessel and spaced apart therefrom and attached to said top flange, a reactor coolant inlet port comprising a generally cylindrical inner inlet conduit attached to said inner pressure vessel and in fluid communication therewith, a generally cylindrical outer inlet conduit attached to said outer pressure vessel and in fluid communication with the space between said inner pressure vessel and said outer pressure vessel, a reactor coolant outlet port comprising a generally cylindrical inner outlet conduit attached to said inner pressure vessel and in fluid communication therewith, a generally cylindrical outer outlet conduit attached to said outer pressure vessel and in fluid communication with the space between said inner pressure vessel and said outer pressure vessel, a low melting point, high boiling point material selected from the group, lead, tin, antimony, bismuth, sodium, potassium, boron and mixtures thereof disposed in the spaced between said inner pressure vessel and said outer pressure vessel, and means for maintaining said incompressible fluid contained in the space between said inner and outer pressure vessel at a predetermined pressure whereby the material of said wall of said inner pressure vessel in maintained in compression during steady state operation of said pressure vessel. means for maintaining said low melting point, high boiling point material at a constant, predetermined multiple of the pressure inside the inner pressure vessel. a first generally cylindrical bellows closed at one end and having its open end attached to said inner pressure vessel with the interior of said first bellows in fluid communication with the space between said inner pressure vessel and said outer pressure vessel, said first bellows projecting into said inner pressure vessel, a second generally cylindrical bellows closed at one end disposed concentric about said first bellows and spaced apart therefrom and having its open end attached to the inside surface of said inner pressure vessel, the closed end of said second bellows being in contact with the closed end of said first bellows, compressible fluid disposed in the space between said first bellows and said second bellows, and means for maintaining said compressible fluid at a predetermined pressure. a top flange, a pressure vessel head adapted to engage said top flange, be connected thereto and maintain a seal therebetween, a first inner pressure vessel attached to said top flange, a second inner pressure vessel disposed concentric about said first inner pressure vessel and spaced apart therefrom and attached to said top flange, an outer pressure vessel disposed concentric about said second inner pressure vessel and spaced apart therefrom and attached to said top flange, a reactor coolant inlet port comprising a generally cylindrical inner inlet conduit attached to said first inner pressure vessel and in fluid communication therewith, a generally cylindrical outer inlet conduit attached to said second inner pressure vessel and in fluid communication with the space between said first inner pressure vessel and said second inner pressure vessel, a reactor coolant outlet port comprising a generally cylindrical inner outlet conduit attached to said first inner pressure vessel and in fluid communication therewith, a generally cylindrical outer outlet conduit attached to said second inner pressure vessel and in fluid communication with the space between said first inner pressure vessel and said second inner pressure vessel, a low melting point, high boiling point material selected from the gorup, lead, tin, antimony, bismuth, sodium, potassium, boron and mixtures thereof disposed in the space between said first inner pressure vessel and said second inner pressure vessel, and between said second inner pressure vessel and said outer pressure vessel, and means for maintaining said low melting point, high boiling point material at a pressure sufficient to maintain the tensile stresses in the wall of said inner pressure vessel below the yield point of the material of said wall during transient pressure rises within said pressure vessel, and means for maintaining said low melting point, high boiling point material at a constant, predetermined multiple of the pressure inside the inner pressure vessel. said means for maintaining said low melting point, high boiling point material at a constant, predetermined multiple of the pressure inside the inner pressure vessel comprises a first generally cylindrical bellows closed at one end and having its open end attached to said inner pressure vessel with the interior of said first bellows in fluid communication with the space between said inner pressure vessel and said outer pressure vessel, said first bellows projecting into said inner pressure vessel, a second generally cylindrical bellows closed at one end disposed concentric about said first bellows and spaced apart therefrom and having its open end attached to the inside surface of said inner pressure vessel, the closed end of said second bellows being in contact with the closed end of said first bellows, compressible fluid disposed in the space between said first bellows and said second bellows, and means for maintaining said compressible fluid at a predetermined pressure. a top flange, a pressure vessel head adapted to engage said top flange, be connected thereto and maintain a seal therebetween, a first inner pressure vessel attached to said top flange, a second inner pressure vessel disposed concentric about said first inner pressure vessel and spaced apart therefrom and attached to said top flange, an outer pressure vessel disposed concentric about said second inner pressure vessel and spaced apart therefrom and attached to said top flange, a reactor coolant inlet port comprising a generally cylindrical inner inlet conduit attached to said first inner pressure vessel and in fluid communication therewith, a generally cylindrical outer inlet conduit attached to said second inner pressure vessel and in fluid communication with the space between said first inner pressure vessel and said second inner pressure vessel, a reactor coolant outlet port comprising a generally cylindrical inner outlet conduit attached to said first inner pressure vessel and in fluid communication therewith, a generally cylindrical outer outlet conduit attached to said second inner pressure vessel and in fluid communication with the space between said first inner pressure vessel and said second inner pressure vessel, a low melting point, high boiling point material selected from the group, lead, tin, antimony, bismuth, sodium, potassium, boron and mixtures thereof disposed in the spaced between said first inner pressure vessel and said second inner pressure vessel, and between said second inner pressure vessel and said outer pressure vessel, and means for maintaining said incompressible fluid contained in the space between said inner and second inner pressure vessel at a predetermined pressure whereby the material of said wall of said inner pressure vessel in maintained in compression during steady state operation of said pressure vessel. means for maintaining said low melting point, high boiling point material at a constant, predetermined multiple of the pressure inside the inner pressure vessel. said means for maintaining said low melting point, high boiling point material at a constant, predetermined multiple of the pressure inside the inner pressure vessel comprises a first generally cylindrical bellows closed at one end and having its open end attached to said inner pressure vessel with the interior of said first bellows in fluid communication with the space between said inner pressure vessel and said outer pressure vessel, said first bellows projecting into said inner pressure vessel, a second generally cylindrical bellows closed at one end disposed concentric about said first bellows and spaced apart therefrom and having its open end attached to the inside surface of said inner pressure vessel, the closed end of said second bellows being in contact with the closed end of said first bellows, compressible fluid disposed in the space between said first bellows and said second bellows, and means for maintaining said compressible fluid at a predetermined pressure. means for detecting leakage of fluid from within said first inner pressure vessel disposed in the spaces between said first inner pressure vessel and said outer pressure vessel. a leak detection channel disposed about the inner periphery of said outer pressure vessel proximate the welded connections thereof to other parts thereof, and a conduit in fluid communication with said leak detection channel and the exterior of said pressure vessel. a leak detection channel disposed about the outer periphery of said first inner pressure vessel proximate the weled connections thereof to other parts thereof, and a conduit in fluid communication with said leak detection channel and the exterior of said pressure vessel. 2. The method of fabricating the pressure vessel as claimed in claim 1 further comprising the steps of 3. The method of fabricating the pressure vessel as claimed in claim 1 further comprising the steps of 4. A pressure vessel comprising 5. The pressure vessel as claimed in claim 4 wherein 6. A pressure vessel comprising 7. The pressure vessel as claimed in claim 6 further comprising 8. The pressure vessel as claimed in claim 7 wherein said means for maintaining said low melting point, high boiling point material at a constant, predetermined multiple of the pressure inside the inner pressure vessel comprises 9. A pressure vessel comprising 10. The pressure vessel as claimed in claim 9 wherein 11. A pressure vessel comprising 12. The pressure vessel as claimed in claim 11 further comprising 13. The pressure vessel as claimed in claim 11 wherein 14. The pressure vessel as claimed in claim 11 further comprising 15. The pressure vessel as claimed in claim 14 wherein said means for detecting leakage of fluid comprises 16. The pressure vessel as claimed in claim 14 wherein said means for detecting leakage of fluid comprises |
041785240 | description | DETAILED DESCRIPTION Now referring to the drawing, there is shown by illustration in FIG. 1 a radioisotope photoelectric generator, or "battery," made in accordance with the teaching of this invention. As shown, the battery includes a radioisotope source 10, such as cobalt 57, cadmium 109, tin 199 m, gadolinium 153, or gold 195. These sources emit photons up to approximately 200 keV energy and have the desirable characteristic that they emit no high-energy charged particles. The "battery" is formed by alternate plates of high-atomic number material 11, such as tantalum, lead or gold, and a low-atomic number material 12, such as aluminum or magnesium. Other materials may also be used for the high-Z plate such as bismuth, platinum, iridium, rhenium, tungsten, tin, or silver. Other materials may also be used for the low Z plate such as beryllium, carbon, calcium and titanium. Plates of like materials are connected to each other electrically and are insulated from the other set of plates by an evacuated spacing or by any suitable material which will not affect the operation thereof. It should be noted that the high-Z plates may contain one or more elements and the low-Z plates may also contain one or more elements. For example the "high-Z plate" alone might consist of a thick low-Z plate, e.g., aluminum (for structural rigidity) covered on both sides with high-Z material less than one electron range thick. For plates composed of more than one element, the materials within an electron range or less of each surface will determine whether it is a high-Z or a low-Z plate, from the operational standpoint. The efficiency of operation of the radioisotope photoelectric generator depends strongly on the combination of high- and low-Z materials used. The efficiency is generally higher the further the high-Z and low-Z plates are separated in atomic number. In order to prevent an outside radiation hazard, the radioisotope source and plates can be contained within a lead housing 13. Electrical leads, 14, 15 are brought out through the housing in order to connect the plates to an electrical load 16. It has been determined that each cell, one tantalum and one aluminum plate, will absorb about 5% of the incident photons for a 50 keV, monoenergetic, gamma ray source, and therefore for a single radioisotope source which emits 50 keV gamma rays, a "battery" may contain about 20 "cells" on each side of the source. FIG. 2 illustrates a "battery" with the plates pointing almost, but not quite, radially outward from the source 10. The "batteries" illustrated in FIGS. 1 and 2 may be made with more than one radioisotope source, that is, different radioisotope sources with the same or different atomic number and furthermore the radioisotope source may be "seeded" or distributed in the plates or insulating material. With two or more sources, alternate plates are electrically secured to each other the same as in the single-source battery and a larger number of plates may be used. Furthermore, a plurality of batteries may be connected electrically in parallel or in series to provide higher voltage or current or a combination thereof. Instead of a "battery" as shown in FIG. 1 with a single source at one end, the single-source could be placed in the middle between any two plates with alternate plates assembled on opposite sides of the radioisotope source. In assembly of the device, a radioisotope source which emits X-rays or gamma rays of less than 1 MeV energy is used. The energy is selected to maximize the efficiency of the unit or to determine the voltage of the power source. The unit is most efficient at about 100 KeV gamma ray energy. The amount of power output of the unit is proportional to the amount of radioisotope used. As an illustration: for a 50 keV gamma ray source, the tantalum plate has a thickness of 5.8 microns (one electron range) or less, and the aluminum plate has a thickness of 21 microns (one electron range) or greater, with a spacing between the plates sufficient to avoid breakdown of the insulator used but less than one electron range. The thickness of the lead housing must be sufficient to attenuate the radiation emitted by the source or sources. For most of the above-named sources, less than 0.5 cm of lead is required. The electrical load (for example a resistor) on the outside of the "battery" which will allow a voltage of 25 KV to develop between the plates has a value determined by the amount of isotope present in the battery. The radioisotope photoelectric generator (RPEG) utilizes a radioisotope which emits X-rays or gamma rays of less than 1 MeV as the energy source. A "battery" is formed by alternate layers of high-Z and low-Z material which are insulated by vacuum or other insulating material as described above and shown in FIGS. 1 and 2. The low-energy photons from the radioactive source interact predominantly with the high-Z material by the photoelectric process, ejecting photoelectrons whose energy extends up to the incident photon energy E (Compton and Auger electrons are also ejected but to a lesser extent). By selecting the thickness of the high-Z material to be less than one electron range (at energy E) and that of the low-Z material to be more than one electron range, there will be a net electron transfer from the high-Z plates to the low-Z plates because electrons are emitted predominantly from the high-Z plates and stop predominantly in the low-Z plates. Photons eject electrons from both high-Z and low-Z materials but a much greater number are ejected from high-Z materials. A potential difference will build up between the high-Z and low-Z plates. An upper limit for this potential difference in kilovolts is the energy E in keV. The high-Z plates are connected electrically together and the low-Z plates are connected electrically together thus forming the "battery." The battery can deliver power to an external electrical load preferably but not necessarily resistive as shown in the figure, the value being chosen to maximize the power delivered to the electrical load, to yield the desired voltage, to control the temperature of the plates, or a combination of such considerations. It has been determined that with plates as described above and with a source of energy of 50 keV with an electrical load selected to allow 25 KV to develop between the plates, the attenuation of the photons incident on any tantalum plate while traversing the tantalum and adjacent aluminum plate is 5.3% of the incident photons. It has also been determined that, with the above mentioned 50 keV source, the net charge, Q, transferred to the tantalum plate is: EQU Q=Ne[(.sup..epsilon. FTa.sup.+.epsilon. BTa)-(.sup..epsilon. FAl.sup.+.epsilon. BAl)] Where N is the number of photons incident on the plates (attenuation in traversing one set of Ta and Al plates is only 5.3% and is neglected), e is the electron charge, and the .epsilon.'s are the forward and back emission efficiencies for Al and Ta. Using values from a published article, "X-ray Induced Electron Emission from Metals," by K. W. Dolan, Sandia Laboratory, Sandia Report No. 74-8642, October 1974, one obtains EQU Q=3.8.times.10.sup.-3 Ne In traversing one cell of the "battery," the 50 keV photons will be attenuated by 5.3% so the efficiency for converting photons to electrons capable of reaching the alternate plates is simply ##EQU1## The power output per photon absorbed/sec is determined as follows for each cell: ##EQU2## where R is the number of 50 keV photons absorbed per second. The total energy absorbed per second by the cell from the 50 keV photons is ##EQU3## Using the power output as estimated above, the efficiency becomes ##EQU4## It must be pointed out that some effects have been neglected in this estimate such as electron absorption in the plates of origin and secondary electron emission from all surfaces. However, the efficiency calculated above is quite attractive and it is a lower limit in the sense that it could be increased substantially by allowing for non-normal incidence of the photon beam (0.degree. would be best) and by making the high-Z plate thinner than 5.8 microns since the photoelectrons are emitted predominantly at 90.degree. to the incident beam at such low photon energies. The photon beam should strike the plates at a small angle to the plates, generally less than 30.degree.. The effects of secondary electron emission can be substantially reduced by including a screen or grid between the plates and applying a small retarding potential to the screen (about 100 v). The efficiency of the RPEG has been determined at 3-4% which is an efficiency comparable to present RTG's. However, for ease of calculation in this determination, the photon beam from the radioactive source is assumed normal to the plates (a worst case). The efficiency will be increased substantially by including in the calculation non-normal incidence of the photons incident on the plates at a very small angle by concentrating the source at the center and positioning the plates almost, but not quite, radially outward as shown in FIG. 2. The RPEG, therefore, has a potential advantage of higher efficiency than the RTG's currently in use. The RPEG develops power at high voltages (typically up to 50 KV) and therefore has the advantage that its output can be set directly at the required voltage for a high-voltage supply rather than requiring a step-up transformer or other device to convert from low to high voltage at a sacrifice in efficiency. The RPEG can be tailored to provide a range of voltages available directly (by selection of different sets of plates and values of electrical load) to provide current limiting, or to control the temperature of the plates. Another advantage of the RPEG over the RTG is that a large percentage of the electrons' energy is dissipated by the field between the plates rather than being dissipated in heat. This reduces the temperature of the RPEG. This advantage can be enhanced by proper selection of the load impedance, plate and insulator thickness, etc., to minimize the photoelectron energy deposited as heat in the plates and insulators. The RPEG, like the RTG, can also be used as a combined electrical power and heat source for remote applications such as in the Artic or for deep space missions. The heat output of the RPEG can be selected by choices of load impedance, plate and insulator thickness, etc. The most essential new feature of the RPEG is the use of materials of different atomic number (Z) in a radiation field to develop a potential difference between the plates. Alternate materials are possible for the dielectric separating the plates. Vacuum or almost any insulator could serve in this role provided it could withstand the high fields. The plates can be almost any high-Z material combined with almost any low-Z material provided that the plates do not melt at high temperature. A screen can be added between the plates with a retarding potential applied to suppress secondary electron emission. There are several alternative methods of source placement. For example, the source can be placed all at the center, distributed throughout the plates or insulators, or different energy sources can be used in different regions of the RPEG. The thicknesses of the plates can be varied and, for example, the high-Z plate and the insulator could be deposited on a thicker low-Z plate for structural rigidity. Obviously many modifications and variations of the present invention are possible in light of the above teachings. It is therefore to be understood that within the scope of the appended claims the invention may be practiced otherwise than as specifically described. |
summary | ||
043895719 | summary | BACKGROUND OF THE INVENTION Charged particle beam systems, which are devices where an electron is brought to a final focus using magnetic or electrostatic lenses, have their performance limited at this time by fourth and higher order aberrations of the lenses. Examples of such electron beam devices are electron microscopes, electron microanalyzers, ion microscopes, and scanning transmission electron microscopes. A system for correcting spherical or third order aberration of lenses in electron beam devices was shown and described in U.S. patent application Ser. No. 088,224, an invention of Albert V. Crewe and David A. Kopf, entitled "Sextupole System for the Correction of Spherical Aberration", filed Oct. 25, 1979 U.S. Pat. No. 4,303,864. In that application, a sextupole focusing arrangement was described for compensating for spherical aberration in charged particle beam devices. In that arrangement, a sextupole was positioned between two focusing lenses. The third order aberrations of a sextupole are cylindrically symmetric and of opposite sign to those of the upstream focusing lens. By producing a beam crossover in the center of the sextupole, its aberrations were used to correct the third-order aberration of the upstream round lens. However, no workable solution for eliminating higher than third order aberration in charged particle beam systems was known. These higher order aberrations limit the usefulness of charged particle beam devices in that probe sizes cannot be reduced below a specified minimum, the sizes of acceptable images are larger than desired, and magnification of the images must be limited to an unacceptably low level. It is therefore an object of this invention to provide a means for compensating for third order, as well as higher than third order aberration in a charged particle beam system. Another object of this invention is to provide a means for compensating for higher than third order aberration in charged particle beam systems having as few elements as possible, preferably no more than three elements. Additional objects, advantages and novel features of the invention will be set forth in part in the description which follows, and in part will become apparent to those skilled in the art upon examination of the following, or may be learned by practice of the invention. The objects and advantages of the invention may be realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims. SUMMARY OF THE INVENTION In an electron beam device in which an electron beam is focused in a final image plane, there is provided a means for eliminating spherical and higher-order aberrations. Two sextupole electromagnetic lenses are positioned between two extreme focusing lenses and an intermediate focusing lens is positioned between the two sextupole electromagnetic lenses. The downstream sextupole electromagnetic lens is operated so as to have a lesser strength than the upstream electromagnetic lens. Interaction of the sextupoles and lenses compensates for third and higher order aberrations. |
abstract | A vol-oxidizer for spent nuclear fuel is provided. The vol-oxidizer includes a reactor module in which the spent nuclear fuel is put and oxidized; a heating module to heat the reactor module to a high temperature; a utility module to control an inner state of the reactor module, being connected with the reactor module; a support module to support the heating module and the reactor module; a drive module to supply a driving force for transferring the spent nuclear fuel within the reactor module; a valve module to control discharge of the spent nuclear fuel being oxidized in the reactor module; and a collecting container module to collect the spent nuclear fuel being oxidized and discharged from the reactor module. The respective modules are assembled and disassembled with respect to one another. |
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046879412 | summary | The invention relates to a glove box with tubing and/or hose lines for radioactive and/or toxic liquids or gases and with shutoff valves which are inserted into these tubing and/or hose lines and which, respectively, include an actuating member with a drive device and a valve body. Such a glove box is already known and in use. In this heretofore known glove box, not only the tubing and/or hose lines, but also the shutoff valves and the corresponding actuating members as well as the drive devices are arranged in the glove box. The liquids or gases in the tubing or hose lines are not only radioactive or toxic, but also frequently corrosive. Therefore, the drive devices, especially, which are generally electric or pneumatic servomotors, must be protected by special housings of corrosion-resistant material such as alloy steel, in order to prevent these drive devices from being damaged or even destroyed by escaping liquids or gases upon the occurrence of leaks in the tubing or hose lines. Further requirements are electric power supply and acknowledgement lines which lead to the drive devices in the glove box and also extend out of the glove box. These electric lines are not only subject to corrosion, but they also become embrittled under the influence of radioactive radiation and must therefore be constructed as corrosion and radiation-resistant cables with an insulating jacket of special insulating material. Furthermore, the bushings or lead-ins for these cables through the wall of the glove box must be sealed specially. It is accordingly an object of the invention to provide a glove box which avoids corrosion-sensitive parts which would otherwise require intensive servicing. With the foregoing and other objects in view, there is provided, in accordance with the invention, a glove box with tubular lines for harmful fluids and with shutoff valves inserted in the tubular lines and, respectively, formed of an actuating member and a valve body, comprising an exterior wall forming part of the glove box, the shutoff valves being disposed adjacent one another on the wall, the valve bodies of the shutoff valves being located on the inside of the exterior wall, and the actuating members of the valve bodies being located on the outside of the exterior wall; and drive means for driving the actuating members also located on the outside of the exterior wall and having a device for docking with the respective actuating members, the drive means being guidably displaceable, in a plane parallel to the exterior wall, to the respective actuating members for docking with the respective actuating members. The actuating members of the shutoff valves are brought to the outside through the wall of a glove box. The bushings or lead-ins for these actuating members can be made gas and liquid-tight without difficulty. Drive devices for these shutoff valves need not be located at all inside the glove box. Only a single drive device, for which no special corrosion-resistant housing is required, need be provided on the outside of the wall of the glove box for the drive members of a plurality of shutoff valves. The travel of the drive devices to the actuating member of the individual shutoff valves and the adjustment of these actuating members by this drive device can be controlled by an electronic control device in such a manner that the liquids or gases are distributed in accordance with a defined program to the individual tubing or hose lines in the glove box. If necessary, the actuating members of the shutoff valves can even be adjusted manually without requiring operating personnel to perform any work inside the glove box and without any endangerment from radioactive radiation. The connection of the individual shutoff valves to one another by the tubing or hose lines in the glove box can be effected with especially good visibility if, in accordance with another feature of the invention, the tubular lines include rectilinear main tubes whereon the valve bodies are seated, the main tubes extending spaced from and parallel to one another on the inner surface of the exterior wall. The main tubes may be in the form of feed lines or collecting tubes which are brought, for example, through the wall of the glove box to tanks located outside the glove box. In addition, the main tubes can be constructed as tube plates which are formed with bores and constitute a quite well-defined circulatory loop for the distribution of the liquids or gases flowing through therethrough. The shutoff valves can be assembled close together on these tube plates especially easily. In accordance with a further feature of the invention, the drive means comprise a sensor for detecting the setting of a respective actuating member and for adjusting the setting of the docking device for docking with the actuating member at the respective setting. In accordance with an added feature of the invention, the docking device comprises a sensor for detecting a proper docking of the drive means with the actuating member. In accordance with an additional feature of the invention, the fluids are radioactive. In accordance with a further feature of the invention, the fluids are toxic. In accordance with again another feature of the invention, the tubular lines are tubes. In accordance with a concomittant feature of the invention, the tubular lines are hose lines. 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 glove box, 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. |
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044903286 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention concerns a gas cooled high temperature nuclear reactor and a bottom shield therefor having a base plate, arranged under the reactor core and over the bottom of the prestressed concrete pressure vessel in a horizontal manner. 2. Background of the Prior Art The dead weight of the reactor core and of the bottom and side reflectors and the thermal side shield are supported by a bottom shield in one type of nuclear reactor in a concrete pressure vessel. The loads resulting from rod forces in the reactor and possible external forces such as earthquakes must be absorbed by the bottom shield and transmitted to the prestressed concrete pressure vessel. The bottom shield also serves to protect the bottom liner of the prestressed concrete pressure vessel against excessive nuclear radiation and temperatures. SUMMARY OF THE INVENTION It is an object of the invention to provide a gas cooled high temperature reactor with a bottom shield capable of functioning in a manner so that the above-cited thermal, nuclear and weight considerations are provided for as ideally as possible. Thus, any detrimental action by thermal,nuclear and weight forces are presented by an economical and simple construction of a bottom shield. The objects are attained according to the invention by a bottom shield comprising at least two plates arranged over and coaxially with respect to each other. These two plates will be referred to hereinafter as the first and second plates or as the upper and lower plates respectively. Each of the plates comprise several parts partially connected with each other with the lower plate being arranged at an axial and vertical distance, respectively, both from the bottom liner of the prestressed concrete pressure vessel and the upper plate. In advantageous embodimentsof the invention, the bottom shield may contain a first plate above the second plate with a second plate having a greater thickness than the thickness of the first plate. In addition, a means for spacing the first plate from the second plate may comprise a plurality of bearing supports. Similarly, a means for spacing the second plate from the bottom liner of the cavity in the concrete pressure vessel of a nuclear reactor installation advantageously comprises a plurality of bearing supports. The bearing supports between the bottom liner and the second plate are spaced at a greater distance from one another than the bearing supports between the first and second plates. The individual plate members are comprised of circular segments having overlapping areas which form the plate member seams. In the overlapping area, pins are positioned forming the seams of the interconnected plate members. An annular supporting ring formed of a plurality of interconnected supporting ring members surrounds the second or lower plate in an advantageous embodiment of the instant invention. The first or upper plate member may advantageously display a ribbed lower surface to aid in the circulation of cooling gas between the first and second plates. In addition, the upper plate may advantageously carry a plurality of sheet metal plates secured in spaced relationship to one another on the upper surface. As an advantageous alternative, the upper surface of the first plate may display a waffle like pattern. The bottom shield arranged in this manner including the support arrangement provides the essential advantage that the loads to be transmitted are transferred to the prestressed concrete pressure vessel in a uniform distribution. Furthermore, the distances of the upper and the lower bearing supports are chosen so that the different stiffnesses of the two bottom shield plates are balanced with respect to each other. Further advantages of the invention are described in the following. During the operation of the nuclear reactor according to the present invention, the temperature profile over the plate thicknesses is substantially reduced as the result of the division in a thin upper and a thick lower plate with a cooling gap provided between them. Because of the segmented construction with concentric partition gaps, whereby the segments are mechanically joined only within the inner half of the radius, the individual segments are able to expand freely in relation to each other, since the radial temperature profile may be maintained very small. The division of the lower plate into a circular plate and an annular ring further results in the fact that parts stressed differently by the temperature and the imposition of loads are allowed to move freely. By virtue of the special configuration of the top side of the upper plate, which may be provided both with a flat or a waffle structure, an additional bypass flow may be effected for cooling purposes. When the top side of the upper plate has a flat configuration and consists of individual cover plates with expansion gaps arranged between them, this may be used as thermal insulation and to facilitate the sliding of the graphite structures. |
description | This application claims priority to U.S. Provisional Application Ser. No. 61/933,952, filed Jan. 31, 2014, entitled “Apparatus And Method To Remotely Inspect Piping And Piping Attachment Welds.” 1. Field This invention pertains generally to nondestructive inspection tools and methods and more particularly to such tools and methods that can be employed for inspecting difficult to access locations such as around the welds on jet pumps of boiling water reactors. 2. Related Art A nuclear reactor produces electrical power by heating water in a reactor pressure vessel that contains a nuclear core of fissile material in order to generate steam which is used in turn to drive a steam turbine. A reactor pressure vessel of a boiling water reactor typically has a generally cylindrical shape and is closed at both ends, e.g., by a bottom head and a removable top head. A top guide typically is spaced above a core plate within the reactor pressure vessel. A core shroud, or shroud, typically surrounds the core and is supported by a shroud support structure. Particularly, the shroud has a generally cylindrical shape and surrounds both the core plate and the top guide. There is a space or annulus located between the cylindrical reactor pressure vessel and the cylindrically shaped shroud. In a boiling water reactor, hollow tubular jet pumps, positioned within the shroud annulus, provide the required reactor core water flow. The upper portion of the jet pumps, known as the inlet mixer, is laterally positioned and supported against two opposing rigid contacts within restrainer brackets by a gravity actuated wedge. The restrainer brackets support the inlet mixer by attaching to the adjacent jet pump riser pipe. The lower portion of the jet pump, known as the diffuser, is coupled to the inlet mixer by a slip joint. The slip joint between the jet pump inlet mixer and the jet pump diffuser collar has about 0.015 inch diametrical operating clearance which accommodates the relative axial thermal expansion movement between the upper and lower parts of the jet pump and permits leakage flow from the driving pressure inside the pump. The inlet mixer and the diffuser, due to their large size, are formed by welding a plurality of cylindrical sections together. Specifically, the respective ends of adjacent cylindrical sections are joined with a circumferential weld. During operation of the reactor, the circumferential weld joints may experience intergranular stress corrosion cracking and irradiation assisted stress corrosion cracking in the weld heat affected zones which can diminish the structural integrity of the jet pump. Various other components and structures in the nuclear reactor have experienced similar defects because of the harsh environment. Accordingly, it is important to examine periodically these components to assess their structural integrity and determine the need for repair. Ultrasonic inspection is a known technique for detecting cracks in nuclear reactor components. Many of the areas in a nuclear reactor that need to be inspected may have limited access and therefore, are difficult to assess using an inspection tool. For example, the jet pump riser pipe and elbow welds are periodically examined for cracking. The presence of cracking can diminish the structural integrity of a jet pump riser pipe and elbow and in extreme cases adversely impact reactor coolant flow. However, the jet pump riser pipe and elbow are difficult to access. Installation access is limited to the annular space between the outside of the shroud and the inside of the reactor pressure vessel, between adjacent jet pumps. Scanning operation access is additionally restricted within the narrow space between the jet pump riser pipe and vessel, shroud, or other welded attachments such as the riser brace or restrainer brackets. Furthermore, the inspection areas in a nuclear reactor can be highly radioactive and can pose safety risks for personnel working in these areas. Thus, inspection of these areas for the most part can require a robotic device which can be installed remotely and operated within the narrowly restricted space. Inspecting and repairing nuclear reactors, such as boiling water reactors, typically involves manually controlled poles and ropes to manipulate servicing devices and/or positioning of these devices. During reactor shutdown, servicing of some components requires installation of inspection manipulators or devices 30 to 100 feet deep within the reactor coolant. Relatively long durations are required to install or remove manipulators and can impact the plant shutdown duration. In addition, different inspection scopes can require several different manipulators or reconfigurations requiring additional manipulator installations and removals and costs. The long durations cannot only impact plant shutdown durations, but also increase personnel radiation and contamination exposure. Plant utilities have a desire to reduce the number of manipulator installations and removals to reduce radiological exposure as well as costs and plant outage impact. In addition, the plant utilities have a desire to reduce costs and operate as productively as possible. Thus, it is an object of this invention to minimize the number of reconfigurations and the number of tools required to perform inspections. In addition, it is a further object of this invention to provide a means to inspect difficult to access components such as jet pump riser piping areas that have previously been inaccessible with existing tooling. These and other objects are achieved by an automated inspection assembly that includes a number of subassemblies. One of the subassemblies comprises a frame subassembly having a first side and a second side and a length and width with the width substantially larger than a thickness of the frame subassembly between the first side and the second side. The length has a first end and a second end along a longitudinal dimension and the frame assembly is configured to form the main support structure for the automated inspection apparatus. A second subassembly is the positioning arm subassembly which is coupled to the frame subassembly and includes a support arm remotely operable to extend out from and retract toward the first side. A kicker arm is remotely operable to extend out from or retract toward the second side. The support arm and the kicker arm in the extended position are operable to wedge the frame subassembly between a member to be inspected and a surface opposed to the member to be inspected. A scanning subassembly is also supported from the frame subassembly and is configured to scan at least a portion of the member to conduct the desired nondestructive examination. In one embodiment, the automated inspection assembly further includes an orientation pivot subassembly that includes a rotational pivot joint coupling. The orientation pivot subassembly is connected to the first end of the frame subassembly. The orientation pivot subassembly is configured, through the rotational pivot joint coupling to orient the frame subassembly in one of two vertical positions with either the first end up or the second end up. In still another embodiment, the automated inspection assembly includes a lead in/gripper subassembly attached to the second end of the frame subassembly and configured to position the automated inspection assembly at the location to be inspected. Desirably, a scanning subassembly is supported from the second end of the frame subassembly and includes a substantially horseshoe-shaped scan head sized to receive the member to be scanned, a scan head wrist coupling joint, a linear hanger coupling joint and a linear drive box operable to move one or more transducers along the member to be inspected in a vertical, horizontal, radial, and a circumferential direction relative to the member, with the frame assembly in-line with the member. In still another embodiment, the support arm is formed in the general shape of a fork. When the member to be inspected is a pipe or a pipe elbow, preferably the support arm has a generally “U” shaped outer end configured to accept the pipe or pipe elbow within the opening of the “U”. Desirably, the support arm is configured to rotate out from the frame subassembly. Similarly, it is preferable that the kicker arm is configured to rotate out from the frame assembly. At least one of the kicker arm and the support arm is operable to move hydraulically. Preferably, tin one such embodiment, the support arm is operated to move hydraulically while the kicker arm is operated to move pneumatically. In such an arrangement, the support arm may comprise two support limbs spaced along the longitudinal dimension of the frame assembly with each of the limbs being operable to extend out from or retract towards the first side of the frame subassembly. Desirably, in such an arrangement, the outward ends of the kicker arm is positioned along the second side of the frame subassembly so as to engage the opposed surface at an elevation in between a first and second elevation on the member to be inspected that the two limbs respectively contact. Preferably, the scanning assembly houses at least one inspection device selected from a group of sensors comprising an ultrasonic transducer, eddy current transducer and video image capture device. The automated inspection assembly also preferably includes a mounting cup configured to remotely couple to the frame subassembly to vertically and horizontally position the automated inspection assembly relative to the member to be inspected. This invention also contemplates a method of inspecting a pipe comprising the steps of transporting a scanning assembly to the pipe; remotely wedging the scanning assembly between the pipe and an opposing surface to support the scanning assembly in a desired position; and scanning a surface of the pipe. The method may also include the steps of: positioning the scanning assembly at a desired location along the pipe; and extending opposing arms from opposite sides of the frame subassembly of the scanning assembly to contact both the pipe and the opposing surface. FIG. 1 is a sectional view, with parts cut away, of a boiling water nuclear reactor pressure vessel 10. The reactor pressure vessel 10 has a generally cylindrical shape and is closed at one end by a bottom head 12 and at its other end by a removable top head 14. A sidewall 16 extends from the bottom head 12 to the top head 14. Sidewall 16 includes a flange 18 upon which the top head 14 is sealed. A cylindrical-shaped core shroud 20 within the pressure vessel 10 surrounds a reactor core 22. Shroud 20 is supported at one end by a shroud support 24 and includes a removable shroud head 26 at the other end. An annulus 28 is formed between the shroud 20 and the sidewall 16. A pump deck 30, which has a ring shape, extends between the shroud support 24 and the reactor pressure vessel sidewall 16. The pump deck 30 includes a plurality of circular openings 32, with each opening housing a jet pump assembly 34. Jet pump assemblies 34 are circumferentially distributed around the core shroud 20. Heat is generated within the core 22 from fuel bundles 36 of fissionable material. Water circulated up through the core 22 is at least partially converted to steam. Steam separators 38 separate steam from water, which is recirculated. Residual water is removed from the steam by the steam dryers 40. The steam exits the reactor vessel 10 through a steam output nozzle 42 near the vessel top head 14. The amount of heat generated within the core 22 is regulated by inserting and withdrawing control rods 44 of neutron absorbing material, such as for example, hafnium. To the extent that the control rods 44 are inserted into the fuel bundles 36, they absorb neutrons that would otherwise be available to promote the fission chain reaction which generates the heat in the core 22. Control rod guide tubes 46 direct the vertical motion of the control rods 44 during insertion and withdrawal. Control rod drives 48 effect the insertion and withdrawal of the control rods 44. The control rod drives 48 extend through the bottom head 12. The fuel bundles 36 are aligned by a core plate 50 located at the base of the core 22. A top guide 52 aligns the fuel bundles 36 as they are lowered into the core 22. Core plate 50 and top guide 52 are supported by the core shroud 20. FIG. 2 is a perspective view of a portion of the reactor vessel and shroud, with parts cut away to show some of the details of a jet pump assembly 34. An inlet nozzle 54 extends through the sidewall 16 of the reactor pressure vessel 10 and is coupled to a jet pump assembly 34. Jet pump assembly 34 includes a thermal sleeve 56 that extends through the inlet nozzle 54, a lower elbow 55 (only partially visible in FIG. 2), and a riser pipe 58. The riser pipe 58 extends between and substantially parallel to the shroud 20 and reactor pressure vessel sidewall 16. Riser braces 60 stabilize riser pipe 58 within the reactor pressure vessel 10. Riser pipe 58 is coupled to jet pumps 62 by a transition assembly 64. Each jet pump 62 includes a jet pump nozzle 66, a suction inlet 68, and inlet mixer 70, and a diffuser 72. Jet pump nozzle 66 is positioned in the suction inlet 68 which is located at a first end 74 of the inlet mixer 70. Diffuser 72 is coupled to a second end 76 of the inlet mixer 72 by a slip joint 78. Because of their large size, both inlet mixer 70 and diffuser 72 are formed from multiple cylindrical sections. Circumferential weld joints 80 join the cylindrical sections together. FIG. 3 is a perspective view of one embodiment of an automated inspection assembly 82 within the purview of the claims set forth hereafter, that can be employed to access difficult to reach areas around the circumferential welds of the jet pump assembly, especially in areas between the jet pump assembly and the shroud. The embodiments of the automated inspection assembly 82 illustrated in FIG. 3-5 have a number of subassemblies including a frame subassembly 84, a positioning arm subassembly 86 (also referred to as the support arm assembly), a lead in or gripper subassembly 88, a mounting cup subassembly 90 and a scanning subassembly 92. The frame subassembly 84 serves as the main support structure for the apparatus and includes mounting features 94 and guide rails 96 for mounting and securing all the other subassemblies. The positioning arms subassembly 86 includes at least two spaced hydraulically actuated fork arms 98 each having a generally “U” shape with an opening between the tines of the fork large enough to at least partially fit around the pipe or elbow to be scanned. The dual forks 98 are actuated by the hydraulic cylinder 100 and can be detached from the rails 96 and replaced with forks with shorter or longer tines to accommodate different applications. The fork arms 98 (i.e., positioning arms) can rotate from a flat position substantially against the frame assembly 84 to at least an approximately vertical position. This can be seen from the side view in FIG. 4. A kicker arm 102 that extends from the opposite side of the frame subassembly 84 and is operated by a pneumatic cylinder 104 to rotate from a substantially flat position parallel to the frame subassembly 84 to a substantially vertical position as shown in the side view in FIG. 4, and works with the positioning arms 98 to support the inspection assembly 82. The kicker arm 102 may be part of the frame subassembly 84 or it may be part of the positioning arms subassembly and is preferably located in between the fork arms along the elongated dimension of the frame subassembly. The coordinated motions of the positioning arms subassembly 86 utilized for the forks 98 and the kicker arm 102 allows precise positioning of the apparatus frame subassembly 84 relative to the examination pipe, e.g., the riser 58. The pneumatic drive 104 for the kicker arm 102 provides a much softer operating force on the kicker arm that is spring like as compared to the more rigid hydraulic force 100 to the fork arms 98 which maintains the scanning head when fully deployed at a fixed distance from the member to be inspected, while the spring like movement of the kicker arm assures the automated inspection assembly remains wedged in position. The orientation pivot subassembly 106 includes a rotational pivot joint 108 attaching the orientation pivot subassembly to the frame subassembly 84. The swivel adaptor arm 110 with the junction box 112 is constrained with a quick release pin coupling 114 to change configurations (as shown in FIGS. 3 and 5). The arm 110 is symmetric for simple reconfiguration from a clockwise installation to a counter clockwise installation. This pivoting motion provides a means to orient the tool in an upright or upside down orientation to accommodate different examinations. The lead in or gripper subassembly 88 attaches to the frame subassembly 84, extends out under the scanning subassembly and serves the purpose of vertically positioning the apparatus by either resting on a pipe elbow, or gripping a pipe attachment. The lead in/side rails 130 attach to the frame and lead the tool onto the pipe elbow and support the weight of the tool. The lead in/side rails can be replaced with the gripper 128 which can be used to hang the automated inspection assembly 82 from the riser brace. A mounting cup assembly 116 shown in FIG. 8 can be remotely coupled to the frame subassembly 84 and is utilized to vertically position the inspection assembly 82 relative to the pipe end or transition piece of the pipe. The scanning subassembly 92 includes a horseshoe-shaped scan head 118 sized to receive the pipe, a scan head wrist coupling joint 120, a linear hanger coupling joint 122 and a linear drive box 124. The scan head 118 houses movably coupled transducers 126 and provides a means for moving these transducers along the pipe welds to perform inspections. The scanning subassembly 92 coupled to the frame subassembly 84 provides a means to move the transducers along the pipe in a vertical, horizontal, radial, and circumferential direction relative to the pipe or pipe elbow with the longitudinal direction of the frame oriented in-line with the pipe. The scan head wrist coupling joint 120 is hung from the X-axis and the scan head wrist coupling joint, linear hanger coupling joint 122 and linear drive box 124 enables 90 degree rotation and approximately 9.5 inch stroke in the Y direction (the direction of the longitudinal dimension of the frame assembly) and 1.25 inch stroke in the X direction (perpendicular to the flat surface of the frame assembly), which makes the automated inspection assembly very versatile. A more detailed understanding of the drive system of the scanning subassembly can be had by reference to the enlarged views shown in FIGS. 9 and 10. The motor 132 drives the sensors 126 around the track 134 on the inside surface of the generally “U” shaped scanning head 138 and the pneumatic cylinder 136 moves the sensors 126 radially so the sensors can come in close contact with the surface of the member to be inspected, where close contact is required. The motor 140 provides the wrist motion through the gear assembly 142 that enables the 90 degree rotation of the scanning head 138. The motor 144 through the gear assembly 146 and the linear hanger coupling joint 122 provides the 1.25 inch stoke in the X direction. Movement in the Y direction is achieved through the motor 148 and drive gear 150 which rides on a toothed track on the frame subassembly 84. The frame subassembly 84 coupled to the positioning arms subassembly 86, provides a means to position the frame within a variable annular gap between a pipe and a wall, e.g., the shroud, or pressure vessel, so the scanning subassembly drive system can finely position the scanning head 138 around the pipe or elbow and the sensors 126 over the portion of the member to be scanned. This unique design provides a means to perform remotely controlled automated piping inspections in limited access areas on a variety of welds and plant conditions. FIG. 5 shows the automated inspection assembly illustrated in FIG. 4 with the orientation pivot subassembly 106 pivoted in the opposite direction and a gripper assembly 128 in place of the lead in side rails on subassembly 88. The gripper assembly can be employed to attach to various features of the pipe attachment components to further support the inspection assembly 82. A side rail 130 can also be used to protect the scan head 118 and probes 126. FIG. 6 shows the lead in subassembly 88 positioned over an elbow 55 with the fork arms 98 pressured against the riser pipe 58 and the kicker arm 102 pressured against the vessel wall 16 to wedge the frame assembly 84 in a scanning position between the vessel wall and the riser pipe 58. FIG. 7 shows a front view of the cross section of the elbow shown in FIG. 6 showing a better view of the side rail 130 and scanning subassembly 92. FIG. 8 shows the inspection subassembly 82 wedged between the vessel wall 16 (not shown) and the inlet mixer 74 where the inspection subassembly can be lowered to inspect the circumferential welds 80 by rotating the scanning subassembly head down 90 degrees to extend around a portion of the surface of the weld to be scanned while the frame subassembly 84 is in-line with the inlet mixer 74. The invention also contemplates a method of inspecting a pipe employing the foregoing apparatus, comprising the steps of: transporting a scanning assembly to the pipe; remotely wedging the scanning assembly between the pipe and an opposing surface to support the scanning assembly in a desired position; and scanning a surface of the pipe. The method may also include the steps of positioning the scanning assembly at the desired location along the pipe and extending opposing arms from opposite sides of the frame subassembly of the scanning assembly to contact both the pipe and the opposing surface. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof. |
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abstract | A grazing incidence reflector (300) for EUV radiation includes a first mirror layer (310) and a multilayer mirror structure (320) beneath the first mirror layer. The first mirror layer reflects at least partially EUV radiation incident on the reflector with grazing incidence angles in a first range, and the first mirror layer transmits EUV radiation in a second range of incidence angles, which overlaps and extends beyond the first range of incidence angles. The multilayer mirror structure reflects EUV radiation that is incident on the reflector with grazing incidence angles in a second range that penetrates through the first mirror layer. A grazing incidence reflector can be used in a lithographic apparatus and in manufacturing a device by a lithographic process. |
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055725627 | abstract | Fine, sub-micron line features and patterns are created in a sensitized layer on a semiconductor wafer by a source of X-ray radiation. The X-ray source emits very low wavelength radiation along a path towards a sensitized surface of a semiconductor wafer. An image mask substrate is disposed in the path of the radiation, and is provided with a patterned opaque material on a surface of a substrate thereof. The substrate is formed of beryllium, which is robust and has a thermal coefficient of expansion closely conforming to that of common image mask carriers. Further, a wide variety of opaqueing materials adhere well to the beryllium substrate, and the substrate is relatively insensitive to moisture. The image mask is spaced sufficiently close to the wafer that radiation passing through the mask forms a corresponding pattern in the surface of the wafer. For X-ray radiation, the opaqueing material is gold, tungsten, platinum, barium, lead, iridium, rhodium, or the like. |
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description | The present invention is a continuation and claims priority of U.S. Ser. No. 10/249,931, filed May 20, 2003 now U.S. Pat. No. 6,968,030. The present invention relates generally to diagnostic imaging and, more particularly, to a method and apparatus of dynamically filtering radiation emitted toward a subject during radiographic imaging. Typically, in radiographic imaging systems, an x-ray source emits x-rays toward a subject or object, such as a patient or a piece of luggage. Hereinafter, the terms “subject” and “object” may be interchangeably used to describe anything capable of being imaged. The beam, after being attenuated by the subject, impinges upon an array of radiation detectors. The intensity of the attenuated beam radiation received at the detector array is typically dependent upon the attenuation of the x-rays. Each detector element of the detector array produces a separate electrical signal indicative of the attenuated beam received by each detector element. The electrical signals are transmitted to a data processing system for analysis which ultimately produces an image. In computed tomography (CT) imaging systems, the x-ray source and the detector array are rotated about a gantry within an imaging plane and around the subject. X-ray sources typically include x-ray tubes, which emit the x-rays as a beam at a focal point. X-ray detectors typically include a collimator for collimating x-ray beams received at the detector, a scintillator for converting x-rays to light energy adjacent the collimator, and a photodiode for receiving the light energy from an adjacent scintillator and producing electrical signals therefrom. Typically, each scintillator of a scintillator array converts x-rays to light energy. Each photodiode detects the light energy and generates a corresponding electrical signal. The outputs of the photodiodes are then transmitted to the data processing system for image reconstruction. There is increasingly a need to reduce radiation dosage projected toward a subject during an imaging session. It is generally well known that significant dose reduction may be achieved by using a “bowtie” filter to shape the intensity profile of an x-ray beam. Surface dose reductions may be as much as 50% using a bowtie filter. Generally, the bowtie filter is constructed such that the profile used to filter the radiation projected toward the subject is constant through the imaging process. That is, the bowtie filter is constructed to filter radiation emitted toward a subject such that the sum of the radiation path length through the filter plus the path length through the subject is roughly constant. One drawback of this construction is that a subject's shape is not constant. For example, a patient's shape changes as a function of view angle as the x-ray tube and detector rotate around the patient and because the filter has a constant filtering profile at different points in the imaging process, portions of the subject may be over-radiated whereas at other times, other portions will be under-radiated. As a result, a choice must be made between having under-radiated portions during the imaging process or increasing the dosage such that a minimum radiation dosage throughout the imaging session is maintained with some portion being over-radiated. A number of developments have been made to address this situation. The CT scanner may be equipped with multiple beam shaping filters. Multiple filters, however, increases the manufacturing costs of the CT scanner as well as requiring additional controls to insure the appropriate filters are being used during the scan session. Another approach is to dynamically control the amount of x-ray dosage emitted by the projection source as a function of view angle. This approach, however, requires significant and complex controls to alter the amount of radiation projected toward the subject during the scan that significantly increases the costs of the CT system nor takes into account shape of the pre-patient filter. Therefore, it would be desirable to design a pre-subject filter having a filter that rotates and effectively changes its filtering and been shaping properties as a function of view angle as the x-ray tube and detector rotate around a subject. The present invention is a directed method and apparatus for CT data acquisition using a rotatable pre-subject filter having multiple filtering profiles to dynamically control radiation exposure to a subject that overcomes the aforementioned drawbacks. Specifically, a filter is disclosed that is shaped to have varying filtering profiles. The filter is designed to rotate about its lengthwise axis such that the filtering profile used may be varied as a function of the view angle of the projection source. Therefore, in accordance with one aspect, the present invention includes a pre-subject filter assembly for a CT imaging system having a detector assembly and a high frequency electromagnetic energy projection source configured to rotate about a subject during an imaging session. The filter assembly includes a beam shaping filter having a generally cylindrical shaped body defining multiple filtering profiles. The filter is constructed to rotate about an axis of rotation that extends through the body. A controller is provided to cause rotation of the filter about the axis of rotation during an imaging session to dynamically filter high frequency electromagnetic energy projected toward the subject as a function of view angle. In accordance with another aspect, a CT system includes a rotatable gantry having an opening to receive a subject to be scanned. A moveable high frequency electromagnetic energy projection source projects high frequency electromagnetic energy such as x-rays toward the subject at least two view angles during an imaging session. A pre-subject includes a filter that is rotatable about itself relative to an axis of rotation extending the filter during the imaging session. The filter includes multiple filtering profiles. As such, the filter moves circumferentially around the subject during the imaging session but also rotates or spins relative to itself during the imaging session such that multiple filtering profiles are usable during the imaging session. The CT system further includes a scintillator array and a photodiode array that rotate synchronously with the filter and x-ray source during the imaging session. The scintillator and photodiode arrays collectively form a detector array that detects x-rays attenuated by the subject during the imaging session and outputs electrical signals indicative of the attenuation. The detector output is then input to a DAS configured to reconstruct an image of the subject. The CT system further includes a computer programmed to rotate the filter about the axis of rotation such that a first view angle a first filtering profile is presented and at a second view angle a second filtering profile is presented. In accordance with yet another aspect of the present invention, a method of reducing x-ray exposure during CT data acquisition includes positioning a subject to be scanned in a scanning bay and projecting x-rays toward the subject from an x-ray source at a first view angle. A first profile of a multi-profile filter is then positioned between the x-ray source and the subject when the x-ray source is projecting x-rays at the first view angle. The x-ray source is then rotated to a second view angle and projects x-rays toward the subject therefrom. Simultaneously therewith, the multi-profile filter is caused to rotate about an axis of rotation that extends through a length of the filter such that a second profile is positioned between the x-ray source and the subject when the x-ray source is projecting x-rays from the second view angle. Various other features, objects and advantages of the present invention will be made apparent from the following detailed description and the drawings. The present invention is described with respect to a radiographic imaging system such as the CT system shown in FIGS. 1–2. However, it will be appreciated by those skilled in the art that the present invention is equally applicable for use with other radiographic imaging systems utilizing a rotating projection source and a pre-subject filter Moreover, the present invention will be described with respect to the emission and detection of x-rays. However, one skilled in the art will further appreciate, that the present invention is equally applicable for the emission and detection of other high frequency electromagnetic energy. Referring to FIGS. 1 and 2, a “third generation” CT imaging system 10 is shown as including a gantry 12. The present invention, however, is applicable with other CT systems. Gantry 12 has an x-ray source 14 that projects a beam of x-rays 16 through filter 15 toward a detector array 18 on the opposite side of the gantry 12. Detector array 18 is formed by a plurality of detectors 20 which together sense the projected x-rays that pass through a medical patient 22. Each detector 20 produces an electrical signal that represents the intensity of an impinging x-ray beam and hence the attenuated beam as it passes through the patient 22. During a scan to acquire x-ray projection data, gantry 12 and the components mounted thereon rotate about a center of rotation 24. Rotation of gantry 12 and the operation of x-ray source 14 are governed by a control mechanism 26 of CT system 10. Control mechanism 26 includes an x-ray controller 28 that provides power and timing signals to an x-ray source 14, a gantry motor controller 30 that controls the rotational speed and position of gantry 12, and filter controller 33 that controls filter 15. As will be described in greater detail below, filter 15 is rotated or caused to spin about a lengthwise axis of rotation by motor assembly 17. A data acquisition system (DAS) 32 in control mechanism 26 samples analog data from detectors 20 and converts the data to digital signals for subsequent processing. An image reconstructor 34 receives sampled and digitized x-ray data from DAS 32 and performs high speed reconstruction. The reconstructed image is applied as an input to a computer 36 which stores the image in a mass storage device 38. Computer 36 also receives commands and scanning parameters from an operator via console 40 that has a keyboard. An associated cathode ray tube display 42 allows the operator to observe the reconstructed image and other data from computer 36. The operator supplied commands and parameters are used by computer 36 to provide control signals and information to DAS 32, x-ray controller 28 and gantry motor controller 30. In addition, computer 36 operates a table motor controller 44 which controls a motorized table 46 to position patient 22 and gantry 12. Particularly, table 46 moves portions of patient 22 through a gantry opening 48. As shown in FIGS. 3 and 4, detector array 18 includes a plurality of scintillators 57 forming a scintillator array 56. A collimator (not shown) is positioned above scintillator array 56 to collimate x-ray beams 16 before such beams impinge upon scintillator array 56. In one embodiment, shown in FIG. 3, detector array 18 includes 57 detectors 20, each detector 20 having an array size of 16×16. As a result, array 18 has 16 rows and 912 columns (16×57 detectors) which allows 16 simultaneous slices of data to be collected with each rotation of gantry 12. Switch arrays 80 and 82, FIG. 4, are multi-dimensional semiconductor arrays coupled between scintillator array 56 and DAS 32. Switch arrays 80 and 82 include a plurality of field effect transistors (FET) (not shown) arranged as multi-dimensional array. The FET array includes a number of electrical leads connected to each of the respective photodiodes 60 and a number of output leads electrically connected to DAS 32 via a flexible electrical interface 84. Particularly, about one-half of photodiode outputs are electrically connected to switch 80 with the other one-half of photodiode outputs electrically connected to switch 82. Additionally, a reflector layer (not shown) may be interposed between each scintillator 57 to reduce light scattering from adjacent scintillators. Each detector 20 is secured to a detector frame 77, FIG. 3, by mounting brackets 79. Switch arrays 80 and 82 further include a decoder (not shown) that enables, disables, or combines photodiode outputs in accordance with a desired number of slices and slice resolutions for each slice. Decoder, in one embodiment, is a decoder chip or a FET controller as known in the art. Decoder includes a plurality of output and control lines coupled to switch arrays 80 and 82 and DAS 32. In one embodiment defined as a 16 slice mode, decoder enables switch arrays 80 and 82 so that all rows of the photodiode array 52 are activated, resulting in 16 simultaneous slices of data for processing by DAS 32. Of course, many other slice combinations are possible. For example, decoder may also select from other slice modes, including one, two, and four-slice modes. As shown in FIG. 5, by transmitting the appropriate decoder instructions, switch arrays 80 and 82 can be configured in the four-slice mode so that the data is collected from four slices of one or more rows of photodiode array 52. Depending upon the specific configuration of switch arrays 80 and 82, various combinations of photodiodes 60 can be enabled, disabled, or combined so that the slice thickness may consist of one, two, three, or four rows of scintillator array elements 57. Additional examples include, a single slice mode including one slice with slices ranging from 1.25 mm thick to 20 mm thick, and a two slice mode including two slices with slices ranging from 1.25 mm thick to 10 mm thick. Additional modes beyond those described are contemplated. Referring now to FIG. 6, a statically shaped pre-patient filter 15 is shown in perspective as being connectable to a motor or other drive assembly 17 via a shaft 86 that is operationally connected to the filter 15 via end cap or plate 88. End cap 88 is designed to be adhesively connected to one end 90 of filter 15. A similarly shaped end cap or plate 92 is connected to an opposite end 94 of the filter. The operational significance of the shaft and end caps will be described in greater detail with respect to FIG. 9. The shape of the filter is static or constant, but the shape defines multiple filtering profiles, as hereinafter described. Still referring to FIG. 6, filter 15 has a general bowtie or hourglass shape. In this regard, each of the ends 90, 94 of filter 15 has a diameter that is greater than the diameter of the body 96 that extends between each of the ends. Body 96, however, is uniquely designed to have a depression defined by surface 98 as well as a second depression defined by surface 100. In the embodiment illustrated in FIG. 6, each of the depressions is positioned orthogonally from one another. The orthogonal placement of the depressions illustrates only one example of how the body 96 of filter 15 may be constructed. As such, the depressions may be formed in the body 96 of filter 15 so as to be more acutely or obtusely defined with respect to one another. Further, the body may be constructed to have multiple depressions including more than two depressions. As will be discussed in greater detail below, the depressions define the shape of filter 15 such that a multitude of filtering profiles is possible. Refening now to FIG. 7, a cross-sectional view of filter 15 illustrating a first filtering profile is shown. It is readily apparent in FIG. 7 that the body 96 of filter 15 extending between ends 90 and 94 includes a relatively steep depression defined by surface 100. In this regard, a minimum body diameter is formed at the base 102 of the depression. The filtering profile of filter 15 that results from the orientation illustrated in FIG. 7 is particularly well suited for pre-patient filtering of an x-ray beam that is being projected toward the patient from a projection source generally positioned at a side of the patient, i.e. when the patient has the thickest cross-section. That is, since a patient generally has a thicker cross-section in the x-direction than in the y-direction, when lying on a patient table, filter 15 is designed such that the filtering profile illustrated in FIG. 7 is presented when x-rays are projected toward a thicker patient cross-section. Simply, filter 15 is designed such that x-rays received at surfaces 104 of filter 15 undergo greater filtering than the x-rays received at surfaces98, 100. This variation in filtering power throughout the filter ensures that the sum of the x-ray beam lengths is uniform at the x-ray detector array interface. Referring now to FIG. 8, another cross-sectional view of filter 15 is shown that is orthogonal to the cross-sectional view illustrated in FIG. 7. In this regard, the filtering profile of filter 15 is defined by surface 98. Similar to the filtering profile illustrated in FIG. 7, the filtering profile of FIG. 8 also has a relative bowtie or hourglass shape, but the depression formed by surface 98 is less steep than the depression illustrated in FIG. 7. In this regard, the filtering profile illustrated in FIG. 8 is particularly well suited for presentation in the x-ray fan beam path when the x-ray source is projecting x-rays toward a patient from a position generally above the patient, i.e. when the patient has the thinnest cross-section. Because patients are generally flat and thin in the x-direction when positioned face-up on a patient table, the filtering profile illustrated in FIG. 8 is also relatively flat. That is, filter 15 is designed to provide less varying filtering across its width but less filtering of the fan beam when the filtering profile illustrated in FIG. 8 is presented in the x-ray beam path. Refening now to FIG. 9, a schematic representation of the patient filter 15 is shown connected to a motor or other drive assembly 17. As indicated previously, filter 15 is positioned within the gantry of the CT system and is designed to rotate with the gantry around the patient or other scan subject during data acquisition. However, filter 15 is also constructed to spin or rotate about its lengthwise axis generally referenced 106 such that more than one filtering profile, such as those illustrated in FIGS. 7 and 8, may be presented in the x-ray beam path as the filter rotates around the patient. Accordingly, motor 17 is connected to an end cap or plate 88 that is affixed to an end of filter 15. On an opposite end, plate 92 is connected to the filter and also has a shaft 108 that is connected to a bearing assembly 110. Motor 17 is designed to rotate shaft 86 in conjunction with a bearing assembly 110 so that filter 15 is caused to spin or rotate about lengthwise axis 106 as a function of view angle of the x-ray projection source. In this regard, motor 17 may be controlled by a filter controller 33, FIG. 2, such that a steeper filtering profile, FIG. 7, is presented when the x-ray projection source is at a side of the patient. The motor may then cause filter 15 to spin or rotate about its lengthwise axis 106 such that the filtering profile illustrated in FIG. 8 is presented when the x-ray projection source is projecting x-rays from a position above the patient. One skilled in the art would readily appreciate that an encoder or other device may be incorporated with the motor such that the rotation of the gantry and the spinning of filter 15 may be synchronized. Motor 17 may also be controlled such that the filter 15 is caused to rotate or spin along its lengthwise axis uniquely for each scan session. That is, a pre-scan may be undertaken whereupon a profile of the patient is acquired and used to input data into a computer program or algorithm to control motor 17 such that the filtering profiles used throughout the imaging session are presented according to the particular contour of the particular patient undergoing the imaging session as opposed to a general control algorithm, or synchronized with table translation. Additionally, the motor may incrementally spin or rotate the filter about its lengthwise axis or provide a timed and smooth continuous rotation as a function of view angle. Referring now to FIG. 10, package/baggage inspection system 200 includes a rotatable gantry 202 having an opening 204 therein through which packages or pieces of baggage may pass. The rotatable gantry 202 houses a high frequency electromagnetic energy source 206, a pre-subject filter 207, as well as a detector assembly 208. A conveyor system 210 is also provided and includes a conveyor belt 212 supported by structure 214 to automatically and continuously pass packages or baggage pieces 216 through opening 204 to be scanned. Objects 216 are fed through opening 204 by conveyor belt 212, imaging data is then acquired, and the conveyor belt 212 removes the packages 216 from opening 204 in a controlled and continuous manner. As a result, postal inspectors, baggage handlers, and other security personnel may non-invasively inspect the contents of packages 216 for explosives, knives, guns, contraband, etc, Therefore, in accordance with one embodiment of the present invention, a beam shaping filter for a CT imaging system includes a first end and a second end, each of which has a circular cross-section. The filter further includes a body formed of radiation beam shaping material that extends between the first end and the second end. A pair of depressions is formed in the body orthogonal from one another such that a first body diameter and a second body diameter are defined. In accordance with another embodiment of the present invention, a CT system includes a rotatable gantry having an opening to receive a subject to be scanned as well as a rotatable high frequency electromagnetic energy projection source configured to project a high frequency electromagnetic energy beam toward the subject at af least two view angles. The rotatable pre-subject filter is also provided and has a static shape that defines at least two filtering profiles such that at a first view angle a first filtering profile filters the high frequency electromagnetic energy beam and at a second view angle a second filtering profile filters the high frequency electromagnetic energy beam. The CT system further includes a scintillator array having a plurality of scintillator cells wherein each cell is configured to detect high frequency electromagnetic energy passing through the subject as well as a photodiode array optically coupled to the scintillator array and comprising a plurality of photodiodes configured to detect light output from a corresponding scintillator cell. A data acquisition system is provided and connected to the photodiode array and configured to receive photodiode outputs which are received by an image reconstructor that is configured to reconstruct an image of the subject. According to another embodiment of the present invention, a method of reducing x-ray exposure during CT data acquisition includes the steps of positioning a subject to be scanned in a scanning bay and projecting x-rays toward the subject from an x-ray source at a first view angle. The method also includes positioning a first profile of a multi-profile filter between x-ray source and the subject when the x-ray source is projecting x-rays at the first view angle. The method further includes the steps of spinning an x-ray source to a second view angle and projecting x-rays toward the subject from the x-ray source at the second view angle. A second profile of the multi-profile filter is then positioned between an x-ray source and the subject when the x-ray source is projecting x-rays at the second view angle. In another embodiment, the present invention includes a pre-subject filter assembly for a CT imaging system having a detector assembly and a high frequency electromagnetic energy projection source configured to rotate about a subject during an imaging session. The filter assembly includes a beam shaping filter having a generally cylindrical shaped body defining multiple filtering profiles. The filter is constructed to rotate about an axis of rotation that extends through its body. A controller is provided to cause rotation of the filter about the axis of rotation during an imaging session to dynamically filter high frequency electromagnetic energy projected toward the subject as a function of view angle. In another embodiment, a CT system includes a rotatable gantry having an opening to receive a subject to be scanned. A moveable high frequency electromagnetic energy projection source projects high frequency electromagnetic energy such as x-rays toward the subject at af least two view angles during an imaging session. A pre-subject filter includes a filter that is rotatable about itself relative to an axis of rotation extending the filter during the imaging session. The filter includes multiple filtering profiles. As such, the filter moves circumferentially around the subject during the imaging session but also rotates or spins relative to itself during the imaging session such that multiple filtering profiles are usable during the imaging session. The CT system further includes a scintillator array and a photodiode array that rotate synchronously with the filter and x-ray source during the imaging session. The scintillator and photodiode arrays collectively form a detector array that detects x-rays attenuated by the subject during the imaging session and outputs electrical signals indicative of the attenuation. The detector output is then input to a DAS configured to reconstruct an image of the subject. The CT system further includes a computer programmed to rotate the filter about the axis of rotation such that a first view angle a first filtering profile is presented and at a second view angle a second filtering profile is presented. In another embodiment of the present invention, a method of reducing x-ray exposure during CT data acquisition includes positioning a subject to be scanned in a scanning bay and projecting x-rays toward the subject from an x-ray source at a first view angle. A first profile of a multi-profile filter is then positioned between the x-ray source and the subject when the x-ray source is projecting x-rays at the first view angle. The x-ray source is then rotated to a second view angle and projects x-rays toward the subject therefrom. Simultaneously therewith, the multi-profile filter is caused to rotate about an axis of rotation that extends through a length of the filter such that a second profile is positioned between the x-ray source and the subject when the x-ray source is projecting x-rays from the second view angle. The present invention has been described in terms of the preferred embodiment, and it is recognized that equivalents, alternatives, and modifications, aside from those expressly stated, are possible and within the scope of the appending claims. |
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claims | 1. A method for producing 99mTc, comprising:providing an aqueous solution comprising dissolved 100Mo-molybdate ions,providing a proton beam having an energy suitable for inducing a 100Mo(p,2n)99Tc nuclear reaction when 100Mo-molybdate ions are irradiated,irradiating the aqueous solution with the proton beam and inducing a 100Mo(p,2n)99Tc nuclear reaction,applying a solvent extraction method to dissolve the 99Tc into solution with the solvent and separate a resulting 99Tc-loaded solvent solution from the aqueous solution containing dissolved 100Mo-molybdate ions. 2. The method of claim 1, wherein the extraction method comprises a solvent extraction method using methyl ethyl ketone. 3. The method of claim 1, comprising recycling dissolved 100Mo-molybdate ions remaining in solution after the 99Tc extraction to additional aqueous solution to be irradiated. 4. The method of claim 1, wherein the aqueous solution with dissolved 100Mo-molybdate ions is an aqueous solution of a 100Mo-molybdate salt, and wherein a nuclear reaction which leads to at least one cation end product is induced in the aqueous solution by irradiating cations of the 100Mo-molybdate salt in solution with the proton beam. 5. The method of claim 4, comprising:after extracting the 99Tc, recycling the remaining solution containing dissolved 100Mo-molybdate ions to additional aqueous solution; andremoving the at least one cation end product before returning to the aqueous solution. 6. The method of claim 4, comprising after extracting the 99Tc from the aqueous solution, cleansing the extracted 99mTc of impurities resulting from the nuclear reaction which leads to at least one cation end product. 7. The method of claim 4, wherein the 100Mo-molybdate salt comprises 6Li2100MoO4, and wherein the at least one cation end product comprises 3H. 8. The method of claim 4, wherein the 100Mo-molybdate salt comprises Na2100MoO4, and wherein the cation end product comprises 18F. 9. The method of claim 4, wherein the 100Mo-molybdate salt comprises K2100MoO4, and wherein the cation end product comprises Ca ions. 10. The method of claim 4, comprising after extracting the 99mTc, returning the remaining dissolved 100Mo-molybdate ions to the aqueous solution and removing the at least one cation end product using an ion exchanger. 11. The method of claim 4, comprising after extracting the 99Tc from the aqueous solution, using an ion exchanger to cleanse the extracted 99Tc of impurities resulting from the at least one cation end product. |
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claims | 1. A method of providing a neutron source, the method comprising:providing a nuclear reactor neutron source, the nuclear reactor neutron source comprising:an enclosure delimiting a chamber,a nuclear reactor core arranged inside the chamber,the nuclear reactor core is configured to produce neutrons from a nuclear fuel element inside the nuclear reactor core;installing a beam generator arranged to generate a beam directed into the chamber;wherein the beam generator forms part of a spallation neutron source for providing neutrons;removing the nuclear fuel element of the nuclear reactor core from the chamber,wherein removing the nuclear fuel element forms a void inside the chamber;andinstalling, inside the void, a target configured to eject neutrons upon impact of the beamwherein impact by the beam onto the target causes neutrons to be ejected from the target and emitted from the chamber through a neutron beam outlet arranged through the enclosure,wherein the neutron beam outlet provides a passage for neutrons to leave the enclosure. 2. The method according to claim 1, wherein the chamber of the provided nuclear reactor neutron source further comprises a moderator, and wherein the act of installing the target comprises arranging the target inside the moderator. 3. The method according to claim 2, wherein the moderator of the provided nuclear reactor neutron source comprises a material selected from the group consisting of H2O, D2O, liquid or solid hydrogen or deuterium, liquid or solid methane, mesithelene, and ice. 4. The method according to claim 1, wherein the chamber of the provided nuclear reactor neutron source further comprises a reflector, and wherein the act of installing the target comprises arranging the target inside the reflector. 5. The method according to claim 4, wherein the reflector of the provided nuclear reactor neutron source comprises a material selected from the group consisting of graphite, beryllium, steel, tungsten carbide nickel, tungsten, heavy water, lead, and alloys of these. 6. The method according to claim 1, wherein the provided nuclear reactor neutron source further comprises an additional neutron beam outlet arranged through the enclosure to provide an additional neutron passage for neutrons from the chamber, and wherein the act of installing of the beam generator comprises directing the beam into the chamber via the additional neutron passage onto the target. 7. The method according to claim 1, wherein the enclosure of the provided nuclear reactor neutron source further comprises a thermal column or an access shaft, and wherein the act of installing of the beam generator comprises directing the beam into the chamber via the thermal column or the access shaft onto the target. 8. The method according to claim 1, wherein the act of installing a beam generator comprises installing the beam generator or a portion of the beam generator inside the chamber. 9. The method according to claim 1, wherein the act of installing a target comprises installing a target comprising a material selected from the group consisting of mercury, tantalum, lead, liquid lead-bismuth alloy, tungsten, rhenium, alloys of these, or beryllium and lithium. 10. The method according to claim 1, wherein the installing a beam generator arranged to generate a beam directed into the chamber comprises installing the beam generator to form part of a spallation neutron source for providing neutrons. 11. The method according to claim 1, wherein the installing a beam generator arranged to generate a beam directed into the chamber comprises installing the beam generator to form part of a compact neutron source for providing neutrons. 12. The method according to claim 1, wherein the chamber of the provided nuclear reactor neutron source further comprises a hot or cold neutron source, and wherein the act of installing a target comprises arranging the target adjacent to the hot or cold neutron source. |
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claims | 1. A pressure relief system for a containment in a nuclear power facility, the pressure relief system comprising:a shut-off valve;a pressure relief line guided through the containment and being closed by said shut-off valve, said pressure relief line conducting a pressure relieving flow out of the containment when said shut-off valve is open;a wet scrubber disposed in a portion of said pressure relief line inside the containment, wherein said wet scrubber is a Venturi scrubber having a Venturi tube disposed in a washing fluid tank, said Venturi tube having a throat point with a suction port for Venturi injection of the washing fluid into the pressure relief flow;a reservoir disposed inside the containment and fluidically connected to an inner space of the containment such that any overpressure, with respect to surroundings outside the containment, prevailing in the containment is transferred at least in part to said reservoir;a supply line leading from said reservoir to said wet scrubber for supplying said wet scrubber with a fluid, acting as a washing fluid, from said reservoir;said wet scrubber having a washing fluid tank sealed off from an inner space of the containment;a throttle portion formed in a portion of said pressure relief line which is upstream of a point at which said pressure relief line opens inside said washing fluid tank; andwherein said wet scrubber has an inner tank inside said washing fluid tank for the washing fluid, said inner tank surrounding said Venturi tube only in a region of said throat point and said suction port of said Venturi tube, a capacity of said inner tank is substantially smaller than a capacity of said washing fluid tank. 2. The pressure relief system according to claim 1, wherein said reservoir is a pool which is open to the inner space of the containment. 3. The pressure relief system according to claim 1, further comprising a filling level control means for the washing fluid supplied by said supply line, said filling level control means being a float valve inside said washing fluid tank. 4. The pressure relief system according to claim 1, wherein said throttle portion is configured such that, at a start of a pressure-relieving process and at a pressure of at least 2 bar absolute inside the containment, a pressure inside said washing fluid tank is, in contrast, reduced by at least 0.3 bar. 5. The pressure relief system according to claim 1, further comprising a throttle valve disposed in said pressure relief line downstream of said wet scrubber when viewed in a direction of flow of the pressure relief flow, and is configured for critical pressure relief of the pressure relief flow. 6. The pressure relief system according to claim 1, wherein said reservoir is a sump of a pressurized water reactor. 7. The pressure relief system according to claim 1, wherein said wet scrubber is kept dry when in a standby mode by a sealing device configured to open under accident conditions. 8. A nuclear power facility, comprising:a containment; andthe pressure relief system according to claim 1. 9. A method for operating a pressure relief system, including the pressure relief system of claim 1, the method comprising the steps of:during a pressure-relieving process, when the shut-off valve is open, conveying the fluid by means of the supply line from the reservoir to the wet scrubber as a result of a pressure differential, which adjusts according to a flow, between the containment and the wet scrubber. |
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046876251 | claims | 1. Piping of a nuclear reactor containment vessel comprising: pipes in the interior of the containment vessel passing through a wall of the containment vessel and extending to the outside of the containment vessel; that portion of said pipes inside said containment vessel having no shut-off valve therein; outer shut-off valves each provided at a portion of the pipe which is located at the outside of the containment vessel; and isolating vessel means for covering a portion of the pipe extending from the containment vessel up to and including a portion of the pipe where the outer shut-off valve is located for isolating the covered portions of the pipe from atmosphere, thereby eliminating a requirement for an inner shut-off valve on the portion of the pipe in the interior of the containment vessel and enabling a reduction in size of the containment vessel. 2. Piping of a nuclear reactor containment vessel, as claimed in claim 1, wherein the containment vessel contains a boiling water nuclear reactor pressure vessel, and the pipes having an outer shut-off valve at the portion of the pipe located outside the containment vessel pass into the interior of the containment vessel without inner shut-off valves being provided therefor. 3. Piping as claimed in claim 1, wherein said isolating vessel means include bellows constructions having flexibility. 4. Piping as claimed in claim 3, wherein said isolating vessel means include steam leak detectable transducers. 5. Piping as claimed in claim 1, wherein one of said pipes in a main steam pipe and one of said outer shut-off valves is a main steam shut-off valve. 6. Piping as claimed in claim 5, wherein said isolating vessels include steam leak detectable transducers. |
description | The present application is a continuation of U.S. patent application Ser. No. 10/219,765, filed on Aug. 16, 2002, the entire disclosure of which is incorporated herein by reference, claiming priority on Japanese Patent Applications No. 2001-253752, filed on Aug. 24, 2001 and No. 2002-62817, filed on Mar. 8, 2002, the entire contents of which are incorporated herein by reference. The present invention relates to methods for evaluating microscope images such as from scanning electron microscopes and scanning ion microscopes and relates in particular to a method for evaluating image resolution and a microscope having an evaluation function. In charged particle microscopes such as scanning electron microscopes (SEM) and scanning ion microscopes (SIM), the related techniques for evaluating image resolution are roughly divided into two methods. One is the gap method disclosed in JP-A-45265/1993 for evaluating image resolution by viewing a microscope image of gold particles deposited by evaporation onto a carbon substrate as a specimen and then finding the minimum visually discernible gap separating the particles at two points as seen on the monitor display of the microscope. The other is the FFT method as disclosed in JP-A-24640/1999 for evaluating image resolution by frequency analysis of observation image data that was subjected to two-dimensional Fourier transform (FFT). The first gap method of the related art has the problem that high accuracy and repeatability of image resolution evaluation values cannot be expected because the size and shape of the gold particles to be imaged are not uniform and because the human evaluator may not be objective when visually determining the minimum gap between two particles. The second FFT method of the related art also has the problem that resolution is visually evaluated with a resolution determination curve plotted on a power spectrum graph for frequency analysis, so subjectivity on the part of the evaluator still cannot be eliminated. In the semiconductor manufacturing process in particular, multiple microscopes are used for long periods during the work process for quality control of semiconductor devices. But image resolution is still evaluated by the methods of the related art, so the differential (instrumental error) in image resolution between microscopes and changes in resolution over time cannot be accurately controlled, causing problems making it impossible to minimize variations in device quality in the device manufacturing process. The present invention has the object of providing an image evaluation method for objectively determining image resolution of microscope images, and a microscope having a resolution evaluation function. To achieve the above objects, the present invention provides an image evaluation method for evaluating image resolution, wherein the resolution in partial regions of the image is obtained over an entire area or a portion of the image, averaging is performed over an entire area or a portion of the image, and the averaged value becomes the resolution evaluation value for an entire area or a portion of the image. Other objects and examples of the present invention are described in the detailed description of the preferred embodiments. Preferred embodiments of the present invention are next described in detail based on the following drawings. The outline of a scanning electron microscope (called SEM hereafter) is explained below as a preferred embodiment of the present invention. The electron lens system in FIG. 1 focuses a charged particle beam (electron beam) 2 emitted from a charged particle source (electron gun) 1 onto a specimen 4 by an electron lens 3 and scans the beam over the specimen 4 in any desired sequence. Secondary particles (secondary electrons) 5 generated from the surface of the specimen 4 by irradiation of the electron beam are detected by a secondary particle detector system 6. The output of the detector system 6 is input as image data to a control system 7 (control processor) having an image processing and control function. The specimen 4 can be moved along in any direction in the three dimensions by use of an X-Y-Z stage 8. The control system 7 also controls the charged particle source (electron gun) 1, the electron lens 3, the secondary particle detector system 6, X-Y-Z stage 8, and an image display device 9. In this embodiment, the electron beam 2 is two-dimensionally scanned over the surface of the specimen 4 with a scanning coil (not shown in drawing). The signal detected with a secondary electron detector in the secondary particle detector system 6 is transferred to an image memory after being amplified by a signal amplifier in the control system 7 and is finally displayed as an image of the specimen on the image display device 9. The secondary signal detector may be a secondary electron or reflected electron detector, photodetector or X-ray detector. The address signals corresponding to the memory positions in the image memory are generated in the control system 7 or a separately installed computer, and supplied to a scanning coil after being converted into analog signals. When for example, the image memory has 512×512 pixels, the X-direction address signal is a digital signal representing 0 to 512 repeatedly. The Y-direction address signal is also a digital signal representing 0 to 512 repeatedly but is incremented by 1 each time the X-direction address signal reaches 512 from 0. These digital signals are converted into analog signals. Since the addresses in the image memory correspond to addresses of deflection signals for scanning the electron beam, a two-dimensional image in the electron beam deflection range determined by the scanning coil is stored in the image memory. The signals in the image memory can be sequentially read out in time series by a readout address generator circuit synchronized with a readout clock. The signals that were read out according to each address are converted into an analog signal and input to the image display device 9 as a brightness-modulated signal. The image memory has a function for storing images (image data) while superimposing (synthesizing) them in order to improve the signal-to-noise ratio. For example, one complete image is created by storing images obtained with 8 two-dimensional scans while superimposing them. In other words, a final image is created by synthesizing images that were formed by one or more X-Y scans. The number of images (frame integration count) for creating one complete image can be set arbitrarily to an optimum value in view of factors such as the efficiency of generating secondary electrons. Another final image can also be created as needed by further superimposing images that were created by integrating multiple image frames. When or after the desired number of images has been stored, blanking may be performed with the primary electron beam to stop the input of information to the image memory. When the frame integration count is set to 8, a sequence may be made so that the first image frame is deleted when the ninth image frame is input so that there are always 8 image frames. Weighted averaging can also be performed, for example, by adding the product of the image integrated in the image memory times (×) ⅞, to the ninth image when the ninth image is input. The control system 7 has an input device (not shown in drawing) that specifies the image acquisition conditions (scanning speed, number of images to be integrated), field-of-view correction method and how images should be output or saved. In addition, the control system 7 incorporates a memory medium (not shown in drawing) for storing various types of data. This embodiment of the present invention also provides a function that forms a line profile based on the detected secondary electrons or reflected electrons. Such a line profile should be formed based on the quantity of detected electrons when a specimen is scanned one-dimensionally or two-dimensionally with the primary electron beam, or based on the brightness information of the specimen image. A line profile obtained this way will be used, for example, to measure the dimensions of patterns formed on a semiconductor wafer. The schematic diagram in FIG. 1 was explained with the control system 7 as an in integral or semi-integral unit of the scanning electron microscope. However, the embodiment of the present invention is not limited by this example and may have an external control processor installed separately from the scanning electron microscope to perform the processing explained below. This case will require transfer media for sending the signals detected with the secondary signal detector to the control processor or sending the signals from the control processor to the electron lens and deflector of the scanning electron microscope, and also an input/output terminal for inputting and outputting the signals sent by way of the transfer media. Furthermore, this embodiment of the present invention also provides a function that for example, allows storing the observation conditions (measurement points, optical conditions for electron scanning microscope, etc.) in advance as a guide to help assist in viewing multiple points on a semiconductor wafer. Measurement and observation can easily be performed according to the contents of this guide. A program intended to perform the processing explained below may be stored in the memory medium and executed with a control processor having an image memory and supplying signals necessary for the scanning electron microscope. In other words, the following embodiment of the present invention may be established as a program invention used for charged-particle beam equipment such as scanning electron microscopes with an image processor. A specimen is placed on stage 8 of this embodiment for evaluating image resolution as explained below. Particles of gold having a large atomic number are typically deposited by evaporation on a carbon substrate and used as a specimen for evaluating image resolution. FIGS. 2A and 2B show SEM images of typical gold particle specimens. These are digital images consisting of 512×512 pixels. Image resolution is calculated by the evaluation algorithm explained below. The concept for evaluating image resolution is shown in FIG. 8 along with the gap resolution evaluation method, and the flow chart for calculating image resolution is shown in FIG. 9. A detailed description is given in the following embodiments. In the present embodiment, a microscope image is treated as an image of a three-dimensional object viewed from one direction, and the gradient and curvature of the object surface at each pixel point P are first obtained. Next, the hypothetical shortest distance required for the object to be distinguished from the surrounding area, or in other words, the local resolution is calculated from the gradient and the minimum contrast needed to recognize the object. The weighted harmonic mean of the local resolution is then calculated over the entire image, and the mean value is viewed as a typical resolution of the image. A specific example for calculating image resolution is explained below using the flow chart shown in FIG. 9. A microscope image is treated as an image of a three-dimensional object viewed in one direction and the geometrical features such as the gradient and radius of curvature of the object shape are calculated by range image analysis. The image is a digital image consisting of n×n pixels. The position of a pixel at an arbitrary point on the image is considered as (x, y), and the density at the pixel as “z”. To calculate the gradient and radius of curvature of the object shape at point P(x, y), a local region (partial region) comprised of m×m pixels (normally, m=5) with the center at point P is clipped out and the shape of the local surface is approximated by multi-degree equation z(x, y). Here, a quadratic equation is used as a multi-degree equation for approximating with a curved surface, as expressed in the quadratic equation (Eq. 1) below.z(x,y)=ax2+by2+cxy+dx+ey+f (Eq. 1) To obtain a fitting or matching with the curved surface, the coefficients “a” to “f” are determined so that the sum-square value of the fitting error is minimal. Once these coefficients are determined, the gradient g(x, y; θ) [θ is azimuth angle of gradient] of the local surface can be found by the following geometrical calculation (Eq. 2).g(x,y;θ)={(δz/δx)2+(δz/δy)2}1/2 (Eq. 2a)θ=arctan {(δz/δy)/(δz/δx)} (Eq. 2b) The radius of curvature, Rc, can also be similarly computed. Applying this approximated curved surface is also effective in eliminating the noise of the image. More specifically, by clipping out a part of an image containing noise and approximating it with a curved surface or plane, the noise can be partially eliminated, allowing calculation of the gradient with fewer adverse effects from noise. In view of the noise elimination effect and reduction in calculation time, a local region of a 5×5 pixel size (m=5) is preferable and a quadratic equation is suitable for this size. When a local region of a 3×3 pixel size (m=3) is used, the calculation time becomes shorter, but the effect achieved from eliminating the frequency component noise, especially noise exceeding a period of 3 pixels on the image pattern will be poor compared to cases using the quadratic equation of m=5. On the other hand, the 3×3 pixel size was used as the local region size for images having less noise and containing a large quantity of frequency components having a period of less than 5 pixels on the image pattern because the gradient calculation accuracy is better. When applying a plane (display functional equation is linear) as the approximate surface instead of the curved surface, the calculation time becomes shorter but the noise elimination effect is poor compared to cases using the curved surface for the quadratic equation display. Here, an azimuth angle representing the range from a specific azimuth angle θi−π/n to θi+π/n is expressed as θi (i=1, 2, . . . , n; n=36), and the resolution obtained by finding a harmonic mean in the image while weighting with the weight Wp of the local image resolution Rp having the azimuth angle, is viewed as the azimuth image resolution R(θi). The image resolution R over the entire image is obtained by finding the geometric mean of this azimuth image resolution R(θi) by using the whole azimuth angle. The image resolution calculated this way is not greatly affected even, for example, if unexpected noise intrudes into a portion of the microscope image.1/R(θi)=[∫{Wp(x,y)/Rp(x,y;θi)}dxdy]/[∫Wp(x,y)dxdy] (Eq. 3a)Rp(x,y;θi)=2ΔC/|g(x,y;θi)| (Eq. 3b)R={R(θ1), R(θ2), . . . , R(θn)}1/n (Eq. 4) Here, ΔC is the threshold contrast needed to discern a gap on the object shape that corresponds to the resolution, and is proportional to the maximum amplitude of the expected density value E{z} calculated from Eq. 1. When the proportional coefficient is considered as Kc, ΔC is given as follows (normally, Kc=0.1).ΔC=Kc·[E{z(max)}−E{z(min)}] (Eq. 5) The local weight Wp can be considered as the gradient g only (Eq. 6a) or the product (Eq. 6b) of g times [E{z(x, y)}−E{z(min)}]. The latter is used when evaluating an image in which the gradient, g, is large and more attention should be paid to bright portions. This greatly relieves the effects of brightness on the resolution when brightness is changed.Wp(x,y)=|g(x,y)| (Eq. 6a)[E{z(x,y)}−E{z(min)}]·|g(x,y)|] (Eq. 6b) Here, one way to calculate the local resolution with high accuracy when the applied surface is a curved surface is by calculating the Wp (Eq. 6) limited to cases where assuming that ½ of that value (=Rp/2) is less than the absolute value of the minimum radius of the curvature, and setting Wp=0 in all other cases. The standard deviation σ of density (corresponding to image noise) which is a parameter for evaluating image quality can also be calculated by using the local noise σp.σ={1/(n−2)}{Σ(σp2)}1/2(m=5) (Eq. 7a)σp={1/m)}{Σ[z(x,y)−E{z(x,y)}]2}1/2 (Eq. 7b) The embodiment of the present resolution evaluation method is explained using FIGS. 2A and 2B. Shown in FIG. 2A and FIG. 2B are images (512×512 pixels) of a specimen prepared for resolution evaluation, photographed with an SEM under different optical conditions. Image resolution Rcg by the present evaluation method and image resolution Rgap by the gap method of the related art are shown by the resolution ratio between the conditions A and B, along with evaluation errors caused by multiple human evaluators. In either resolution, the R(b)/R(a) ratio is larger than 1, but the gap method of the related art shows an error due to multiple evaluators, which is as large as ±50%. However, the error occurring in the present evaluation method is 0 as long as the same image data is used. The resolution evaluation algorithm stated above determines all parameters required for calculation by using only the information possessed by the image. The present evaluation method uses an algorithm that is not vulnerable to subjective impressions of human evaluators and has the following advantages: (1) Subjectivity of the evaluator accompanying the conventional gap method of resolution evaluation can be eliminated, so images can be objectively evaluated. (2) Calculation is performed on real space so that calculation parameters are easy to understand as a physical quantity. (Calculation is performed on frequency space by the FFT method.) (3) Not only resolution but also the standard deviation of density (image noise) can be calculated and evaluated. (4) Signal-to-noise ratio, image quality, viewing magnification accuracy (or error) can be calculated and evaluated. (5) Not susceptible to changes in brightness and image noise. (6) Applicable to specimens with a high directional pattern (Not possible with the FFT method). Not susceptible to effects occurring from contamination on the specimen. (7) Resolution can be evaluated among multiple microscopes based on the same specimen (image), allowing the instrumental error (differential between microscopes) to be easily found. In particular, advantage (1) allows objective comparison of image resolution between microscopes of the same type and is very helpful to the users when selecting the microscope model. The present embodiment was explained as an example for obtaining image resolution over the entire specimen image, but preferred embodiments are not limited to it. Resolution in a specific region of the specimen image (smaller than the entire image but larger than a local region used for calculating the local resolution Rp) may also be calculated. For instance, an image region selecting device (pointing device, etc.) not shown in the drawing may be connected to the control system 7 to select any desired size at any position on the image, so that image resolution in a region corresponding to the selected image is computed. This method eliminates background information not directly relating to the measurement object and acquires actual image resolution in the region to be measured. In addition, the throughput for calculating image resolution can be improved since the calculation is performed only on a specific region. As the second embodiment, a scanning electron microscope (SEM) utilizing the present image evaluation method is explained below. The scanning electron microscope is just one example, so the present image evaluation method can be applied to most inspection devices in which images must be accurately evaluated. On a semiconductor device production line, as shown in FIG. 7, multiple SEMs 701 to 704 are connected by a network to the master computer 705 for measuring and controlling the length of semiconductor device patterns. Each SEM has an image resolution computing function based on the above image resolution evaluation method, which is installed in the computer of the SEM control system. Self-evaluation of image resolution can be implemented by an instruction from the equipment operator. The resolution evaluation value appears on the monitor that also displays the microscope image. On the SEMs that have been used to measure and control the length of device patterns for extended periods of time, image resolution of each SEM is periodically evaluated by using the specimen for image resolution evaluation, displayed and recorded, along with the information on changes in the evaluation value. These periodically evaluated resolution values are stored in the master computer 705 where the data is collectively managed together with the information from other SEMs. If an image resolution evaluation value falls outside the allowable range or value, the operator is informed of the error on that microscope and master computer. The master computer 705 has an image display monitor and an image processor as explained earlier. The display monitor shows that the image resolution evaluation value is outside the allowable range or value. As specific display formats, information on changes in evaluation values on multiple SEMs may be graphically represented while distinguishing the allowable range from the area outside it. In another possible format, illustrations of multiple SEM models may be displayed on the monitor as shown in FIG. 7 and the illustration of the model on the display may start flashing if its evaluation value falls outside the allowable range or value. Using these display formats prevents the measurement accuracy from deteriorating even if a measurement error occurs due to a differential occurring between microscopes (instrumental error). In this embodiment, if the output image from an inspection device indicates a resolution evaluation value outside the allowable range, the error is displayed or stored in memory so that the instrumental error between each inspection device can be controlled while maintaining high resolution. On the microscope where the error occurred, a command for adjusting the optical system which is one inspection parameter is issued by an instruction from the operator or from the program installed in the control system. The lens system is then controlled by the signal based on the evaluation value through the control system of the microscope, so that the image resolution is set within the specified allowable range. Information on the adjustment process is stored in the SEM that caused the error and also sent to the master computer. In this example, each SEM evaluated its own image resolution; however the master computer for system control may evaluate the resolution of images sent from each SEM and send the evaluation value signal back to each SEM. Based on this evaluation value signal, each SEM controls the lens system through the control system of the microscope in the same way as described above in order to set the image resolution within the specified allowable range. In this way, differences in resolution occurring among multiple SEMs can be simultaneously measured and controlled. If an error is found, the above method is used to make an adjustment. This reduces differences in device quality between each production process and maintains uniform quality. Uniform quality in device production can also be maintained in the same way among factories in different locations by communicating this information. An embodiment is next described in detail using the previously described image resolution evaluation specimen. In this embodiment, a dimensional calibration specimen (JP-A-31363/1996) was used as an image resolution evaluation specimen having an absolute dimensional pattern. FIG. 4 shows an SEM image of this calibration specimen. The pitch dimension of the grid pattern formed on a silicon substrate is an extremely accurate 0.240±0.001 μm. Using this specimen allows highly accurate magnification calibration of SEM images. This means that pixels used as size units for microscope digital images can now be converted into or converted from actual dimensional units (for example, nanometers) with high accuracy, allowing the resolution evaluation value R in the present invention to be displayed in length units (for example, nanometers). However, because the pattern of the specimen used here consists of vertical lines, the image resolution is not calculated from the average in all azimuth directions, but instead uses the image resolution in an azimuth direction (horizontal direction in the case of FIG. 4) showing a minimum value among azimuth image resolution values calculated from Eq. 3a. The resolution evaluation value R is convenient during actual use because it is displayed in units of actual dimensions (length) that do not depend on the viewing magnification. In pixel unit display on the other hand, a specified number of pixels is allocated to the image resolution value, making it convenient to directly calculate the optimum viewing magnification. In an image resolution display, length units are usually used, but the pixel unit or both units can also be simultaneously displayed at the discretion of the equipment operator. By presetting the number of pixels to be allocated for image resolution, the viewing magnification can be automatically set for various specimens. Image resolution and signal-to-noise ratio (SNR) are parameters used for evaluating image quality of a microscope images. When the average of the expected density value E{z} is Zav, the signal S can be approximated as S=Zav−E{z(min)}, and the noise N can be approximated as N=σ. (See Eq. 7.), the SNR can then be obtained as follows.SNR=[Zav−E{z(min)}]/σ (Eq. 8) If the fine structure pattern is the focus of the image quality parameter, and C is utilized as the maximum amount of information (bit units) per pixel in that structure pattern, then C can be calculated from the following equation using the image resolution value R of pixel units.C={ log2(1+SNR)}/R2 (Eq. 9) The image can therefore now be evaluated in terms of image resolution, noise, and image quality. By setting a threshold level for at least one of these evaluation parameters to determine whether the image is a pass (acceptable) or a fail (reject), the image that was obtained can be judged as acceptable or a reject based on these pass/fail parameters as well as each evaluation value. The image display device displays different messages according to the evaluation value and pass/fail results (for example, whether the threshold level is exceeded, to what extent the threshold level was exceeded, etc.). The messages to be displayed can also be selected by the equipment operator. The messages may be stored in the control processor memory for readout later on. The viewing magnification accuracy (or error) is discussed next. The image resolution evaluation value (unit: μm) of a microscope image taken with a dimensional calibration specimen under a particular viewing magnification is considered as R, the absolute pitch dimension of the dimensional calibration specimen as L, and the error as Δx (units: μm). The dimensional calibration specimen has dimensions of L=0.240 μm and Δx=0.001 μm in this example. On the surface of this specimen, a repetitive pattern having a square-wave cross section is fabricated by silicon anisotropic etching. Each square-wave edge sharply rises perpendicular to the specimen surface. After correcting the image rotation so that the grid pattern image of the dimensional calibration specimen is vertical on the display monitor, the viewing magnification is calibrated while measuring the pitch dimension at multiple points (about 10 to 20 points) at a magnification of 5 to 100,000, so that the average is 0.240 μm. In other words, the scan area of the electron beam is changed so that the measured average value approaches 0.240 μm. The pitch dimensions of the pattern itself have an error of a few nanometers so the measurements made at these multiple points are averaged to reduce the error. At this point, the magnification error E and magnification accuracy P on the percentage (%) display are calculated as follows.E=100×{(Δx)2+R2}1/2/L (Eq. 10)P=100−E (Eq. 11) Verification using general scanning electron microscopes confirmed that respective viewing magnification errors E within ±2% and ±0.5% were obtained. These results prove that, like image resolution, the viewing magnification accuracy and error can be controlled with high accuracy versus microscope differentials (instrumental error) and changes over time not only for an individual microscope but also for multiple microscopes. Changes in data over elapsed time on these evaluation parameters can also be displayed as needed on the image display device in a graphic or spreadsheet format by an instruction from the equipment operator. FIG. 5 shows typical graphical changes in the image resolution over one month, displayed on the image display monitor of SEM equipment A and B. The figure clearly shows that image resolutions on both units of equipment are held within the range between 3.2 and 3.8. If the resolution deviates from this range, the equipment should be serviced. This graph also proves that due to an instrument error the resolution of equipment B is 0.1 better than equipment A, even though the image resolutions of both pieces of equipment are within the allowable range. Based on this instrument error, a coefficient for correcting the instrumental error between the two units of equipment is found and used to correct the measurement length value. As a result, semiconductor device production can now be controlled while the instrument error between the microscopes is minimized. FIGS. 6A and 6B show an example of evaluation of astigmatic images using azimuth resolution. FIG. 6A shows microscope images of gold particles with correction parameter s.p. values at 0, 40 and 80 hours on the astigmatic image control system installed in the microscope control system 7. Azimuth image resolutions at this point are plotted on a polar coordinate in FIG. 6B. The azimuth image resolution curve is nearly circular at s.p.=0, but increases at azimuths of 17 and 35 (direction of arrow) as the s.p. value increases to 40 and then 80, while exhibiting little change at azimuths of 8 and 26 (direction of arrow). These changes indicate the extent of image blur in FIG. 6A. This image blur corresponds to the beam spread in the microscope. (See the image at s.p.=80 in FIG. 6A.) A microscope capable of automatically performing focus adjustment and astigmatic correction with high accuracy can be achieved by installing in the computer in the control system 7, a program combining the azimuth image resolution evaluation method with the optimization method to minimize evaluation values in all azimuth directions. The above embodiment was explained using SEM, but similar effects were attained from a scanning transmission electron microscope (STEM) and scanning ion microscope (SIM). In the STEM, the specimens are thin films and the brightness signals used for image construction are secondary electrons, primary electrons transmitting through the specimen, or X-rays emitted by excitation of primary electrons. In the case of SIM, since spattering damage occurs on the specimen surface from ion beams which are not used by the SEM or STEM, the same point on the specimen cannot be observed at high magnification. Therefore, because the present evaluation method is “applicable to specimens with a high directional pattern” mentioned as the advantage (5) earlier, a wider range of specimens can now be used for resolution evaluation. The present evaluation method has proven very effective in improving the operability and the accuracy of image resolution during actual use. The above embodiments were explained using microscopes as the device of the embodiment. However, the image resolution evaluation method of the present invention can also be used in equipment utilizing microscope images for beam positioning and setting the beam irradiation area, for example of charged-particle beam systems such as focused ion beam machining systems, electron beam diagnostic equipment, electron beam exposure systems. Therefore, the “microscopes” mentioned here include “charged-particle beam systems” utilizing microscope images. The present invention uses a technique by which a microscope image is treated as an image of a three-dimensional object viewed from one direction and is approximated to a curved surface (or plane) described with a multi-order (or linear) function z=f(x, y) for each local region. This technique is also very effective as an image processing method for reducing image noise. The density distribution after image processing is definitely the E{z(x, y)} distribution of Eq. 7. FIGS. 3A and 3B show the image processing effects on SEM images of a resolution standard specimen. FIG. 3A is the original image and FIG. 3B is the processed image. The image of FIG. 3B appears smooth because of the noise elimination effect. The present invention can be applied not only to SEM and SIM but also to any microscope using the gap method and FFT method of the related art as the image resolution evaluation method, for example, optical microscopes, and scanning probe microscopes. Likewise, the image processing method of the present invention has the effect of eliminating noise on all microscope images as well as SEM and SIM. The microscope image resolution evaluation process of the present invention is further not susceptible to the subjective impressions of the evaluators, so image resolution evaluation values can be obtained with high reliability and repeatability. In the image resolution evaluation method using a magnification calibration specimen, optical characteristics are precisely adjusted by making use of the evaluation value signal so that the desired image resolution, magnification, image noise and image quality can be maintained and controlled with a high degree of accuracy. Furthermore, when the present invention is applied to multiple microscopes used for quality control in semiconductor device production, the differential (instrumental error) in image resolution between each microscope and changes in resolution over time can be accurately controlled, making it possible to minimize variations or irregularities in device quality during the device manufacturing process. |
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claims | 1. A system for accident tolerant oxide fuel, said system comprising:a fuel assembly comprising at least one disk placed axially between fuel pellets composed of UO2, wherein said at least one disk possesses a higher thermal conductivity material than that of said UO2 to provide enhanced heat rejection thereof; andwherein said fuel pellets comprise annular pellets having an annulus filled with a higher thermal conductivity material from that of said at least one disk,wherein each annulus among said annular pellets is filled with rod-shaped inserts. 2. The system of claim 1 wherein at least one of said higher thermal conductivity material comprises at least one of the following: Si, Zr, Al2O, Nb or Mo. 3. A system for accident tolerant oxide fuel, said system comprising:at least one disk placed axially between fuel pellets comprising UO2, wherein said at least one disk possesses a higher thermal conductivity material than that of said UO2 to provide enhanced heat rejection thereof;wherein said fuel pellets comprise annular pellets having an annulus filled with a higher thermal conductivity material from that of said at least one disk; anda cladding coating comprising zircaloy coated with a material that provides stability and high melting capability and self-healing properties at elevated temperatures,wherein each annulus among said annular pellets is filled with rod-shaped inserts. 4. The system of claim 3 wherein said material coated on said zircaloy comprises ZrSi2. 5. The system of claim 3 wherein said material coated on said zircaloy comprises at least one silicide. 6. The system of claim 3 wherein said material that provides stability and high melting capability and self-healing properties at elevated temperatures is ZrSi, and wherein said higher thermal conductivity material comprises at least one of the following: Si, Zr, Al2O3, Nb, or Mo. |
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claims | 1. An aspheric lens fabricating method comprising:designing a desired aspheric surface;producing a basic lens of an arbitrary shape;comparing the desired aspheric surface with the basic lens and thus setting a deviation region;performing ion beam milling for milling the deviation region by exposing the deviation region to an ion beam by rotating and inclining the basic lens and by controlling exposure time according to a processing depth of the deviation region; andprocessing a shape of the basic lens by comparison with the desired aspheric surface by the ion beam milling and making the basic lens consistent with the desired aspheric surface. 2. The method of claim 1, wherein the basic lens is an ideal spherical shape than can be massively produced easily. 3. The method of claim 1, wherein the desired aspheric surface has a parabolic shape. 4. The method of claim 1, wherein the desired aspheric surface has an elliptical shape. 5. The method of claim 1, wherein comparing the basic lens processed by ion beam milling with the desired aspheric surface is performed by a stylus profiling method. 6. The method of claim 1, wherein comparing the basic lens processed by ion beam milling with the desired aspheric surface is performed by an interferometer. 7. The method of claim 1, wherein in the step of performing an ion beam milling, a plurality of the basic lenses are mounted at a specimen mounting unit. 8. The method of claim 7, wherein the specimen mounting unit not only rotates the basic lenses mounted thereto but also inclines the basic lenses. 9. The method of claim 1, wherein the basic lens is formed of a mixture between hereto materials. |
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059848537 | claims | 1. The method of invasively delivering radiation to a therapy location in a living body, comprising: a) forming a miniaturized source of ionizing radiation on a support comprising at least one plate member, including a pair of electrodes in the form of an anode and a cathode on the support; b) inserting the source into a living body and directing the source at the therapy location; and c) activating the source of ionizing radiation by selectively energizing and deenergizing the source so as to direct a highly accelerated electron emission from the cathode to the anode in the vicinity of the therapy location. 2. The method of claim 1, wherein the high electron emission is created by applying a positive voltage on the anode to cause electrons to be emitted from the cathode. 3. The method of claim 2, wherein the voltage is in kV range. 4. The method of claim 1, wherein the high electron emission is created by heating the cathode to a high temperature, giving rise to thermal emission of electrons toward the anode. 5. The method of claim 1, including, retarding the electron emission by impact at the anode. 6. The method of claim 1, including forming the anode of a metal selected from a group consisting of tungsten, molybdenum, cobalt and aluminium. 7. The method of claim 1, including forming the anode of a metal having an atomic number higher than 50. 8. The method of claim 7, including depositing the anode metal in the form of a thin film on a first plate member. 9. The method of claim 8, including forming the cathode with a thin deposited film of a material having a low work function on a second plate member. 10. The method of claim 9, including forming a miniature platform formed by the plate members. 11. The method of claim 10, including connecting the anode and cathode to a voltage source by deposition of metallic strips. 12. The method of claim 1 wherein the support includes a first plate and a second plate. 13. The method of claim 12, including forming the first and second plates of a semiconductor material and doping leads therein, and connecting the anode and cathode to a voltage source by doping leads onto the semiconductor. 14. The method of claim 12, including forming a microcavity in one of the first and second plates, and forming at least the cathode in the microcavity and the anode on the other plate. 15. The method of claim 12, including forming the microcavity by etching off a sacrificial layer between the first and second plates. 16. The method of claim 12, including adhesively bonding adjoining surfaces of the first and second plates. 17. The method of claim 12, including forming at least one of the first and second plates of simple crystalline silicone. 18. The method of claim 1, including forming the anode at least partly of a metal having an atomic number greater than 20. 19. The method of claim 1, including forming a third electrode in the form of a gate within the microcavity, controlling the electronic current emitted toward the anode. 20. The method of claim 1, including isolating the electrodes to prevent short circuits and current leakage. 21. The method of claim 1, including forming a sharp tip in the microcavity and forming the cathode on the tip. 22. The method of claim 1, including determining a desired radiation level and duration of radiation, and energizing the source to achieve the desired level and duration. 23. The method of claim 22, including deenergizing the source and withdrawing the source from the location. 24. The method of claim 22, including repeating the steps of inserting the source, deenergizing the source and withdrawing the source at selected time intervals. 25. The method of claim 1, including encompassing the source within a cannula element comprising material transparent to emitted radiation, penetrating tissue of the living body and directing the source in the cannula to the therapy location. 26. The method of claim 25, including connecting electrical leads to the anode and the cathode, connecting insulated wires to the leads and extending the wires through the cannula to a power source. 27. The method of claim 1, including mounting the source near a distal end of a wire. 28. The method of claim 27, including inserting a catheter into body tissue, and directing the source and wire through the catheter to the therapy location. |
description | This invention was made with Government support under contract: W911NF-12-1-0605, awarded by the U.S. Army. The Government has certain rights in this invention. The present disclosure relates to methods, devices, and systems for positional control of ions in an ion trap. An ion trap can use a combination of electrical and magnetic fields to capture one or more ions in a potential well. Ions can be trapped for a number of purposes, which may include mass spectrometry, research, and/or controlling quantum states, for example. Ions can be transported along a path in some regions of an ion trap, and can have their motion restricted in other regions of an ion trap. As an example, electric and/or magnetic fields can be used to transport and/or capture ions (e.g., charged particles). Some ion traps make use of electrodes to transport and/or capture ions, for example, by providing static and/or oscillating electric fields that can interact with the ion. It may be desirable to provide differing degrees of positional control to such ions as they move through different regions of an ion trap; however, providing differing degrees of positional control over ions in an ion trap can be problematic using conventional methods, which can employ electrodes of uniform pitch to provide positional control. The embodiments of the present disclosure relate to methods, apparatuses, and systems for design, fabrication, and use of an ion trap with variable pitch electrodes. As described herein, different issues which can arise from the use of some previous approaches to ion trap technology can be overcome. One such issue can arise from use of electrodes that are formed on uniform pitch in an ion trap. Forming electrodes on uniform pitch in an ion trap can limit positional control over ions in an ion trap, for example, by providing a uniform electric field that can interact with the ion. Stated differently, positional control of ions in an ion trap can be limited to a single degree of positional control over the ions if the ions are transported and/or positioned using electrodes that are formed on uniform pitch. In the following detailed description, reference is made to the accompanying figures that form a part hereof. The figures show by way of illustration how one or more embodiments of the disclosure may be practiced. The embodiments are described in sufficient detail to enable those of ordinary skill in the art to practice one or more embodiments of this disclosure. It is to be understood that other embodiments may be utilized and that process, electrical, and/or structural changes may be made without departing from the scope of the present disclosure. As will be appreciated, elements shown in the various embodiments herein can be added, exchanged, combined, and/or eliminated so as to provide a number of additional embodiments of the present disclosure. The proportion and the relative scale of the elements provided in the figures are intended to illustrate the embodiments of the present disclosure, and should not be taken in a limiting sense. It should be noted that although many of the figures provided herein provide visual views of example optical bench configurations and example alignments of optical fibers, the embodiments of the present disclosure can be accomplished by using different configurations, materials, and/or components. Further, as used herein, “a” or “a number of” something can refer to one or more such things. For example, “a number of optical components” can refer to one or more optical components. FIG. 1 provides an illustration of an example ion trap 100 according to the present disclosure. As illustrated in FIG. 1, the ion trap 100 can include a plurality of vias 109-1, 109-2, . . . , 109-N (referred to generally herein as “vias 109”). A plurality of capacitors 110-1, 110-2, . . . , 110-N (referred to generally herein as “capacitors 110”) can be included and positioned such that a respective capacitor 110-1, for example radially encompasses a respective via 109-1, for example. The ion trap 100 can be fabricated using anisotropic and deep reactive ion (DRIE) etching techniques, among other suitable techniques. The plurality of capacitors 110 can be formed on a first pitch 120-1. As used herein, “pitch” refers to a distance between various similar objects. For example, as illustrated in FIG. 1, a first capacitor (e.g., 110-1) can be formed adjacent to a second capacitor (e.g., 110-2), and the distance (e.g., first pitch 120-1) between the two capacitors in the x-direction is then the pitch on which the two capacitors 110-2, 110-2 are formed. As a further example, a pitch (e.g., 122-1) associated with an electrode (e.g., 112-1) can be a distance between the rails of the electrode. In the example of FIG. 1, the ion trap 100 can include a first region 114 and a second region 116. In some embodiments, first region 114 can include a plurality of vias 109 and a plurality of capacitors 110. The second region 116 can include a plurality of electrodes 112-1, 112-1, . . . , 112-N (referred to generally herein as “electrodes 112”), and a control region 118. In some embodiments, respective electrodes among the plurality of electrodes 112 can be formed on a pitch that is different from the first pitch 120. For example, electrode 112-2 can be formed on a second pitch 122-1 that is different from the first pitch 120-1. As a further example, electrodes 120-N can be formed on a pitch 122-N that is different than the first pitch 120-1. Examples are not so limited; however, and respective electrodes of the plurality of electrodes 112 can be formed at a pitch that is different both from the first pitch 120-1 and a pitch (e.g., 122-1) on which a different respective electrode is formed. That is, electrode 112-N can be formed on a pitch 122-N that is different than the first pitch 120-1 and different from pitch 122-1, for example. In some embodiments, the pitch of respective electrodes of the plurality of electrodes 112 can vary along a length of a respective electrode (e.g., 112-1). For example, in the first region 114, an electrode 112-1 can have a pitch that is the same as the first pitch 120-1, and a pitch that is different than the first pitch 120-1 in the second region 116. In some embodiments, the rails of a respective electrode 112 can taper continuously from the first pitch to the second pitch. In some embodiments, an apparatus can include an ion trap 100 and a plurality of variable pitch electrodes 112 disposed on the ion trap 100. A respective electrode (e.g., 112-1) of the plurality of electrodes 112 can have a first pitch 121-1 in a first region 114 of the ion trap 100 and a second pitch 122-1 in a second region 116 of the ion trap 100. A plurality of capacitors 110 can be disposed in the first region 114. In some embodiments, a respective capacitor (e.g., 110-1) of the plurality of capacitors 110 can be formed on the first pitch 120-1. The capacitors 110 can be trench capacitors, for example. In some embodiments, the first pitch can be between 50 microns and 70 microns, and the second pitch can be less than 50 microns. Embodiments are not so limited; however, and the second pitch can be greater than 70 microns, for example. As discussed in further detail in connection with FIG. 2, providing electrodes 112 on a different pitch (e.g., 121-1, . . . , 121-N, 122-1, . . . , 122-N) than a pitch 120-1 associated with the capacitors 110 can allow for ions to be transported with varying degrees of positional control in the ion trap 100. For example, coarse positional control over ions in the ion trap 100 can be provided in a first region 114, while fine positional control over ions in the ion trap 100 can be provided in a second region 116. FIG. 2 illustrates a portion of an example ion trap 200 according to the present disclosure. In some embodiments, a pitch on which a respective electrode (e.g., 212-1) is formed can vary along a length of the respective electrode (e.g., 212-1). That is, the pitch of a respective electrode (e.g., 212-1) can be tapered such that a pitch at one end of the electrode (e.g., 212-1) is different than a pitch at the opposite end of the electrode (e.g., 212-1). For example, with respect to electrode 212-1, pitch 221-1 can be different than pitch 220-1, and can also be different than pitch 222-1. In some embodiments, the capacitors 210 can be trench capacitors. As an example, trench capacitors 210 can be formed such that a trench region of at least one of the plurality of capacitors 210 extends to a depth of between 200 and 400 microns from the surface of the ion trap. In some embodiments, at least one of the plurality of capacitors 210 can have a capacitance between 50 and 250 picofarads. For example, at least one of the capacitors 210 can have a capacitance of 100 picofarads. In some embodiments, an ion trap apparatus can include an apparatus body, a plurality of vias 209 disposed on the body, and a plurality of electrodes 212. Each respective electrode (e.g., 212-1) can be electrically coupled to a respective capacitor (e.g., 210-2). A first pitch 220-1 of each respective electrode 212 can be the same as a pitch 220-1 of the respective capacitor (e.g., 210-2) in a first region 214 of the body, and a second pitch (e.g., 222-1) of each respective electrode 212 can be different than the pitch 220 of the respective capacitor 210 in a second region 216 of the body. Advantageously, this can allow for variable positional control of an ion in the different regions. For example, coarse positional control can be provided in first region 214, and fine positional control can be provided in second region 216 and in the control region 218. In some embodiments, the pitch of a respective electrode (e.g., 212-1) can be tapered from the first pitch 220-1 to the second pitch 222-1 such that a distance between the rails of the respective electrode (e.g., 212-1) changes as a distance from the respective capacitor (e.g., 210-2) changes. An example method 330 of fabrication for one or more embodiments contained herein is presented below. In some embodiments, an ion trap can be formed from a plurality of alternating metal and dielectric layers that can be formed together in a sequential order. For instance, anisotropic etching or deep reactive ion etching (DRIE) can be used to form portions of the ion trap. Anisotropic etching and DRIE are different etching techniques in the context of device fabrication. FIG. 3 illustrates an example flow chart of an example method 330 for forming an ion trap with variable pitch electrodes. In this embodiment, the process can include forming a plurality of vias through an ion trap apparatus, at block 332. For example, in the embodiment of FIG. 2, the ion trap includes a plurality of vias 209 that can be formed through the substrate. At block 334, the method 330 includes forming a plurality of capacitors in the ion trap apparatus such that a respective via (e.g., 209) is substantially encircled by a respective capacitor (e.g., 210-1) of the plurality of capacitors 210. In some embodiments at least one of the capacitors can be a trench capacitor. In various embodiments, the method 330 can include forming a plurality of electrodes, wherein a respective electrode is electrically coupled to the respective capacitor of the plurality of capacitors, and wherein the respective electrode is formed at a first pitch in a first region of the ion trap apparatus and is formed at a second pitch in a second region of the ion trap apparatus. In some embodiments, a pitch associated with a respective electrode can taper from the first pitch to the second pitch such that a distance between the rails of the electrodes changes as a distance from a respective capacitor changes. The method 330 can also include forming at least one of the plurality of capacitors to a depth between 250 and 350 microns below a surface of the ion trap apparatus. In some embodiments, the method can include filling a trench region of at least one of the plurality of capacitors with a doped polysilicon material. For example, the sidewalls of at least one of the plurality of capacitors can be oxidized and subsequently filled with a polysilicon. In some embodiments, the polysilicon can be a boron-doped polysilicon, for example 1.0×1025 m−3 boron-doped polysilicon. In some embodiments, the method 330 can include forming the plurality of electrodes out of a metal, e.g., gold or other suitable metal. The electrodes can be formed such that a width of a respective rail of an electrode is between 1 micron and 2 microns. The method 330 can include controlling a position of an ion in the ion trap with a first level of positional control in the first region of the trap, and controlling the position of an ion in the ion trap with a second level of positional control in the second region of the trap. In some embodiments, the first level of positional control and the second level of positional control can be different. For example, a comparatively coarse level of positional control over the ion can be provided in the first region of the trap and a comparatively fine level of positional control over the ion can be provided in the second region of the trap. Although specific embodiments have been illustrated and described herein, those of ordinary skill in the art will appreciate that any arrangement calculated to achieve the same techniques can be substituted for the specific embodiments shown. This disclosure is intended to cover any and all adaptations or variations of various embodiments of the disclosure. It is to be understood that the above description has been made in an illustrative fashion, and not a restrictive one. Combination of the above embodiments, and other embodiments not specifically described herein will be apparent to those of skill in the art upon reviewing the above description. The scope of the various embodiments of the disclosure includes any other applications in which the above structures and methods are used. In the foregoing Detailed Description, various features are grouped together in example embodiments illustrated in the figures for the purpose of streamlining the disclosure. Rather, inventive subject matter lies in less than all features of a single disclosed embodiment. |
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description | A practical method for determining the operating limit minimum critical power ratio (OLMCPR) of a Boiling Water Reactor (BWR) nuclear core is disclosed here. This practical improvement results in the realization of greater operating margins for the core which in turn results in more efficient and cost effective core operation and/or configurations. This is a more direct approach towards demonstrating compliance of a nuclear reactor with USNRC licensing requirements than processes conventionally used for such purposes. A data processing system is disclosed including a computer having memory and various I/O or display devices that is specifically programmed for providing simulation of transient operational events in a BWR and for a subsequent compilation and display of one or more response histogram(s) that incorporate all the inherent xe2x80x9cuncertaintiesxe2x80x9d associated with reactor plant initial state conditions and other parameter(s) of interest or importance. A method is used to calculate a generic bias in change in critical power ratio during a transient event (xcex94CPR/ICPR) and uses the resulting Probability Distribution Function (PDF) to predict a more accurate OLMCPR without first calculating a SLMCPR. From a large number of experimental trials that take many factors into account, a PDF for the transient xcex94CPR/ICPR is created and the standard deviation in xcex94CPR/ICPR is determined for each transient event. A nominal xcex94CPR/ICPR for the transient event starting from nominal initial conditions is also determined. Histograms of individual rod CPR values for the minimum point in the transient are created by drawing random values of initial CPR and transient xcex94CPR/ICPR uncertainty. The initial critical power ratios (ICPR) are converted, or translated, to MCPRs by a common random value of xcex94CPR/ICPR. From the MCPR values, the percentage of NRSBT is calculated for each trial. If the percentage of NRSBT is greater than 0.1%, initial operating conditions are changed and the process is repeated until the NRSBT is equal to 0.1%. The NRSBT distribution histogram is analyzed using statistical methods to determine the xe2x80x9ccentral tendencyxe2x80x9d of the distribution. Typically the mean or median is used as a statistic to quantify central tendency. The value of this statistic is defined here as the nominal value. In the discussions that follow, examples are given where the mean value is chosen as the nominal value although the present invention is not limited to this choice. Use of the median value or the value of some other statistic for central tendency as the nominal value is also contemplated as part of the present invention. The uncertainty in the nominal value of the statistic that is used to quantify central tendency is expressed in terms of a xe2x80x9cconfidence intervalxe2x80x9d for the nominal value. A confidence interval is defined such that there is a specified probability (usually of 50% or greater) that the interval contains the nominal value. For example, a 95% probability that the interval bounds the mean, defines a 95% confidence interval for the mean. The specified probability used to establish this confidence interval is called the xe2x80x9clevel of confidencexe2x80x9d or confidence level. The susceptibility to boiling transition during the transient is quantified statistically as either (1) the probability that a single rod in the core is susceptible to boiling transition or (2) the expected fraction of total rods in the core susceptible to boiling transition. Such a statistical relationship is possible because each individual trial value of NRSBT has been determined by summing the probabilities that individual fuel rods have CPR values less than 1.0 during the transient. The nominal value for each NRSBT distribution can also by the present invention be associated with the distribution of initial rod CPR values for all fuel rods in the core. It is by this process that a relationship can be established between the minimal initial MCPR value for all fuel rods in the core and the probability and confidence level that the fuel rods will be susceptible to boiling transition during the transient. The minimal initial MCPR value for the core when determined in this way using the probability and confidence level established by the USNRC design basis requirement for the number of rods not susceptible to boiling transition during the AOO transient, is by definition, the minimum Operating Limit MCPR required to demonstrate compliance. In accordance with one aspect, the present invention is a system including a data processing apparatus programmed to execute specific routines for simulating BWR core operating conditions and for calculating and statistically demonstrating the OLMCPR of a reactor in accordance with the improved method of the present invention as described in detail below. FIG. 8 shows a block diagram of an example data processing system, contemplated for performing the multi-dimensional simulation of reactor core transient response and for the direct evaluation of OLMCPR for a BWR reactor core in accordance with the present invention. Essentially, the system includes a central processing unit 801 (CPU), a storage memory 802, and a user interfacing I/O devices 803 and, optionally, one or more displays 804. Storage memory 802 includes a data base (not shown) of reactor plant state information, parameter values and routines for implementing multi-dimensional simulations of core operating conditions and evaluating OLMCPR in accordance with the improved method of the present invention as described herein below. A statistical study will be performed for each type of AOO, for each class of BWR plant type, and for each fuel type to determine the generic transient bias and uncertainty in the xcex94CPR/ICPR. Enough trials (on the order of one hundred) are made starting with the nominal conditions, using random variations in the model and plant parameters. Uncertainties in initial conditions that contribute to the xcex94CPR/ICPR (e.g., core power) are also included in the perturbations. The data are utilized to determine bias and standard deviation on the transient xcex94CPR/ICPR. A flow chart for the process of the present invention is shown in FIG. 9. Block 909 remains unvaried throughout the calculation of the OLMCPR, and the xcex94CPR/ICPR for individual transient events for each reactor type and fuel type must be determined before the process is used. FIG. 10 shows the resulting graph of xcex94CPR/ICPR for one specific type of AOO. Histogram 1000 shows the number of trials 1002 with a resulting CPR 1001 for each rod versus the corresponding CPR 1001 values. The PDF 1003 represents the distribution of CPR before the transient event. Each CPR value then changes according to individual xcex94CPR/ICPR 1006 values. The aggregate of the transient CPR values yields the PDF 1004 during the transient event. The nominal xcex94CPR/ICPR 1005 is defined to be the difference in the nominal CPR value of the PDF 1003 and the nominal CPR value of the PDF 1004. The calculation of the OLMCPR is as follows. Step 1: Assume a set of base core operating conditions using the parameters to run the plant generates a core MCPR equal to the OLMCPR as shown by block 901. Step 2: Using the parameters, such as core power, core flow, core pressure, bundle power and others, that predict a general bundle CPR set forth in block 907, determine the ICPR for each bundle in the core, as shown by block 902. Step 3: Using parameters, such as rod placement within each bundle and rod power distribution, that change each bundle CPR into individual rod CPR values set forth in block 908, determine the ICPR for each rod in the core, as shown by block 903. Step 4: Using a randomly drawn individual xcex94CPR/ICPR 1006 value from the graph of the appropriate transient represented in FIG. 10, MCPR values are projected for corresponding values of ICPR according to Equation 4. In FIG. 11, this process is represented by Shift 1109. Histogram 1100 shows the number of rods at a specific CPR value 1102 versus the corresponding CPR value 1101. The histogram 1107 is translated to histogram 1108 during the transient using a randomly selected xcex94CPR/ICPR 1006 value. Lowest CPR value 1105 becomes lowest CPR value 1106, and lowest CPR rod PDF 1103 becomes lowest CPR rod 1104. MCPR i = ICPR i ( 1 - ( Δ xe2x80x83 C xe2x80x83 P xe2x80x83 R I xe2x80x83 C xe2x80x83 P xe2x80x83 R ) 1 ) ( Equation xe2x80x83 4 ) Step 5: Using the ECPR probability distribution shown as PDF 1104 and set forth in block 910, determine the percentage of NRSBT in the core by summing the probabilities of each rod in the core that is subject to boiling transition as shown by block 905. This summation is performed using Equation 3, shown above. Step 6: Vary the parameters set forth in blocks 907 and 908 for a set number of Monte Carlo statistical trials as shown by block 906. Compile the statistics from all the trials from steps 2 through 5 to generate a probability distribution of NRSBT. Step 7: Compare the value of percentage of NRSBT to 0.1% as shown in block 911. If the percentage is greater than 0.1%, reset the core parameters to different initial conditions in order to comply with the USNRC regulations as shown in block 912. Similar to Step 1 and block 901, the new initial conditions are assumed to generate the OLMCPR. The determination of NRSBT restarts and runs until the value of NRSBT is equal to 0.1%. Similarly, if the percentage is less than 0.1%, the core parameters are reset to increase the value of NRSBT in order to operate the core more efficiently or to reduce effluents. Step 8: If the percentage of NRSBT equals 0.1%, the assumed value of OLMCPR, which equals core MCPR, complies with the USNRC regulations as shown by block 913. Accordingly, the operating core conditions are set as the assumed parameters. Two assumptions are made for the above estimation of OLMCPR. First, in performing step 4, shown in FIG. 11 as shift 1109 and in FIG. 9 as block 904, the inventors assume that random draws from the xcex94CPR/ICPR distribution are permissible for a perturbation in the initial conditions. Therefore, variations in xcex94CPR/ICPR must be independent of perturbations in initial conditions or have a negative correlation, so that the interaction tends to diminish the individual effects. Second, in performing step 4, the inventors assume that the transient change in the xcex94CPR/ICPR applies to all rods. A demonstration analysis shows that the xcex94CPR/ICPR is not sensitive to the uncertainty in core power, core flow, core pressure, feedwater temperature, and rod peaking factor (R-factor). Of these, one of the most important parameters in the currently approved process is core power. This parameter actually results in an effect opposite the effect on ICPR. If the power increases, the ICPR will decrease but the xcex94CPR/ICPR will also decrease. This will result in an MCPR that would be higher than derived through the currently approved process. Another conservative factor is the intended use of the nominal xcex94CPR/ICPR. If the core was adjusting to a limiting rod pattern to maximize the number of contributing bundles, as is done for the currently approved process, the xcex94CPR/ICPR is 4% lower. Table 1 shows the impact of uncertainty in critical ICPR values on xcex94CPR/ICPR values. Column 101 lists the critical parameter quantities that affect the xcex94CPR/ICPR. Column 102 lists the percentage uncertainty of each parameter corresponding to the standard deviation of the associated PDF. "sgr" is the standard deviation of the PDF corresponding to the uncertainty in parameter quantity. Column 103 lists the change in the xcex94CPR/ICPR corresponding to a change of one standard deviation of each parameter. The xcex94CPR/ICPR is not sensitive to the other unknown parameter in the currently approved process. The axial power distribution is also part of the local power distribution (TIP uncertainty) calculation in the currently approved process. For a very large change in axial power shape (nearly two times higher power in the bottom of the bundle), the sensitivity to xcex94CPR/ICPR is less than 2%, which is insignificant. The other assumption to be validated is that a constant value of xcex94CPR/ICPR can be applied to rods at different ICPR values. As described above, the transient MCPR distribution will be obtained by transforming the ICPR distribution using Equation 4. To further validate this assumption, a specific set of calculations were performed. Benchmark calculations were made for a transient event that included the uncertainties in core power and channel pressure drop as initial conditions, as well as uncertainties in the model. Core power and channel pressure drop uncertainties were chosen, because they are the only currently approved process compatible uncertainties that are also varied in generating the generic uncertainty probability distribution function. MCPR distributions during the transient were generated for two fuel bundles in the core through ninety-eight transient calculations. The two bundles are very close in ICPR values and have identical xcex94CPR/ICPR values. To verify the translation process, ninety-eight Monte Carlo calculations were then performed where only the core power and pressure drop were varied to generate a PDF of ICPRs at the initial operating state. FIG. 12 shows histogram 1200, which is the number of rods 1202 at a certain CPR versus the corresponding CPR 1201 value. PDF 1203 is the ICPR distribution that was created using the Monte Carlo calculations varying core power and pressure drop. PDF 1205 is the corresponding transient MCPR distribution after the process of the invention transformation was applied. PDF 1204 is the reference ICPR distribution. PDF 1206 is the transient MCPR distribution when applying the currently approved process. PDF 1205 and PDF 1206 are very similar in both the most probably value of MCPR and the associated standard deviation of each distribution. Since there is a strong resemblance between the two resulting MCPR distributions, the transformation using the process of the invention is valid. It has been demonstrated that: (1) the xcex94CPR/ICPR is independent relative to the uncertainties that affect the ICPR, or the covariance is such that it is conservative to assume independence and (2) the transient MCPR distribution can be determined by applying the transient xcex94CPR/ICPR uncertainty to the rod ICPR distribution using the proposed approach. An example of the process of the invention is described by FIG. 13. In FIG. 13, histogram 1300 shows the number of rods 1302 of a certain CPR value versus the corresponding CPR value 1301. The PDF 1303 shows the resulting ICPR values from a set of approximately ninety-eight ICPR trials with all uncertainties applied. Ninety-eight new trials were run to generate a xcex94CPR/ICPR distribution for the specific transient event in order to translate the ICPR values to MCPR values. This xcex94CPR/ICPR distribution is not shown in FIG. 13. The xcex94CPR/ICPR distribution was applied using the process of the invention to the ICPR PDF 1303 to obtain the MCPR PDF 1304. The NRSBT was determined using the process of the invention, and the OLMCPR was determined to be 1.26. As a comparison, using the currently approved process, the SLMCPR was determined to be 1.10. Thus, the process described herein is more conservative than the first stage of the currently approved process. Ultimately, however, the currently approved process generates a unnecessarily conservative value after the error factor is added to the SLMCPR value, which yields a OLMCPR value needlessly larger than the process of the invention. Although the improved methods, as described herein below, are preferably implemented using a high speed data processing system capable of processing simulation routines that require highly accurate calculations and multiple reiterations, the present invention is not intended as limited to any one particular type of computer or data processing system. Any generic data processing system having sufficient, speed, storage memory and programmable computational capabilities for implementing statistical data analysis/reduction may be utilized to implement the present invention. |
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description | The present invention is directed to a radionuclide generation system and a method of producing radionuclides from irradiation targets in a commercial nuclear reactor core. Radionuclides are used in various fields of technology and science, as well as for medical purposes. Usually, radionuclides are produced in research reactors or cyclotrons. However, since the number of facilities for commercial production of radionuclides is limited already and expected to decrease, it is desired to provide alternative production sites. EP 1 667 166 A2 relates to a method of producing isotopes in a light water power reactor, whereby one or more targets within the reactor may be irradiated under a neutron flux to produce one or more isotopes. The targets may be assembled into a tubing portion of a fuel rod in one or more fuel bundles that are to be loaded in a core of the reactor at a given outage. Power operations in the reactor irradiate the fuel bundles so as to generate desired isotopes, such as one or more radioisotopes at a desired specific activity or stable isotopes at a desired concentration. The neutron flux density in the core of a commercial nuclear reactor is measured, inter alia, by introducing solid spherical probes into instrumentation tubes passing through the reactor core. It was therefore suggested that instrumentation tubes of commercial nuclear reactors shall be used for producing radionuclides. For example, EP 2 093 773 A2 suggests that existing instrumentation tubes conventionally used for housing neutron detectors may be used to generate radionuclides during normal operation of a commercial nuclear reactor. In particular, spherical irradiation targets are linearly pushed into and removed from the instrumentation tubes. Based on the axial neutron flux profile of the reactor core, the optimum position and exposure time of the targets in the reactor core are determined. A driving gear system is used for moving and holding the irradiation targets in the instrumentation tubes. US 2013/0315361 A1 also relates to an apparatus and methods for producing radioisotopes in multiple instrumentation tubes of operating commercial nuclear reactors. Irradiation targets are inserted and removed from multiple instrumentation tubes and converted to radioisotopes during operation of the nuclear reactor. Positioning irradiation targets are provided to properly position other irradiation targets at desired positions within or near the nuclear core. The positioning targets can be made of an inexpensive inert material or of a magnetic material, and may be held in the instrumentation tube by means of a magnetic latch. After irradiation, the targets are delivered from the instrumentation tube into a harvesting cask, and the positioning targets may be sorted out from the harvesting cask due to their markings or physical properties. WO 2014/107218 A2 discloses a retention assembly including a restricting structure, such as a fork, for selectively blocking the movement of irradiation targets through a pathway and into/out from instrumentation tubes. Positioning targets are provided to prop up the irradiation targets. The positioning targets may be ferromagnetic. A positioning detector is used to operate the restricting fork based on the presence of magnetic members such as the positioning targets. CA 2 792 593 A2 describes an apparatus and methods for producing radioisotopes in instrumentation tubes of operating commercial nuclear reactors. Irradiation targets are inserted and removed from instrumentation tubes during operation and converted to radioisotopes. The irradiation targets may further include a tracking target located at a known position among all other targets that is fabricated of a material that is different from all other targets and permits tracking or locating of the irradiation targets. For example, the first and last irradiation target may be fabricated of a ferromagnetic material that can be tracked with a magnet sensor, or may be fabricated of a material converting to a different isotope product that can be detected with a radiation sensor. Conventional spherical probes for use in a ball measuring system in the core of a commercial nuclear reactor are driven into and out from the instrumentation tubes using pressurized gas. Therefore, the irradiation targets designed for use in the instrumentation tubes must be able to withstand high mechanical loads. In addition, the targets are usually produced from parent material having high isotope purity. Preparation of the irradiation targets is therefore very expensive. However, the neutron flux density in the core of a commercial nuclear reactor is not homogeneous and may be insufficient for converting the irradiation targets to the desired radionuclide at various axial positions of the instrumentation tubes. Generally, the neutron flux density is higher in the middle section of the core than in the areas at the top or bottom thereof. In addition, spacer elements between the fuel rods may also block the neutron flux at specific axial positions in the core. Accordingly, a selective positioning of the targets is required to avoid waste of expensive parent material due to insufficient activation. If the expensive irradiation targets are positioned in the upper or lower part of the nuclear reactor core, the parent material will not be converted completely to the desired radionuclide. The incompletely activated irradiation targets cannot be used in industrial or medical applications and therefore will have to be sorted out and disposed or stored according to their half-life until a re-use is possible. Sorting out of the incompletely activated irradiation targets will increase safety hazards and can be done in hot cells only. Moreover, the time required for separating completely converted irradiation targets from incompletely activated targets causes a depletion of the radionuclides in the converted targets due to radioactive decay. In addition, disposal of the incompletely activated irradiation targets increases the amount of nuclear waste and thus the costs of producing the desired radionuclides. It is an object of the invention to provide a method of producing radionuclides from irradiation targets inserted into an instrumentation tube of a nuclear reactor core, such as the instrumentation tube of a ball measuring system or a Traversing Incore Probe (TIP) system of a commercial nuclear reactor which allows for an effective and economical production of radionuclides during reactor operation. According to another object, an optimum positioning of the irradiation targets in the nuclear reactor core and a fast recovery of the activated irradiation targets from the instrumentation tubes shall be achieved to completely convert the expensive irradiation targets and avoid production of nuclear waste due to incomplete target activation. The above objects are solved by a radionuclide generation method according to claim 1. Advantageous and expedient embodiments of the invention are indicated in the dependent claims which can be combined with each other independently. The invention provides a method of producing radionuclides from irradiation targets in a nuclear reactor core comprising the steps of: providing at least one instrumentation tube system including an instrumentation finger passing through a core of the nuclear reactor; inserting at least one irradiation target and at least one dummy target into the instrumentation finger and activating the irradiation target by exposing the irradiation target to neutron flux in the nuclear reactor core to form a radionuclide; retrieving the dummy target and the irradiation target from the instrumentation finger and separating the dummy target from the irradiation target; wherein the dummy target and the irradiation target are arranged in the instrumentation finger in a linear order such that the dummy target holds the irradiation target at a predetermined axial position in the reactor core, said axial position corresponding to a pre-calculated neutron flux density sufficient for converting the irradiation target to the radionuclide, wherein the dummy target and the irradiation target have different magnetic properties, and wherein the step of separating the dummy target from the irradiation target includes exposing the dummy target and/or the irradiation target to a magnetic field to retain either the dummy target or the irradiation target in the instrumentation tube system and release the other one of the irradiation target or the dummy target from the instrumentation tube system. In a further aspect, the invention provides a radionuclide generation system comprising: an instrumentation tube system of a nuclear reactor including at least one instrumentation finger passing through a core of the nuclear reactor wherein the instrumentation tube system is configured to permit insertion and removal of irradiation targets and dummy targets into the instrumentation finger, and wherein the dummy targets and the irradiation targets have different magnetic properties; a target drive system configured to insert the irradiation targets and dummy targets into the instrumentation finger in a predetermined linear order and to remove the irradiation targets and dummy targets from the instrumentation finger; a core monitoring system and an instrumentation and control unit linked to each other and configured to calculate an optimum axial irradiation position and time for the irradiation targets based on the actual state of the nuclear reactor as provided by the core monitoring system; a target removal system configured to retrieve the irradiation targets and dummy targets from the instrumentation tube system and separate the irradiation targets from the dummy targets, the target removal system including means for exposing the dummy targets and/or the irradiation targets to a magnetic field to retain either the dummy targets or the irradiation targets in the instrumentation tube system and release the other one of the irradiation targets or the dummy targets from the instrumentation tube system. The invention contemplates that a commercial nuclear reactor whose main purpose is the generation of electrical power can be used for producing radionuclides. In particular, existing or planned ball measuring systems or other instrumentation tube systems of such commercial reactors can be modified and/or supplemented to enable an effective and efficient production of radionuclides. According to the invention, at least some of the instrumentation tubes for example of a ball measuring system or a Transversing Incore Probe (TIP) system are used to guide the irradiation targets into the reactor core and to lead the irradiation targets out of the reactor core after complete activation of the parent material of the irradiation target. The activation of the targets is optimized by positioning the irradiation targets in predetermined areas of the reactor core having a neutron flux sufficient for converting the parent material completely into the desired radionuclide. The proper positioning of the irradiation targets is achieved by means of dummy targets made of an inert material and sequencing the dummy targets and the irradiation targets in the instrumentation tube system so as to form a column of the targets in the instrumentation finger wherein the irradiation targets are at a pre-calculated optimum axial position in the reactor core and the other positions are occupied by the inert dummy targets. The invention further suggests separating the activated irradiation targets from the dummy targets by means of different magnetic properties, for example by using a number of magnets positioned at a discharge section of the instrumentation tube system, in order to retain one of the dummy targets or irradiation targets within the tube system and deliver the other one of the dummy targets or the irradiation targets to a storage container. Separating the targets by means of different magnetic properties allows for a fast remote processing of the activated irradiation targets and may also avoid the use of additional mechanical systems such as counters, indexers and gateways to determine the exact position of the dummy targets and irradiation targets. The dummy targets can be made of inexpensive inert materials and can be re-used after a short decay time so that the amount of radioactive waste is further reduced. Accordingly, the optimum positioning of the irradiation targets by means of inexpensive dummy targets in an instrumentation tube system of a commercial nuclear reactor in combination with separating the activated irradiation targets from the dummy targets due to different magnetic properties provides an effective and economical production of radionuclides during reactor operation, and also avoids production of nuclear waste due to incomplete target activation. According to a preferred embodiment, the instrumentation tube system is derived from a conventional ball measuring system of a pressurized water reactor, such as an EPR™ or Siemens™ PWR nuclear reactor. As it is known to a person skilled in the art, the main purpose of a ball measuring system is to measure the neutron flux density in the core of the nuclear reactor. The ball measuring system comprises an instrumentation tube system and includes a pneumatically operated drive system configured to insert spherical probes, so called “aeroballs”, into the instrumentation finger of the tube system, which pass through the reactor core over the entire axial length of the core, and to remove the aeroballs from the instrumentation finger after activation of the probes and guide the probes through the instrumentation tube system to a detection room thereby preserving the linear order of the probes. The person skilled in the art will however recognize that the invention is not limited to use of a ball measuring system of a PWR reactor. Rather, it is also possible to use the instrumentation tubes of the Traversing Incore Probe (TIP) system of a boiling water reactor (BWR) or the view ports of a CANDU reactor. According to the invention, the dummy targets and the irradiation targets have different magnetic properties. Preferably, either the dummy targets or the irradiation targets are magnetically attractable. More preferably, the dummy targets are made from a ferromagnetic material such as ferrite, iron or iron alloys, including ferritic stainless steel. Most preferably, the irradiation targets and the dummy targets have a round shape, preferably a spherical shape, so that the targets are rollable and can easily be processed in the instrumentation tube system using pressurized gas such as air or nitrogen. The irradiation targets and the dummy targets are inserted into the instrumentation finger of the instrumentation tube system in the core of the nuclear reactor. Preferably, the instrumentation tube system comprises a discharge tube which is connected to the instrumentation finger and which is located outside the nuclear reactor core. More preferably, the discharge tube has an outlet coupled to a storage container for receiving one of the irradiation targets or dummy targets removed from the instrumentation tube system through the discharge tube. The target drive system is configured to insert the irradiation targets and dummy targets through the instrumentation tube system into the instrumentation finger in a predetermined linear order and to remove the irradiation targets and dummy targets from the instrumentation finger. According to a preferred embodiment, the target drive system is pneumatically operated allowing for a fast processing of the irradiation targets and the dummy targets using pressurized gas such as nitrogen or air. More preferably, the target drive system comprises one or more pneumatically operated valve batteries for separate control of the insertion and transport of the irradiation targets and dummy targets in the instrumentation tube system. Based on this separation of controls, the regular ball measuring system for determining the neutron flux in the core and the radionuclide generation system according to the invention can be driven separately. The instrumentation and control system is configured such that operation of the valves of the target drive system is at least partly automated to achieve a safe and reliable operation of the target drive system. The target drive system may further comprise a gate device for discharging the irradiation targets and/or the dummy targets into a storage container. According to a preferred embodiment of the invention, the target drive system comprises sensors for monitoring the presence and runtime of the irradiation targets and the dummy targets passing through the instrumentation tube system, in particular in and out of the instrumentation finger. More preferably, the sensors of the target drive system are responsive to a variation of the magnetic flux as the irradiation targets and/or the dummy targets pass by the sensors. In addition, or as an alternative, activity sensors sensing the radiation of the irradiation targets and/or the dummy targets may be used. According to a preferred embodiment, the core monitoring system to and the instrumentation and control unit are configured such that the irradiation process for converting the irradiation target to the radionuclide is optimized by considering the actual state of the reactor, especially the current neutron flux, fuel burn-up, reactor power and/or loading. Thus, an optimum axial irradiation position and irradiation time can be calculated for optimum results. It is however not important whether the actual calculation is performed in the instrumentation and control unit or by the adapted core monitoring system of the ball measuring system. Accordingly, both alternatives are subject of the invention. Preferably, the information provided by the core monitoring system to the instrumentation and control unit includes at least one of the following: neutron flux (from ex- or in-core detectors), activation values from an existing ball measuring system, burn-up, reactor power, loading, rod position(s), flow rate, inlet-temperature, pressure, and time synchronization. The more information about the reactor is considered as input data, the more accurate will be the results of the calculation of the optimum axial irradiation position and irradiation time. The before mentioned parameters may include real-time values and any derivatives, like developments over time. The instrumentation and control unit may be advantageously configured to automatically control the pressure in the instrumentation tube system, in particular after each insertion of irradiation targets and/or dummy targets by the target drive system. The target removal system preferably includes at least one magnet, more preferably at least one magnet selected from at least one of a permanent magnet or a solenoid, arranged at the instrumentation tube system as a means to expose the targets to a magnetic field. According to a preferred embodiment, the instrumentation tube system comprises the discharge tube connected to the instrumentation finger, and the magnet or solenoid is arranged at the discharge tube. Most preferably, the at least on magnet is movable along a longitudinal axis of the discharge tube. Still more preferably, the target removal system comprises a magnetic lock arranged at the discharge tube for blocking flow of the irradiation targets and/or dummy targets out of the discharge tube into a storage container. Preferably, the magnetic lock is arranged proximate the outlet of the discharge tube, downstream of the at least one magnet with respect to the flow direction of the irradiation targets out of the discharge tube. More preferably, the magnetic lock comprises a retainer means such as a pin crossing the discharge tube so that the targets are retained in the discharge tube until the at least one magnet is activated. Most preferably, the at least one magnet and/or the magnetic lock are remotely controlled by the instrumentation and control unit so that an automatic and fast processing of the irradiation targets is achieved. In the method of the present invention, the irradiation targets are preferably positioned in the center section of the instrumentation finger in the nuclear reactor core, and the dummy targets are preferably positioned in at least one end section of the instrumentation finger, i. e., in the upper and/or lower part of the reactor core where the neutron flux density is insufficient for completely activating the irradiation targets. The exact position of the irradiation targets in the instrumentation finger is determined by the core monitoring system and/or instrumentation and control unit. As a general rule, the center section of the instrumentation finger useful for radionuclide generation extends over about 3-4 meters, and the end portions extend over 0.5 to 1 meter in a commercial pressurized water reactor. These values may vary according to the reactor type and the actual operation status of the reactor, and will be different for boiling water reactors and CANDU reactors, respectively. In a preferred embodiment, the instrumentation tube system penetrates the pressure vessel cover of the nuclear reactor, and the irradiation targets and dummy targets are inserted into the instrumentation finger from the top of the reactor core through the pressure vessel cover. Preferably, the dummy targets are arranged at the lower end section of the instrumentation fingers, and the irradiation targets are arranged above the dummy targets. Most preferably, the space above the irradiation targets will remain empty. According to another embodiment, the irradiation targets in the instrumentation finger may be separated by one or more dummy targets thereby defining irradiation target subsections. The irradiation targets in the irradiation target subsections preferably have the same or different material properties. More preferably, the irradiation targets in adjacent subsections differ with respect to the parent material used as a precursor for radionuclide generation. Thus, it is possible to produce different radionuclides in a one-step operation. Alternatively, it is also possible to insert the irradiation targets and dummy targets to the instrumentation finger from a drywell at the bottom of the reactor core, for example if the TIP system of a boiling water reactor is used as the instrumentation tube system. In this case, additional means for retaining the irradiation targets and dummy targets in the instrumentation finger are provided. The irradiation targets are activated in the instrumentation finger for a period of time sufficient for complete conversion of the parent material of the irradiation targets to the desired radionuclide, as determined by the online core monitoring system and instrumentation and control system. The time to achieve complete conversion of the parent material will depend on the reactor type and status, neutron flux conditions, type of parent material and various other parameters known to a person skilled in the art, and may range from several hours to days, preferably 1 to 10 days. Complete conversion means a conversion rate of the parent material providing a radionuclide content suitable for medical or industrial application of the irradiation targets. After activation, the irradiation targets and dummy targets are preferably transferred out of the instrumentation fingers into the discharge tube, more preferably using pressurized gas, by preserving the linear order of the dummy targets and the irradiation targets. The discharge tube is blocked by the magnetic lock so as to prevent the irradiation targets and dummy targets from leaving the instrumentation tube system. For separating the irradiation targets from the dummy targets and selectively removing the irradiation targets from the instrumentation tube system, the irradiation targets and the dummy targets are exposed to a magnet field, preferably by remotely activating the one or more magnets arranged at the discharge tube adjacent the magnetic targets. The magnetic lock is then opened, and the non-magnetic targets, preferably non-magnetic irradiation targets, are released from the discharge tube under the action of gravity and passed into a storage container whereas the magnetic dummy targets are retained in the discharge tube by the action of the magnetic field. In a further embodiment, if one ore more magnetic targets, such as the dummy targets, are positioned ahead of the non-magnetic targets, such as the irradiation targets, proximate an outlet of the discharge tube, pressurized gas is used to drive the non-magnetic irradiation targets back into the instrumentation finger or a holding section in the instrumentation tube system while the magnetic dummy targets are retained in the discharge tube by means of the activated magnetic field. The magnetic lock can then be opened and the dummy targets be released from the discharge tube by switching off the magnetic field. The method of the invention will also be applicable vice-versa if the dummy targets are non-magnetic and the irradiation targets are ferromagnetic. The invention thus provides a fast and effective method to completely activate the irradiation targets in the instrumentation fingers during normal reactor operation, and separating the activated irradiation targets from inert dummy targets by means of their different magnetic properties. FIG. 1 illustrates the basic setup of a radionuclide generation system within a commercial nuclear power plant, in particular a power plant with an EPR™ or Siemens DWR nuclear reactor. As opposed to a research reactor, the purpose of a commercial nuclear reactor is the production of electrical power. Commercial nuclear reactors typically have a power rating of 100+ Megawatt electric. The basis of the radionuclide generation system described in the example embodiments is derived from a commercial ball measuring system, the main purpose of which is to measure the neutron flux density in the core of the nuclear reactor using a plurality of spherical probes, so called aeroballs, arranged in a linear order thereby forming an aero ball column. The ball measuring system includes a pneumatically operated drive system configured to insert the aeroballs into an instrumentation finger extending into and passing the core through its entire axial length, and to remove the aeroballs from the instrumentation finger after activation. For easier reference, hereafter the radionuclide generation system based on a commercial ball measuring system will be also referred to as MAS (Medical Aeroball System). In the following, the major components of the MAS, which are provided in addition to those of the commercial ball measurement system, or which are supplemented or modified, will be further described. FIG. 1 shows a commercial nuclear reactor comprising an instrumentation tube system 12 including at least one instrumentation finger 14 passing through a core 10 of the nuclear reactor. The instrumentation tube system 12 is configured to permit insertion and removal of irradiation targets 16 and dummy targets 18 (cf. FIG. 3) into the instrumentation finger 14. According to the present invention, the commercial ball measuring system of the pressurized water reactor is adapted to also handle irradiation targets 16 and dummy targets 18 having a round or spherical shape and having a diameter corresponding to the clearance of the instrumentation finger of the ball measuring system. Preferably, the diameter of the targets 16, 18 is in the range of between 1 to 3 mm, preferably about 1.7 mm. The instrumentation tube system 12 penetrates a reactor confinement 11 and the pressure vessel cover of the nuclear reactor, with the instrumentation finger 14 extending from the top to the bottom of the reactor core 10 over substantially the entire axial length of the reactor core 10. An end of the instrumentation finger 14 at the bottom of the reactor core 10 is closed and/or provided with a stop so that the irradiation targets 16 and dummy targets 18 inserted into the instrumentation finger form a column wherein each target 16, 18 is at a predefined axial position. A separation component (not shown) may be provided for splitting the tubes of the instrumentation tube system 12 at a cable bridge above the reactor pressure vessel cover, via which the tubes are led out of the reactor confinement 11, and/or at a connector board. Preferably, several humidity sensors are provided in the instrumentation tube system 12 to detect any ingress of primary coolant (or any other liquid) into the MAS. It is understood that the instrumentation fingers 14 used for the MAS are in direct contact with the primary cooling water surrounding the core of the nuclear reactor. The humidity sensors may be based on spark plugs which are modified for measuring electrical resistance. Further sensors are preferably provided for monitoring the presence and runtime of the targets 16, 18 passing through the instrumentation tube system. These sensors are preferably arranged at the tubes penetrating the reactor core 10. The measuring principle may be based on the detection of a variation of the magnetic flux as the irradiation targets 16 and/or the dummy targets 18 pass by the sensors for measuring transport time and completeness indication. Preferably, the sensors are used to monitor that all irradiation targets 16 have left the instrumentation finger 14 during the removal process while the targets pass the sensors. In addition, or as an alternative, activity sensors sensing the radiation of the irradiation targets 16 and/or the dummy targets 18 may be used. The instrumentation tube system 12 further comprises a discharge tube 34 which is connected to the instrumentation finger 14 and which is located outside the nuclear reactor core 10. The irradiation targets 16 or dummy targets 18 are removed from the instrumentation tube system 12 through the discharge tube 34. The instrumentation tube system 12 is connected to a target drive system 20 configured to insert the irradiation targets 16 and dummy targets 18 into the instrumentation finger 14 in a predetermined linear order and to drive the irradiation targets 16 and dummy targets 18 out of the instrumentation finger 14. Preferably, the target drive system 20 is pneumatically operated allowing for a fast processing of the irradiation targets 16 and the dummy targets 18 using pressurized gas such as nitrogen or air. Preferably, the target drive system 20 comprises a valve battery (not shown) which is used as an additional pneumatic system for separate control of the targets 16, 18 in the irradiation tube system 12. The valve battery may be implemented as a further subsystem in addition to the valve batteries of the conventional ball measuring system, or a separate target drive system is installed. The target drive system 20 preferably comprises a target filling device (not shown) for inserting the targets 16, 18 into the instrumentation finger 14. A gate system (not shown) including several (electro)-mechanical devices may be used to fill the irradiation targets 16 and dummy targets 18 into the instrumentation tube system 12 for transport to the reactor core 10 and also guide the targets 16, 18 to the discharge tube 34 and the storage containers 36 (FIG. 4). The target drive system 20 cooperates with a target removal system 22 configured to retrieve the irradiation targets 16 and dummy targets 18 from the instrumentation tube system 12 and separate the irradiation targets 16 from the dummy targets 18. In particular, the target removal system 22 includes means 24 for exposing the dummy targets 18 and/or the irradiation targets 16 to a magnetic field to retain either the dummy targets 18 or the irradiation targets 16 in the instrumentation tube system 12 and release the other one of the irradiation targets 16 or the dummy targets 18 from the instrumentation tube system 12 (cf. FIG. 4). An instrumentation and control unit (ICU) 26 is linked to the target drive system 20 and the target removal system 22 as well as an online core monitoring system 28 for controlling activation of the irradiation targets 16, and a fault monitoring system 30. The ICU 26 and/or the online core monitoring system 28 are configured to calculate an optimum axial irradiation position and time for the irradiation targets 16 based on the actual state of the nuclear reactor as provided by the online core monitoring system 28. The ICU 26 is connected via an interface with the adapted online core monitoring system 28 software. The ICU 26 is further connected to the mechanical components of the MAS, including the sensors. For an efficient generation of radionuclides, optimum irradiation conditions and time for the irradiation targets are determined. Practically all relevant input data for this calculation are available from the online core monitoring system 28 of the conventional ball measuring system, for example the POWERTRAX/S™ core monitoring software system available from Areva. The ICU 26, which is linked to the core monitoring system 28 can calculate the optimum irradiation time and axial irradiation position, as well as further parameters, like the amount of irradiation targets 16 in an instrumentation finger 14 defining the actual length of the respective target column and the positions of the individual irradiation targets 16 and dummy targets 18 within the target column. Based on the results of these calculations the ICU 26 and/or an operator operate the mechanical MAS components accordingly. The ICU 26 is also connected to an updated fault monitoring system 30 of the ball measuring system for reporting any errors in the MAS. The online calculation of the optimum irradiation time and axial irradiation position of the irradiation targets is not simply based on the assumption of an estimated constant neutron flux, but rather takes the actual state of the reactor into account, especially at least one of the following parameters: neutron flux, activation values from an existing ball measuring system, burn-up, reactor power, loading, rod position(s), flow rate, inlet-temperature, pressure, and time synchronization. Not only real-time values of these parameters, but also their development over time may be considered. FIG. 2 schematically depicts a diagram providing information on the equipment of the reactor core 10 with MAS instrumentation fingers 14, conventional ball measuring and neutron detector subsystems and their distribution within the core 10 of the nuclear reactor. According to the example shown in FIG. 2, four positions are taken from the conventional ball measuring and neutron detector systems for use in the MAS. FIG. 3 shows a simplified illustration of an instrumentation finger 14 which is used for the MAS. As described above, the instrumentation finger 14 extends from the top to the bottom of the reactor core 10 over substantially the entire axial length of the reactor core 10. The irradiation targets 16 and dummy targets 18 are inserted into the instrumentation finger in a linear order to form a column wherein each target 16, 18 is at a predefined axial position. With the aid of the online core monitoring system 28 it is possible to determine sections 38, 44 of the instrumentation finger 14 in which the neutron flux is too low for producing radionuclides, and sections 40 where the neutron flux is above the required irradiation target demand and thus suitable for producing the desired radionuclides. A humidity sensor 42 is arranged at the instrumentation finger 10. Alternatively or in addition, humidity sensors can be arranged at components of the instrumentation tube system 12 outside the reactor pressure vessel. In order to eliminate any waste of expensive irradiation targets 16, dummy targets 18 are provided and positioned in the lower end section 38 of the irradiation finger 14 having a too low neutron flux density for radionuclide generation. The dummy targets 18 are made of an inert material which is not substantially activated under the conditions in the core 10 of an operating nuclear reactor. The irradiation targets 16 comprise a suitable parent material for generating radionuclides which are to be used for medical and/or other purposes. More preferably, the irradiation targets consist of the parent material which converts to a desired radionuclide upon activating by exposure to neutron flux present in the core of an operating commercial nuclear reactor. Useful parent materials are Mo98 and Yb176 which are converted to Mo99 and Lu177, respectively. It is understood, however, that the invention is not limited to the use of a specific parent material. As shown in FIG. 3, the irradiation targets 16 are positioned above and are held in place by the dummy targets 18 in the sections 40 of the irradiation finger 14 where the neutron flux is sufficient to completely convert the irradiation targets into the desired radionuclides, as determined by the ICU 26 and/or the online core monitoring system 28. In accordance with a preferred embodiment of the invention, an upper section 44 of the instrumentation finger 14 is kept empty. In a further embodiment (not shown), one or more of the irradiation targets 16 in the instrumentation finger 14 may be separated from each other by one or more dummy targets 18 thereby defining irradiation target subsections. The irradiation targets in the irradiation target subsections preferably have the same or different material properties. More preferably, the irradiation targets in adjacent subsections differ with respect to the parent material used as a precursor for radionuclide generation. According to the invention, the dummy targets 18 and the irradiation targets 16 have different magnetic properties. Preferably, either the dummy targets or the irradiation targets are magnetically attractable. More preferably, the dummy targets 18 are made from a ferromagnetic material such as iron or iron alloys, including ferritic stainless steel, or ferrite. For use in a conventional ball measuring system, the irradiation targets 16 and the dummy targets 18 have a round shape, preferably a spherical shape, so that the targets are rollable and can easily be processed in the instrumentation tubes of the ball measuring system by pressurized air and/or gravity. The target removal system 22 is schematically shown in FIG. 4. The means 24 for exposing the irradiation targets 16 and/or dummy targets 18 to a magnetic field preferably include one or more permanent magnets or solenoids 46 associated to the irradiation tube system 12. A discharge tube 34 is connected to the instrumentation finger 14 through the instrumentation tube system 12 (FIG. 1) and is configured to receive the irradiation targets 16 and dummy targets 18 driven out of the instrumentation finger after activation is completed. The discharge tube 34 is located outside the reactor core 10. One or more solenoids 46 are surrounding the discharge tube 34. Preferably, the solenoids 46 are movably arranged along a longitudinal axis of the discharge tube 34. The discharge tube 34 is further provided with a magnetic lock 48 for blocking flow of the irradiation targets 16 and/or dummy targets 18 out of the discharge tube. In the embodiment shown in FIG. 4, the magnetic lock 48 comprises a pin 50 crossing the discharge tube 34 so that the targets 16, 18 are retained in the discharge tube 34 until the electromagnetic coils 46 are activated. The one or more magnets or solenoids 46 and/or the magnetic lock 48 are remotely controlled by the ICU 26 so that an automatic and fast processing of the irradiation targets 16 and the dummy targets 18 is achieved. The discharge tube 34 further comprises a gas inlet 52 linked to the target drive system 20 to blow pressurized gas such as air or nitrogen into the instrumentation tube system 12 and drive the irradiation targets 16 and dummy targets 18 out of the instrumentation finger 14 into the discharge tube 34. In addition, the discharge tube 35 comprises an outlet 54 coupled to one or more storage containers 36, 36′ and/or an intermediate storage tank 56 to selectively receive the irradiation targets 16 or dummy targets 18 released from the discharge tube 34. The storage containers 36, 36′ and or the intermediate storage tank 56 may be movable to match with the outlet 54. Alternatively, the outlet may include a gate 58 to selectively direct the irradiation targets 16 or dummy targets 18 to the appropriate container 36, 36′ and tank 56. For operating the radionuclide generation system or MAS of the present invention, the irradiation targets 16 and the dummy targets 18 are inserted into the instrumentation finger 14 using the target drive system 20, and the irradiation targets 16 are activated by exposure to neutron flux in the nuclear reactor core when in power generating operation to form a radionuclide. Preferably, the dummy targets 18 are made of a ferromagnetic material, and the irradiation targets 16 are substantially non-magnetic or paramagnetic. The dummy targets 18 and the irradiation targets 16 are arranged in the instrumentation finger 14 in a pre-calculated linear order such that the dummy targets 18 hold the irradiation targets 16 at a predetermined axial position in the reactor core. The optimum axial position of the irradiation targets 16 is calculated by the ICU 26 and/or the online core monitoring system 28 and corresponds to a neutron flux density sufficient for completely converting the irradiation targets 16 to the radionuclide during a predetermined period of time. The remaining positions in the instrumentation finger 14 are occupied by the dummy targets 18 which hold the irradiation targets 16 in place. After activation of the irradiation targets 16 and conversion to the desired radionuclide, the dummy targets 18 and the activated irradiation targets 16 are retrieved from the instrumentation finger 14, and the dummy targets 18 are separated from the activated irradiation targets 16 using the target removal system 22. The dummy targets 18 and/or the activated irradiation targets 16 are exposed to a magnetic field to retain either the dummy targets 18 or the activated irradiation targets 16 in the instrumentation tube system 12 and release the other one of the activated irradiation targets 16 or the dummy targets 18 from the instrumentation tube system 12. Operation of the MAS may is preferably monitored and controlled at an operator station via a process unit. The process unit is installed at a separate control cabinet in a control cabinet room (not shown). The process unit is equipped with a display and, inter alia, allows to control specific parameters of the MAS valve batteries. At the operator station the state of the irradiation targets 16 during irradiation and the remaining irradiation time can be monitored. When the calculated irradiation time of a set of targets 16, 18 in an instrumentation finger 14 is reached, a message prompts the operator to start the removal process with respect to this instrumentation finger 14. The operation of the various valves of the target drive system 20 is partly automated so that repeating actions are performed more safely and more reliably. After each insertion of irradiation targets 16 and dummy targets 18 into the instrumentation tube system 12, the pressure in the tube system is checked and regulated in a fully automated manner. The ICU 26 also collects further digital signals representative of certain system conditions. Especially, the signals of the humidity sensors allow a leakage monitoring, i. e. to detect whether any primary coolant has entered the tube system of the MAS. The electric power for the MAS components, including the valve batteries and the process unit of the MAS cabinet, is provided by a load cabinet of the ball measuring system. To this end, a further power inverter with appropriate fuses is installed in the load cabinet. It is also possible to use an additional 24 volt supply incorporated in the control cabinet room. The irradiation targets 16 are preferably positioned in the center section 40 of the instrumentation finger 14 in the reactor core 10, and the dummy targets 18 are preferably positioned in the end section 38 and/or the upper section 44 of the instrumentation finger 14, i. e., in the upper and/or lower part of the reactor core 10 where the neutron flux density is insufficient for completely activating the irradiation targets 16 (cf. FIG. 3). The exact position of the irradiation targets 16 in instrumentation finger 14 is pre-calculated by the ICU 26 and/or the online core monitoring system 28. The irradiation targets 16 and dummy targets 18 are preferably inserted into the instrumentation finger 14 from the top of the reactor core 10 through the pressure vessel cover. Preferably, the dummy targets 18 are arranged at the lower portion of the instrumentation finger 14, and the irradiation targets 16 are arranged above the dummy targets 18. As shown in FIG. 3, the space above the irradiation targets 16 in the upper section 44 can remain empty. After activation is completed, the irradiation targets 16 and the dummy targets 18 are transferred out of the instrumentation finger 14 into the discharge tube 34 using pressurized gas such as nitrogen or air provided by the target drive system 20 (cf. FIG. 4). The linear order of the dummy targets 18 and the irradiation targets 16 in the instrumentation finger 14 is preserved in the discharge tube 34 so that the irradiation targets 16 are proximate to the outlet 54 of the discharge tube 34. The discharge tube 34 is blocked by the magnetic lock 48 providing a stop for the targets 16, 18 and to prevent the irradiation targets 16 and dummy targets 18 from leaving the tube system 12. The valves of the drive system 20 are then closed and the pressure in the instrumentation tube system 12 is relieved. For separating the irradiation targets 16 from the dummy targets 18 and selectively removing the irradiation targets 16 from the discharge tube 34, the irradiation targets 16 and the dummy targets 18 are exposed to a magnet field created by activating the solenoids 46 arranged at the discharge tube. More preferably, the solenoids 46 are moved along the longitudinal axis of the discharge tube 34 and arranged adjacent to the ferromagnetic dummy targets 18 so that each dummy target 18 is associated to and exposed to the magnetic field of a solenoid 46. The magnetic lock 48 is then opened, and the non-magnetic irradiation targets 16 are released from the discharge tube 34 under the action of gravity and passed into the storage container 36 for further processing and transportation to the application site whereas the magnetic dummy targets 18 are kept in the discharge tube 34 by the action of the magnetic field generated by the solenoids 46. The solenoids 26 arranged at the discharge tube 34 can also be used for slowing down the targets 16, 18 when driven out of the instrumentation finger 14 in order to prevent the targets from being damaged. Alternatively, it is possible to stop the targets 16, 18 in the discharge tube using the solenoids 46 only, without activating the magnetic lock 48. After the irradiation targets 16 are separated from the dummy targets 18 and are harvested in the storage container 36, the outlet is coupled to the intermediate storage tank 56, the magnetic field is switched off and the dummy targets 18 are transferred to the intermediate storage tank 56 under the action of gravity for further use after a short decay period. Alternatively, some or all of the dummy 18 targets can be driven back into the instrumentation finger 14 using pressurized gas from the target drive system 20, and new irradiation targets 16 be inserted into the instrumentation tube system. According to another embodiment, the irradiation targets 16 in the instrumentation finger 14 may be separated by one or more dummy targets 18 thereby defining irradiation target subsections. The irradiation targets 16 in the irradiation target subsections can have the same or different material properties. Also in this embodiment, the activated irradiation targets 16 and dummy targets 18 are driven out of the instrumentation finger 14 into the discharge tube 34 preserving the linear order of the dummy targets 18 and the irradiation targets 16. The discharge tube 34 is blocked by the magnetic lock 48 so as to prevent the irradiation targets and dummy targets from leaving the tube system 12. If one or more ferromagnetic dummy targets 18 are now positioned ahead of the irradiation targets 16 proximate the outlet 54, the solenoids 46 are arranged adjacent these dummy targets 18. Pressurized gas from the target drive system 20 is used to drive the non-magnetic irradiation targets 16 back into the instrumentation finger 14 or a holding section in the instrumentation tube system 12 while the magnetic dummy targets 18 are retained in the discharge tube 34 by means of the activated magnetic field. The magnetic field is then switched off, the magnetic lock 48 is opened, and the dummy targets 18 are released from the discharge tube 34 under the action of gravity and transferred into the intermediate storage tank 56 coupled to the outlet 54. In the next step, the magnetic lock 48 is closed and the irradiation targets 16 and the remaining dummy targets 18 are driven out of the instrumentation finger 14 or holding section using pressurized gas from the target drive system 20. The activated irradiation targets 16 are now positioned proximate the magnetic lock 48 and the discharge tube outlet 54 and can be separated from the dummy targets 18 as described above. If necessary, the separation steps can be repeated until all irradiation targets are selectively harvested from the tube system in the respective storage containers 36, 36′. The method of the invention will also be applicable vice-versa if the dummy targets 18 are non-magnetic and the irradiation targets 16 are ferromagnetic. The radionuclide generation system according to the invention can also be installed in a nuclear power plant having no conventional ball measuring system. The ball measuring system as described above only provides a basis to facilitate an installation of the radionuclide generation system since no additional instrumentation tubes, fingers and the like need to be installed only for the MAS. Possible reactor types for such an application include boiling water reactors and CANDU (CANada Deuterium Uranium) reactors. |
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abstract | An optical fiber penetration is disposed in a sleeve provided through a partition wall that separates a first space and a second space. The optical fiber penetration includes a first optical fiber cable and a second optical fiber cable each having a thin tube formed of metal and an optical fiber strand inserted in the thin tube, a cylindrical body that is formed of metal and is disposed in an axial direction of the sleeve, an interior of which includes the first optical fiber cable on a side of the first space and the second optical fiber cable on a side of the second space, an internal connector configured to connect the first optical fiber cable with the second optical fiber cable in the interior of the cylindrical body, and a first lid and a second lid configured to close one end and the other end of the cylindrical body. |
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claims | 1. A neutron source rod assembly for a nuclear reactor comprising;a neutron source positioning rodlet assembly having an upper coupling for connecting to a mounting assembly and an elongated substantially round body sized to slidingly fit within a guide thimble tube of a nuclear fuel assembly, the body extending a preselected distance from the upper coupling along an axis coinciding with an elongated dimension of the substantially round body and terminating in a lower coupling; anda source capsule assembly sealably enclosing a neutron source material, with the source capsule assembly configured as a separate component from the neutron source positioning rodlet assembly, structured to be boxed and shipped separately, sized to slidingly fit within the guide thimble tube and having an upper coupling configured to mate with the lower coupling of the neutron source positioning rodlet assembly and wherein the neutron source positioning rodlet assembly and the source capsule assembly are configured to be fixedly connected to each other at an assembly site and the preselected distance substantially extends from the mounting assembly to an elevation in a core of the nuclear reactor at which the neutron source material is to be situated. 2. The neutron source rod assembly of claim 1 wherein the neutron source positioning rodlet assembly lower coupling and the source capsule assembly upper coupling are mechanical couplings that are configured to mate with each other. 3. The neutron source rod assembly of claim 2 wherein the neutron source positioning rodlet assembly lower coupling is one of either a male or female threaded coupling and the source capsule assembly upper coupling is another of the male or female threaded coupling that the one of either the male or female threaded coupling on the neutron source positioning rodlet assembly lower coupling will mate with. 4. The neutron source rod assembly of claim 1 wherein the source capsule assembly has an elongated axial dimension that is sufficient to contain sufficient neutron source material to provide the neutrons required for a specified X-core source range detector count rate during startup for the nuclear reactor, but is substantially shorter than the elongated dimension of the neutron source positioning rodlet assembly. 5. The neutron source rod assembly of claim 1 wherein the neutron source positioning rodlet assembly is a solid rod. 6. The neutron source rod assembly of claim 1 wherein the neutron source positioning rodlet assembly is at least in part formed from a hollow tube that is capped at each end. 7. The neutron source rod assembly of claim 6 wherein the neutron source material is a primary neutron source and the hollow tube contains a secondary neutron source material. 8. The neutron source rod assembly of claim 1 wherein the lower coupling of the neutron source positioning rodlet assembly and the upper coupling of the source capsule assembly lock together when fully connected. 9. The neutron source rod assembly of claim 8 wherein the lower coupling of the neutron source positioning rodlet assembly and the upper coupling of the source capsule assembly are threaded couplings and lock together when fully tightened. 10. A method of manufacturing a neutron source rod assembly comprising a mounting assembly, a plurality of thimble plug rodlets or a plurality of burnable neutron poison rodlets, a neutron source positioning rodlet assembly and a source capsule assembly, comprising the steps of:manufacturing the mounting assembly, the plurality of thimble plug rodlets or the plurality of burnable neutron poison rodlets and the neutron source positioning rodlet assembly in a first manufacturing facility remote from an assembly site at which the neutron source rod assembly is intended to inserted into a fuel assembly, wherein the neutron source positioning rodlet assembly comprises an upper coupling for connecting to the mounting assembly and an elongated substantially round body sized to slidingly fit within a guide thimble tube of a nuclear fuel assembly, the body extending a preselected distance from the upper coupling along an axis coinciding with an elongated dimension of the substantially round body and terminating in a lower coupling;manufacturing the source capsule assembly in a second manufacturing facility remote from the assembly site, the source capsule assembly sealably enclosing a neutron source material, with the source capsule assembly sized to slidingly fit within the guide thimble tube at the assembly site and having an upper coupling configured to mate with the lower coupling of the neutron source positioning rodlet assembly and wherein the preselected distance substantially extends from the mounting assembly to an elevation in the fuel assembly at which the neutron source material is to be situated;shipping the mounting assembly, the plurality of thimble plug rodlets or the plurality of burnable neutron poison rodlets and the neutron source positioning rodlet assembly to the assembly site;shipping the source capsule assembly to the assembly site;assembling the source capsule assembly to the lower coupling of the neutron source positioning rodlet assembly at the assembly site; andinserting the neutron source rod assembly into a fuel assembly at the assembly site. 11. The method of claim 10 wherein the assembly site is a nuclear reactor at which the neutron source rod assembly is used. 12. The method of claim 10 wherein the assembly site is a nuclear fuel assembly manufacturing facility. 13. The method of claim 10 wherein the mounting assembly, the plurality of thimble plug rodlets or the plurality of burnable neutron poison rodlets and the neutron source positioning rodlet assembly are assembled together at the first manufacturing facility. 14. The method of claim 10 wherein the mounting assembly, the plurality of thimble plug rodlets or the plurality of burnable neutron poison rodlets and the neutron source positioning rodlet assembly are assembled together at the assembly site. |
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description | As used herein, the term xe2x80x9cfloatablyxe2x80x9d or xe2x80x9cfloatablexe2x80x9d indicates a freedom to move, a small amount, in several directions while still being retained. That is, when directed to an ultrasonic probe floatably mounted in a housing, floatable indicates that the probe, while still being retained by the housing, has the freedom to roll, pitch, and yaw, as well as freedom to move laterally. As shown in FIG. 1, an ultrasonic testing end effector 10 in accordance with the present invention is shown. The end effector is composed of two main components the wrist assembly 12 and a probe assembly 14. The wrist assembly 12, as shown in FIG. 2, includes the wrist motor assembly 20 and its housing 30, and the wrist shaft 40 and its housing 50. The probe assembly 14, as shown in FIG. 3, includes the clutch assembly 60 and its housing 70, a probe motor assembly 100 and its housing 110, a probe carriage assembly 130 and its housing 150, the probe 160, and camera assembly 180. Each of these assemblies will be described fully below. However, generally speaking, the probe assembly 14 rotates in a horizontal plane below the wrist assembly 12. As shown in FIG. 2, the end effector 10 is coupled to a robotic arm by mounting bracket 16. Mounting bracket 16 supports the wrist motor assembly 20. The wrist motor assembly includes a wrist motor 22 which has a resolver 26 and a wrist axle 24 ending in a motor gear 28. The wrist motor 22, the upper portion of the wrist axle 24, and resolver 26 are enclosed within a wrist motor assembly housing 30. The wrist motor 22 can be any type of common motor that drives a rotating axle. The resolver 26 is a device which tracks the rotational motion of the wrist axle so that the angular orientation of the wrist axle 22 can be determined electronically by such means as a computer. The wrist motor assembly housing 30 has an upper surface 34 and a lower surface 36. The upper surface 34 has a medial hole therethrough (not shown). A wrist motor assembly coupling 32 is sealably connected to the upper surface 34 about the medial hole and provides a port which is coupleable to a power source and pressurized air hose. The coupling 32 also provides access for data wires connecting the resolver 26 to a computer or digital converter. The wrist motor assembly housing lower surface 36 has a medial hole therethrough which allows the lower portion of the wrist axle 24 to pass through. The wrist axle 24 could be coupled with the clutch assembly 60 (described below) directly. In the preferred embodiment, however, the wrist axle motor gear 28 is rotatably coupled with a wrist shaft 40. The wrist shaft 40 is cylindrical and includes a wrist shaft body 41, a wrist gear 42 and a wrist shaft lower end 44. The wrist shaft lower end 44 may have a greater circumference than the wrist shaft body 41. The teeth of the wrist gear 42 are rotatably coupled with the teeth of the motor gear 28. The wrist shaft lower end 44 is generally circular except for a wrist shaft detent 46. The motor gear 28, wrist shaft body 41 and wrist gear 42 are enclosed within the wrist shaft housing 50. The wrist shaft housing 50 has an upper end 52 and a lower end 54. Both the wrist shaft housing upper end 52 and the wrist shaft housing lower end 54 have openings therethrough. A wrist motor housing seal 38 is disposed between the wrist motor assembly housing 30 and the wrist shaft housing upper surface 52. The wrist motor housing seal provides a water tight seal between the wrist motor assembly housing 30 and the wrist shaft housing 50. The wrist motor assembly 20 is attached to the wrist shaft housing upper end 52 with the wrist axle 24 passing through the wrist shaft housing upper end 52 opening. A wrist axle seal 25 is disposed between the wrist shaft housing 50 and the wrist axle 24. The motor gear 28 is disposed at the lower end of the wrist axle 24 within the wrist shaft housing 50. The wrist shaft 40 is aligned in parallel with the wrist axle 24. The wrist shaft body 41 extends through the wrist shaft housing 50 and passes through the opening in the wrist shaft housing lower surface 54. As shown in FIG. 3, a clutch pin assembly 60, which includes a clutch pin 62, a clutch pin housing 64, and a clutch pin spring 66, is enclosed within the clutch assembly housing 70. The clutch assembly housing includes a clutch assembly housing upper surface 74 with an opening therethrough, a clutch assembly housing lower surface 76 and a clutch assembly housing back plate 72. The clutch assembly housing upper surface 74 is rotatably connected to the wrist shaft housing lower surface 54. The wrist shaft housing lower surface 54 forms a bearing surface 56 at the point of contact between the wrist shaft housing lower surface 54 and the clutch assembly housing upper surface 74. The opening in the wrist shaft housing lower surface 54 and the opening in the clutch assembly housing upper surface 74 are aligned to allow the wrist shaft body 41 to pass from the wrist shaft housing 50 into the clutch assembly housing 70. The wrist shaft lower end 44 and wrist shaft detent 46 are disposed within the clutch assembly housing 70. The clutch pin housing 64 extends from the clutch assembly housing back plate 72 towards the wrist shaft lower end 44. The clutch pin housing 64 contains the clutch pin spring 66 and the clutch pin 62. The clutch pin spring 66 biases the clutch pin 62 against the wrist shaft lower end 44. As shown in FIG. 1, when properly aligned, the clutch pin end 68 is disposed within the wrist shaft detent 46. The clutch assembly housing 70 is attached to a frame 90. The frame 90 has a frame upper surface 92, a frame lower surface 94, a frame back end 96, and a frame front end 98. A carriage assembly motor assembly 100 is disposed below the frame back end 96. The carriage assembly motor assembly 100 includes a carriage assembly rotation motor 102, a carriage assembly axle 104 extending from the carriage assembly motor 102, a carriage assembly axle coupling 106 and a carriage assembly motor assembly housing 110. The carriage assembly rotation motor 102 is any common motor which can provide a rotational force to the carriage assembly axle 104. The carriage assembly motor assembly housing 110 is watertight and has a carriage assembly motor assembly housing back end 112, a carriage assembly motor assembly housing front end 114, a carriage assembly motor assembly housing coupling 116 and a carriage assembly motor assembly housing seal 118. The carriage assembly motor assembly housing back end 112 has an opening therethrough (not shown). The carriage assembly motor assembly housing coupling 116 is sealably connected to the carriage assembly motor assembly housing back end 112 about the medial hole. The carriage assembly motor assembly housing coupling 116 is coupleable to a power source and a positive pressure air tube. The carriage assembly axle 104 passes through an opening in the carriage assembly motor assembly housing front end 114. The carriage assembly motor assembly housing 110 is sealed from the external environment by the carriage assembly motor assembly housing seal 118, which is disposed annularly around the carriage assembly axle 104. The carriage assembly axle 104 terminates in a carriage assembly axle coupling 106. Attached to, or integral to, the frame front end 98 is a probe carriage housing 150 which is a hollow cylindrical structure forming a probe carriage housing cavity 157. The internal surface of the probe carriage housing cavity 157 provides a bearing surface 152. The probe carriage housing further includes a back end 154 and a front end 155, both having openings therethrough connected to the probe carriage housing cavity. A probe carriage housing stop pin 156 is located adjacent to the probe carriage back end 154, and a probe carriage housing retainer 158 is located adjacent to the probe carriage front end 155. The probe carriage assembly 130 is disposed within the probe carriage housing cavity 157 contacting the probe carriage housing bearing surface 152. The probe carriage assembly 130 is a hollow cylindrical body forming a probe carriage cavity 138 and having a coupling arm 131 extending therefrom. The probe carriage assembly coupling arm 131 terminates in a probe carriage coupling end 132 which is attached to the carriage assembly axle coupling 106 by a probe carriage coupling pin 134. The probe carriage assembly coupling arm 131 is offset from the center of the probe carriage assembly, so as to provide a probe carriage cord access notch 136. Along the inner surface 133 of probe carriage assembly 130 is a longitudinal probe carriage slot 142 having a length and a width. The probe carriage spring 140 and the probe 160 are disposed within the probe carriage cavity 138. The probe carriage spring 140 biases the probe 160 (described below) towards the front end of the probe carriage housing 155. The probe carriage assembly 130 is retained within the probe carriage housing 150 by the probe carriage housing retainer 158. The probe carriage assembly has at least one stop pin 144 extending from the probe carriage assembly back end 148. The probe 160 includes a cylindrical probe body 161 having a front portion 162 and a back portion 163. The front portion 162 terminates in a circular mating surface 164. The mating surface 164 has a groove 166 thereon, which bisects the circular mating surface 164. The probe body back portion 163 is enclosed within an annular ring 168. The annular ring 168 has at least one external projection 170. When the probe 160 is disposed within the probe carriage assembly 130, the annular ring projection 170 is loosely fitted within the probe carriage slot 142. The probe carriage slot 142 and projection 170 limit the horizontal movement of the probe 160 within the probe carriage assembly 130, including limiting range of motion provided by the probe carriage spring 140. The projection 170, which is preferably circular, has a diameter that is smaller than the length and the width of the slot 142. Thus, the probe 160 has a limited range, preferably about 2 degrees, of yaw, pitch, and roll within the probe carriage assembly 130. The probe carriage back end 172 has a cylindrical disk 174 attached thereto. The cylindrical disk 174 provides a mating surface for the probe carriage spring 140. When the probe 160 is disposed within the probe carriage assembly 130, the probe carriage spring 140 biases the probe 160 towards the front end of the probe carriage housing. As shown in FIGS. 1 and 4, in the preferred embodiment, a camera assembly 180 is attached to the frame upper surface 92 proximal to the frame front end 98. The camera assembly 180 includes a camera mounting bracket 184, a camera 182, a camera light mounting bracket 188, and a camera light 186. The camera mounting bracket 184 is attached to the frame 90 and provides support for the camera 182. The camera light mounting bracket 188, is attached to the camera 182 and provides support for the camera light 186. The camera 182 and the camera light 186 are oriented to point towards the probe 160. The camera 182 is connected by a cable (not shown) to a video display. The camera light 186 is connected by a power cord (not shown) to a power source. As shown in FIG. 5, a remotely operated service arm 200 (xe2x80x9cROSAxe2x80x9d or xe2x80x9crobot armxe2x80x9d) is disposed above a cylindrical pressure vessel 190 for a nuclear reactor 190. As shown on FIGS. 5 and 6, within the cylindrical pressure vessel 190 is disposed a baffle 192 constructed of a plurality of flat plates forming a grid-like pattern. The baffle plates are secured to each other by baffle bolts along the outer periphery of the grid. As shown in FIG. 7, the baffle bolts 300 installed in recesses 304 on the baffle plates and locked into place by a lock bar 302 which is welded to the fastener head and the baffle 192. The robot arm 200 is designed to position a variety of end effectors within the pressure vessel 190. As indicated by arrow X on FIG. 5, the robot arm moves toward or away from the pressure vessel wall 199 and, as indicated by arrow Y, the arm can rotate 360xc2x0 about its axis, thus providing access to any point along the circumference of the pressure vessel. The robot arm is further designed to allow its end to travel vertically as indicated by arrow Z, substantially the entire length of the pressure vessel cylinder 190. Because the baffle bolts may be oriented perpendicular to the pressure vessel cylinder wall, the robot arm 200 must provide an additional degree of freedom in order to access the baffle bolt head. The ultrasonic testing end effector 10, and more specifically the wrist assembly 20, provides the additional degree of freedom required to inspect the baffle bolt heads. The ultrasonic testing end effector 10 is mounted on the robot arm 200, so that the wrist axle 24 rotates in a plane perpendicular to the robot arm""s vertical plane of travel. In operation, resolvers 194, 195, 196 on the robot arm 200 track the position of the end of the robot arm 200 and provide feedback to a digital converter 212. Using data from the resolvers and a computer controlled positioning system 214, the operator positions the end effector adjacent to the baffle bolt to be tested. Once in position, the camera 182 will display the baffle bolt on the operator""s monitor 216. The operator then uses controls 218, 220, such as a joy-stick, to manually adjust the position of the mating surface of the probe 164 to be grossly aligned, within xc2x12 degrees of being perpendicular, with the baffle bolt. The operator then adjusts the orientation of the mounting surface groove 166 to be grossly aligned, within xc2x12 degrees, with the baffle bolt lock bar. The mating surface 164 is then moved into contact with the baffle bolt. The mating surface 169 is biased against the baffle bolt by the force of the probe carriage spring 140. In the event of a slight misalignment between the baffle bolt and the mating surface 164, the probe 160 will correct its alignment to be flush due to the play provided by the loose fit between the probe carriage slot 142 and the projection 170. Once the mating surface 164 is flush with the baffle bolt, an ultrasonic test may be performed. Once the test is complete, the end effector 10 can be moved to another baffle bolt. It is preferred to move the end effector either vertically or horizontally rather than diagonally. Because the baffle plates 192 run vertically, it is less likely that the end effector will catch on an edge while moving vertically. Conversely, given the proximity of the vessel wall to the sawtooth edge formed by the baffle plates, it is possible that the end effector will contact the baffle 192 during horizontal movement. Damage to the end effector 10 and the baffle 192 is prevented by the clutch assembly 60. In operation, the rotation of the wrist shaft 40 is translated to the clutch assembly 60 which is fixed to the frame member 90. The contact point between the wrist shaft 40 and the clutch assembly 60 is the clutch pin end 68 which is disposed within the wrist shaft detent 46. When the probe 160, carriage assembly 130 or the frame front end 98, contacts an immovable surface, such as a baffle plate 192, while the end effector 10 is being moved horizontally between baffle bolts, the clutch pin assembly 60, and all assemblies attached thereto will stop moving while the robot arm 200 and wrist motor assembly 20 continue to move. This action forces the clutch pin 62 to slide out of the wrist shaft detent 46 and contact the wrist shaft lower end 44. Because the wrist shaft lower end is cylindrical, there is little friction and the clutch pin assembly 60, and all assemblies attached thereto, will rotate freely about the wrist shaft lower end 44. Both the wrist motor assembly housing 30 and the carriage assembly motor assembly housing 110 have couplings 32, 116 that provide access for pressurized air tubes. In operation the pressurized air tube create a positive pressure within the wrist motor assembly housing 30 and probe motor assembly housing 110 so that any seal leakage results in air escaping from the wrist motor assembly housing 30 or carriage assembly motor assembly housing 110 rather than water infiltration into the housing 30, 110. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular arrangements disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the claims appended and any and all equivalents thereof. |
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046541888 | summary | BACKGROUND OF THE INVENTION The present invention is directed to a reactor shroud shield and method for using it, and more particularly to a gamma-ray attenuating shield which can be installed to protect technicians working in the region of a reactor head. A nuclear reactor typically includes a reactor vessel which is disposed in a concrete well located in the floor of a reactor room. A reactor head is bolted to the reactor vessel, which contains water and nuclear fuel housed in an array of fuel assemblies. During operation the nuclear fuel heats the water, which is circulated by pipes from the reactor vessel to steam generators and then back to the reactor vessel for reheating. In order to regulate the degree of water heating and control the nuclear reaction, control rods of neutron absorbing material are movably mounted within the reactor vessel. Control rod drive mechanisms mounted on the reactor head are used to adjust the control rods, with the control rod drive mechanisms and various control and sensor devices being housed within a shroud extending upward from the reactor head. It will be understood that the term "shroud" refers to the housing for the control rod drive mechanisms, although this housing may occasionally be known by other terms, such as lifting skirt. Control rod drive mechanisms are disclosed, for example, in U.S. Pat. No. 3,766,006. During refueling, the nuclear reaction within the vessel is quenched and technicians wearing protective clothing unbolt the reactor head by removing nuts from studs so that the head can be removed from the reactor vessel. These nuts are large and heavy, and a stud tensioner hoist mounted on a circumferential track is available to assist the technicians in handling the nuts and tools. After the reactor head has been bolted, the reactor room is flooded with water to provide shielding and the reactor head, control rod drive mechanisms, shroud, etc., are hoisted by lifting rods attached to the head to expose the depleted fuel assemblies. The depleted fuel assemblies are removed, under water, to a pool of borated water for short-term storage. The reactor is then refueled and the head assembly, complete with control rod drive mechanisms, etc., is set in place on top the reactor vessel. After the water is drained from the reactor room technicians wearing protective clothing re-enter the reactor room to securely bolt the reactor head to the reactor vessel so that operation can continue. The reactor head itself is a hemispherical body typically made of steel, perhaps 18 cm thick, and substantially attenuates gamma-rays from inside the reactor vessel. However the control rod drive mechanisms, during the course of time, tend to accumulate a residue of fission products. Since the control rod drive mechanisms are shielded by the shroud alone, and not by the reactor head, the technicians in the reactor room during the refueling operation are exposed to radiation from the control rod drive mechanisms. While this exposure is a modest one, the steadily increasing safety standards in the nuclear industry have made it desirable to provide additional shielding to the technicians during the refueling operation. U.S. Pat. No. 4,432,932 provides additional shielding by suspending quilted blankets of lead wool from short mounting arms fixedly attached to the shroud. The technicians install the blankets before undertaking the longer task of bolting or unbolting the reactor head. However, such blankets are unwieldy, and in practice may weigh substantially more than the technicians who install them. Even if the stud tensioner hoist is used to manipulate the blankets as they are being hung or removed, the blankets must still be muscled laterally by the distance between the stud tensioner hoist and the mounting arms on the shroud. This not only increases fatigue, it slows the blanket installation process and consequently increases the exposure to radiation. Moreover the heavy weight and awkward movements that are necessary increase the possibility of slips or other accidents. SUMMARY OF THE INVENTION The primary object of the present invention is to provide a reactor shroud shield which permits shielding members to be installed around the reactor shroud or removed therefrom more expeditiously than in the prior art. Another object of the present invention is to provide a reactor shroud shield having pivotable mounting arms which can swing toward the stud tensioner hoist to facilitate transfer of the shielding members. Another object of the present invention is to provide a reactor shroud shield employing shielding members in the form of overlapping panels which encircle the shroud. These and other objects can be attained by attaching a permanent rail to the shroud or to lifting rods extending upward from the reactor head. The permanent rail encircles the shroud. A plurality of swingout rails are pivotably mounted to the permanent rail, so that they can be swung toward the stud tensioner hoist. One or more hangers are attached to each swingout rail to receive shielding members, which may be in the form of lead panels having suspension members pivotably attached thereto. Each suspension member has an eye element for engagement by the stud tensioner hoist and hook elements for engagement by the hangers attached to the swingout arms. When fully installed, the lead panels encircle the lower portion of the shroud and overlap each other so as to avoid unshielded "seams" between the panels. |
abstract | A fuel support for a nuclear reactor may include: a plurality of fuel support apertures, each fuel support aperture dimensioned for receiving a lower tie plate of a fuel assembly; and a plurality of lumens, each lumen being coupled to a different fuel support aperture. At least one lumen may be configured for attenuating a fluid flow differently than the fluid flow in at least one other of the lumen. |
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claims | 1. A mass spectrometer arrangement having a detection system (12) and comprising:a cathode configuration (6) for emitting electrons (21);a reaction zone (3) having an entrance opening (14) for a supply of neutral particles (20), the reaction zone being operatively connected to the cathode configuration (6) for ionization of the neutral particles (20) in an effective region of the reaction zone to form ions (22);an ion extraction system (4) communicating with the effective region of the reaction zone (3);guidance means (1, 10, 11) for guidance of the ions (22) to the detection system (12) within the mass spectrometer arrangement;evacuation means for evacuation of the mass spectrometer arrangement;the cathode configuration (6) comprising a field emission cathode with an emitter surface (7) and, at a short distance from the emitter surface (7), an extraction grid (9) for extraction of electrons (21) away from the emitter surface, the extraction grid substantially covering the emitter surface (7), andthe emitter surface (7) at least partly encompassing a hollow volume (13) to create a tubular structure around the hollow volume (13),wherein the emitter surface (7) is a surface which has been subjected to etching to thereby form a rough surface. 2. The arrangement as claimed in claim 1, wherein the size of the emitter surface (7) is in the range of 0.5 cm2 to 80 cm2. 3. The arrangement as claimed in claim 1, wherein the emitter surface (7) is substantially shaped as a concave inside-facing surface of longitudinal side walls of a cylinder, and wherein the extraction grid is shaped as a second cylinder radially inside the emitter surface cylinder. 4. The arrangement as claimed in claim 1, wherein the emitter surface (7) is a thin layer deposited on a housing wall (2) formed by one of CVD and PVD. 5. The arrangement as claimed in claim 1, wherein the emitter surface (7) is a roughened surface which has been subjected to etching to form a multiplicity of irregularly distributed prominences. 6. The arrangement as claimed in claim 1, wherein the extraction grid (9) is positioned opposite the emitter surface (7) with insulating spacers (8). 7. The arrangement as claimed in claim 1, wherein the extraction grid (9) is biased with respect to the emitter surface (7) with a positive voltage (VG) and that this voltage is in the range from 70 V to 2000 V. 8. The arrangement as claimed in claim 1, wherein the detector system (12) includes a rod system which is part of a quadrupole mass spectrometer. 9. A mass spectrometer arrangement having a detection system (12) and comprising:a cathode configuration (6) for emitting electrons (21);a reaction zone (3) having an entrance opening (14) for a supply of neutral particles (20), the reaction zone being operatively connected to the cathode configuration (6) for ionization of the neutral particles (20) in an effective region of the reaction zone to form ions (22);an ion extraction system (4) communicating with the effective region of the reaction zone (3);guidance means (1, 10, 11) for guidance of the ions (22) to the detection system (12) within the mass spectrometer arrangement;evacuation means for evacuation of the mass spectrometer arrangement;the cathode configuration (6) comprising a field emission cathode with an emitter surface (7) and, at a short distance from the emitter surface (7), an extraction grid (9) for extraction of electrons (21) away from the emitter surface, the extraction grid substantially covering the emitter surface (7), andthe emitter surface (7) consisting essentially of generally planar surface, the emitter surface (7) being curved and at least partly encompassing a hollow volume (13) to create a hollow tubular structure,wherein, adjoining the hollow volume (13) of the cathode configuration (6) is an electron extraction lens (5) and including, in an axial direction of the mass spectrometer arrangement, an ion extraction lens (4), the reaction zone (3) being located between the electron extraction lens (5) and the ion extraction lens (4) to form a volume and the entrance opening (14) for the neutral particles (20) being disposed peripherally upon said volume of the reaction zone (3). 10. The arrangement as claimed in claim 9, wherein the size of the emitter surface (7) is in the range of 0.5 cm2 to 80 cm2. 11. The arrangement as claimed in claim 9, wherein the size of the emitter surface (7) is in the range of 1.0 cm2 to 50 cm2. 12. The arrangement as claimed in claim 9, wherein the emitter surface (7) forms at least arcuate sector elements that are not divided and forms a closed tubular emitter surface (7) wherein the emitter surfaces are concave and face inwards. 13. The arrangement as claimed in claim 9, wherein the emitter surface (7) is substantially shaped as a concave inside-facing surface of longitudinal side walls of a cylinder, and wherein the extraction grid is shaped as a second cylinder radially inside the emitter surface cylinder. 14. The arrangement as claimed in claim 9, wherein the diameter of the hollow volume (13) is between 0.5 cm and 8.0 cm and its length in the axial direction is between 2.0 em and 8.0 cm. 15. The arrangement as claimed in claim 9, wherein the diameter of the hollow volume (13) is between 0.5 cm and 6.0 cm and its length in the axial direction is between 2.0 cm and 8.0 cm. 16. The arrangement as claimed in claim 9, wherein the emitter surface (7) comprises at least on the surface a layer comprising at least one of the materials selected from the group consisting of: carbon; a metal; a metal mixture; a semiconductor; a carbide; and mixtures thereof. 17. The arrangement as claimed in claim 16, wherein the emitter surface (7) is substantially comprised of at least one of molybdenum, tantalum and corrosion-resistant steel. 18. The arrangement as claimed in claim 16, wherein the emitter surface (7) is a thin layer deposited on a housing wall (2) formed by one of CVD and PVD. 19. The arrangement as claimed in claim 9, wherein the emitter surface (7) is comprised of at least a portion of the surface of one housing wall (2), wherein the housing wall (2) is comprised of one of: metal, metal alloy, and corrosion resistant steel. 20. The arrangement as claimed in claim 9, wherein the emitter surface (7) is a roughened surface which has been subjected to etching to form a multiplicity of irregularly distributed prominences. 21. The arrangement as claimed in claim 9, wherein the emitter surface (7) is a roughened surface that is roughened by one of: mechanically roughened; plasma etching; and chemical etching. 22. The arrangement as claimed in claim 9, wherein the distance between the extraction grid (9) and the emitter surface (7) is in the range from 1.0 μm and 2 mm. 23. The arrangement as claimed in claim 9, wherein the distance between the extraction grid (9) and the emitter surface (7) is in the range from 5.0 μm and 200 μm. 24. The arrangement as claimed in claim 9, wherein the extraction grid (9) has a grid structure with high transmission factor and is made of wire cloth. 25. The arrangement as claimed in claim 9, wherein the extraction grid (9) is positioned opposite the emitter surface (7) with insulating spacers (8). 26. The arrangement as claimed in claim 9, wherein the extraction grid (9) is biased with respect to the emitter surface (7) with a positive voltage (VG) and that this voltage is in the range from 70 V to 2000 V. 27. The arrangement as claimed in claim 9, wherein the extraction grid (9) is biased with respect to the emitter surface (7) with a positive voltage (VG) and that this voltage is in the range from 70 V to 200 V. 28. The arrangement as claimed in claim 9, wherein the detector system (12) includes a rod system which is part of a quadrupole mass spectrometer. 29. The arrangement as claimed in claim 9, wherein the emitter surface comprises roughened generally planar surface which has been roughened by mechanical grinding. 30. A mass spectrometer arrangement having a detection system (12) and comprising:a cathode configuration (6) for emitting electrons (21);a reaction zone (3) having an entrance opening (14) for a supply of neutral particles (20), the reaction zone being operatively connected to the cathode configuration (6) for ionization of the neutral particles (20) in an effective region of the reaction zone to form ions (22);an ion extraction system (4) communicating with the effective region of the reaction zone (3);guidance means (1, 10, 11) for guidance of the ions (22) to the detection system (12) within the mass spectrometer arrangement;evacuation means for evacuation of the mass spectrometer arrangement;the cathode configuration (6) comprising a field emission cathode with an emitter surface (7) and, at a short distance from the emitter surface (7), an extraction grid (9) for extraction of electrons (21) away from the emitter surface, the extraction grid substantially covering the emitter surface (7), andthe emitter surface (7) consisting essentially of generally planar surface, the emitter surface (7) being curved and at least partly encompassing a hollow volume (13) to create a hollow tubular structure,wherein the reaction zone (3) is formed within the hollow volume (13) of the cathode configuration (6) so that the hollow volume (13) is delimited on one side by an ion extraction lens (4) and on an opposite side is located the entrance opening (14) for the neutral particles (20). 31. The arrangement as claimed in claim 30, wherein the size of the emitter surface (7) is in the range of 0.5 cm2 to 80 cm2. 32. The arrangement as claimed in claim 30, wherein the emitter surface (7) is substantially shaped as a concave inside-facing surface of longitudinal side walls of a cylinder, and wherein the extraction grid is shaped as a second cylinder radially inside the emitter surface cylinder. 33. The arrangement as claimed in claim 30, wherein the emitter surface (7) is a thin layer deposited on a housing wall (2) formed by one of CVD and PVD. 34. The arrangement as claimed in claim 30, wherein the emitter surface (7) is a roughened surface which has been subjected to etching to form a multiplicity of irregularly distributed prominences. 35. The arrangement as claimed in claim 30, wherein the extraction grid (9) is positioned opposite the emitter surface (7) with insulating spacers (8). 36. The arrangement as claimed in claim 30, wherein the extraction grid (9) is biased with respect to the emitter surface (7) with a positive voltage (VG) and that this voltage is in the range from 70 V to 2000 V. 37. The arrangement as claimed in claim 30, wherein the detector system (12) includes a rod system which is part of a quadrupole mass spectrometer. 38. The arrangement as claimed in claim 30, wherein the emitter surface comprises roughened generally planar surface which has been roughened by mechanical grinding. 39. A mass spectrometer arrangement having a detection system (12) and comprising:a cathode configuration (6) for emitting electrons (21);a reaction zone (3) having an entrance opening (14) for a supply of neutral particles (20), the reaction zone being operatively connected to the cathode configuration (6) for ionization of the neutral particles (20) in an effective region of the reaction zone to form ions (22);an ion extraction system (4) communicating with the effective region of the reaction zone (3);guidance means (1, 10, 11) for guidance of the ions (22) to the detection system (12) within the mass spectrometer arrangement;evacuation means for evacuation of the mass spectrometer arrangement;the cathode configuration (6) comprising a field emission cathode with an emitter surface (7) and, at a short distance from the emitter surface (7), an extraction grid (9) for extraction of electrons (21) away from the emitter surface, the extraction grid substantially covering the emitter surface (7), andthe emitter surface (7) consisting essentially of generally planar but also rough surface, the emitter surface (7) being curved and at least partly encompassing a hollow volume (13) to create a hollow tubular structure,wherein the reaction zone (3) is located on a longitudinal axis of the mass spectrometer arrangement and is encompassed by a wall which includes, in a radial direction toward the axis, an extraction opening which forms the electron extraction lens (5), and the extraction opening communicating with the hollow volume (13) of the cathode configuration (6), the cathode configuration (6) being positioned orthogonally with respect to the axis and to the reaction zone (3) for a radial feeding of the electrons into the reaction zone (3), and in the wall at least one entrance opening (14) is provided for the introduction of neutral particles (20). 40. The arrangement as claimed in claim 39, wherein the size of the emitter surface (7) is in the range of 0.5 cm2 to 80 cm2. 41. The arrangement as claimed in claim 39, wherein the emitter surface (7) is substantially shaped as a concave inside-facing surface of longitudinal side walls of a cylinder, and wherein the extraction grid is shaped as a second cylinder radially inside the emitter surface cylinder. 42. The arrangement as claimed in claim 39, wherein the emitter surface (7) is a thin layer deposited on a housing wall (2) formed by one of CVD and PVD. 43. The arrangement as claimed in claim 39, wherein the emitter surface (7) is a roughened surface which has been subjected to etching to form a multiplicity of irregularly distributed prominences. 44. The arrangement as claimed in claim 39, wherein the extraction grid (9) is positioned opposite the emitter surface (7) with insulating spacers (8). 45. The arrangement as claimed in claim 39, wherein the extraction grid (9) is biased with respect to the emitter surface (7) with a positive voltage (VG) and that this voltage is in the range from 70 V to 2000 V. 46. The arrangement as claimed in claim 39, wherein the detector system (12) includes a rod system which is part of a quadrupole mass spectrometer. 47. The arrangement as claimed in claim 39, wherein the emitter surface comprises roughened generally planar surface which has been roughened by mechanical grinding. |
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claims | 1. A fluorescent material having a composition formula represented as (Gd1-α-β-γLαCeβTbγ)3+a(Al1-u-vGauScv)5-bO12 (L is at least one element selected from Y and Lu), where a, b, α, β, γ, u and v satisfy the following ranges:wherein a, b, α, β, γ, u and v satisfy the following ranges:0<a≦0.07,0<b≦0.07,0<α≦0.15,0.0003≦β≦0.004,0.03≦γ≦0.15,0.35≦u≦0.70, and0≦v≦0.02. 2. The fluorescent material of claim 1, wherein β satisfies the following range:0.0005≦β≦0.003. 3. The fluorescent material of claim 1, wherein v satisfies the following range:0.003≦v≦0.02. 4. The fluorescent material of claim 1, wherein a=b. 5. The fluorescent material of claim 1, wherein the fluorescent material is a polycrystalline material, and an average crystal grain size of the sintered material is 0.05 μm or more and 5 μm or less. 6. The fluorescent material of claim 1, wherein the fluorescent material is in the form of a plurality of fibers, the average fiber diameter is 1 μm or more and 50 μm or less. 7. A scintillator including fluorescent material of claim 6 and having a plate shape, wherein the fluorescent materials are disposed in parallel to one another so that a longitudinal direction of a fibrous shape of the fluorescent materials is along a thickness direction of the plate shape. 8. A radiation image conversion panel comprising:a scintillator of claim 7; anda photoelectric converter for converting light into an electrical signal. |
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claims | 1. A microminiature thermionic converter comprising: a first electrode comprising a first material having a first work function; a second electrode comprising a second material having a second work function different from the first work function; at least one dielectric spacer deposited using chemical vapor deposition, supporting the second electrode relative to the first electrode such that the second electrode, at its closest approach to the first electrode is separated from the first electrode by a distance ranging from between about 1 micron and about 10 microns thereby defining an interelectrode gap, wherein aggregate cross sectional area associated with the at least one dielectric spacer is sufficiently low that in operation the ratio of watts of thermal conversion of the microminiature thermionic converter to watts of thermal conductivity losses, including losses resulting from flow of thermal energy between the first and second a electrodes via the at least one dielectric spacer, is greater than about 0.15. 2. The microminiature thermionic converter of claim 1 wherein the at least one dielectric spacer comprises material selected from the group consisting of SiO 2 and Si 3 N 4 . claim 1 3. The microminiature thermionic converter of claim 2 wherein the first material is a first oxide material. claim 2 4. The microminiature thermionic converter of claim 3 wherein the second material is a second oxide different from the first oxide material. claim 3 5. The microminiature thermionic converter of claim 4 wherein the first oxide material is selected from the group consisting of BaO, SrO, CaO, Sc 2 O 3 , and a mixture of BaSrCaO, Sc 2 O 3 and metal, and any combinations thereof. claim 4 6. The microminiature thermionic converter of claim 1 wherein the at least one dielectric spacer is disposed between the first electrode and the second electrode. claim 1 7. The microminiature thermionic converter of claim 1 wherein the at least one dielectric spacer is disposed in a position other than between the first electrode and the second electrode. claim 1 8. The microminiature thermionic converter of claim 7 wherein the dielectric spacer comprises two separate elements with the interelectrode gap therebetween. claim 7 9. A microminiature thermionic converter made by a process comprising the steps of: depositing a first electrode layer comprising a first material selected from the group consisting of BaO, SrO, CaO, Sc 2 O 3 , other oxides, and a mixture of BaSrCaO, Sc 2 O 3 and metal, and any combinations thereof, and having a first work function; depositing a dielectric oxide spacer layer; depositing a second electrode layer comprising a second material selected from the group consisting of BaO, SrO, CaO, Sc 2 O 3 , other oxides, and a mixture of BaSrCaO, Sc 2 O 3 and metal; and any combinations thereof having a second work function that is different from the first work function; and removing matter from the dielectric oxide spacer layer thereby forming an interelectrode gap. 10. The microminiature thermionic converter of claim 9 wherein the dielectric oxide spacer layer comprises material selected from the group consisting of SiO 2 and Si 3 N 4 and combinations thereof. claim 9 11. The microminiature thermionic converter of claim 10 wherein the step of removing matter from the dielectric oxide spacer layer comprises a technique selected from the group consisting of claim 10 steps comprising masking at least part of the first electrode layer, masking at least part of the second electrode layer, masking at least two parts of the spacer layer, and etching out an interelectrode gap bound on opposite sides by unetched portions of the spacer layer; steps comprising sputtering particles to disrupt crystal structure in a part of the spacer layer thereby causing the crystal structure to disintegrate in that part of the spacer layer and leave an interelectrode gap; and steps comprising utilizing etching vias cut into at least one of the electrode layers to permit etchant to enter the spacer layer and remove a portion of the spacer layer between the first and second electrode layers, leaving an interelectrode gap. |
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abstract | A method of removing the upper internals assembly from a nuclear reactor pressure vessel for refueling that simultaneously disconnects two or more of the in-core instrument thimble assemblies from the reactor vessel penetrations through which their signal cables extend. The signal cables are connected to the penetrations with an electrical connector that supports the two or more in-core instrumentation thimble assembly signal leads. Before the electrical connector is disconnected, water in the vessel is lowered below the connection so that the process is performed in a dry environment. |
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abstract | A technique is disclosed for electro-optically inducing a force to fabricated samples and/or devices with laser light. The technique uses the interaction of the oscillating electric field of the laser beam in opposition with the electric field produced by an appropriate electric charge carrier to achieve a net repulsive (or attractive) force on the component holding the electric charge. In one embodiment, force is achieved when the field near the charge carrier is modulated at a subharmonic of the electric field oscillation frequency of the laser and the relative phases of the light field and electric charge carrier field are controlled to provide optimal repulsion/attraction. The effect is scalable by applying the technique to an array of charge carrier fields sequentially as well as using higher power lasers and higher carrier field voltages. |
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048448567 | abstract | A process for automatic regulation of the soluble boron content of the cooling water of a pressurized water nuclear reactor in which operating regions (2, 3) of the means of boration and of the means of dilution respectively are determined a priori, corresponding to pairs of values of two control parameters relating to the position of the reactor regulating group in the core and to the deviation of the axial power imbalance relative to the reference axial imbalance. The momentary value of the control parameters is determined continuously during the operation of the reactor, and the stopping or the triggering of the means for boration or for dilution is commanded when the operating point crosses a boundary of an operating region (2, 3).. The invention applies in particular to a nuclear reactor controlled in G mode. |
abstract | An anti-rotation device for preventing separation of a valve stem and valve disc(s) in a gate valve in a nuclear reactor power plant of the type where connection means, such as a wedge or disc hub, is threadedly engaged by the valve stem to actuate the valve and a method for maintenance or repair of the valve without disassembly of the valve or removal of the valve actuator. |
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summary | ||
abstract | Multi-leaf collimators have a guide frame (1) with a plurality of metal plates (3) arranged in a displaceable fashion, by which each individual metal plate can be displaced by an electric motor (M), with the electric motor (M) being a rotary electromechanical motor (M), which operates according to the form-fit principle, with electromechanical actuators. |
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claims | 1. A system for storing exothermic materials, comprising: a first canister adapted to store exothermic materials therein, comprising: a first bottom wall; a side wall extending from said first bottom wall such that said first bottom wall and said side wall form a first hollow structure having a first opening opposite said first bottom wall and defining a first storage volume that exothermic material can be inserted into and supported within; and at least one alignment channel affixed to and spanning at least partially along said side wall between said first bottom wall and said first opening at the exterior of said first hollow structure; and a second canister adapted to store exothermic materials therein, comprising: a bottom support structure adapted to support said first canister when said first bottom wall of said first canister engages said bottom support structure and further adapted to provide at least one bottom opening through which a cooling medium is flowable; an inner wall extending from said bottom support structure, said bottom support structure and said inner wall forming an inner hollow structure adapted to receive at least a portion of said first canister such that at least one open channel is defined between said inner wall of said inner hollow structure and said side wall of said first canister, said inner hollow structure having an inner opening opposite said bottom support structure; a plurality of spacer columns affixed to and spanning at least partially along said inner wall between said bottom support structure and said inner opening at the interior of said inner hollow structure, said spacer columns being adapted to establish said at least one open channel; an outer wall surrounding and spaced from said inner wall; and a second bottom wall extending between said outer wall and said inner wall such that said outer wall, said second bottom wall, and said inner wall form a second hollow structure having a second opening opposite said second bottom wall and defining a second storage volume that exothermic material can be inserted into and supported within, said first canister being maintainable in alignment with respect to said second canister by said alignment channel engaging at least one of said spacer columns. 2. The storage system of claim 1 , wherein said side wall of said first canister and said inner wall of said second canister form a cooling medium flow channel therebetween when said first canister is received within said inner hollow structure of said second canister such that a cooling medium is flowable through said cooling medium flow channel from beneath said bottom support structure through said at least one bottom opening and through said at least one open channel, whereby at least a portion of heat transferred to said side wall of said first canister and said inner wall of said second canister from the exothermic material stored within said first canister is dissipated by the cooling medium flowing through said cooling medium flow channel. claim 1 3. The storage system of claim 1 , wherein said side wall of said first canister and said inner wall of said second canister form a cooling medium flow channel therebetween when said first canister is received within said inner hollow structure of said second canister such that a cooling medium is flowable through said cooling medium flow channel from beneath said bottom support structure through said at least one bottom opening and through said at least one open channel, whereby at least a portion of heat transferred to said inner wall of said second canister and said side wall of said first canister from the exothermic material stored within said second canister is dissipated by the cooling medium flowing through said cooling medium flow channel. claim 1 4. The storage system of claim 1 , wherein said first canister is cylindrically shaped. claim 1 5. The storage system of claim 1 , wherein said outer wall and said inner wall of said second canister are cylindrically shaped such that said second storage volume is annularly shaped. claim 1 6. The storage system of claim 1 , further comprising: claim 1 an overpack defining an overpack interior, said overpack interior being configured to receive said first and second canisters therein such that said overpack encases said first and second canisters. 7. The storage system of claim 1 , further comprising: claim 1 an exothermic material inserted within said first storage volume of said first canister. 8. The storage system of claim 1 , further comprising: claim 1 an exothermic material inserted within said second storage volume of said second canister. 9. The storage system of claim 2 , wherein said cooling medium flow channel is formed, at least in part, by said spacer columns engaging between said side wall of said first canister and said inner wall of said second canister, each of said spacer columns being adapted to maintain a spaced configuration of a portion of said side wall of said first canister and said inner wall of said second canister. claim 2 10. The storage system of claim 6 , wherein said outer wall of said second canister and said overpack are configured to form an outer cooling medium flow channel therebetween when said second canister is received within said overpack interior such that a cooling medium is flowable through said outer cooling medium flow channel, whereby at least a portion of heat transferred to said outer wall of said second canister from the exothermic material stored within said second canister is dissipated by the cooling medium flowing through said outer cooling medium flow channel. claim 6 11. The storage system of claim 7 , wherein said exothermic material is nuclear waste. claim 7 12. The storage system of claim 9 , wherein said spacer columns and said alignment channel engage each other such that rotation of said first canister about a longitudinal axis thereof is prevented. claim 9 13. The storage system of claim 1 , wherein said bottom support structure comprises a plurality of beams arranged to support said first canister when said bottom wall of said first canister engages said beams. claim 1 14. The storage system of claim 8 , wherein said exothermic material is nuclear waste. claim 8 15. The system of claim 1 , wherein: claim 1 said first bottom wall and said side wall are adapted to transfer heat such that heat dissipated from the exothermic material stored within said first storage volume of said first canister is transferred out of said first storage volume through said first bottom wall and said side wall, and said inner wall, said outer wall, and said second bottom wall are adapted to transfer heat such that heat dissipated from the exothermic material stored within said second storage volume of said second canister is transferred out of said second storage volume through said inner wall, said outer wall, and said second bottom wall. |
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description | This application claims the priority benefit of Taiwan application serial no. 101134832, filed on Sep. 21, 2012. The entirety of the above-mentioned patent application is hereby incorporated by reference herein and made a part of this specification. The technical field relates to a material aging apparatus, and more particularly to, a light aging equipment. Solar energy is an inexhaustible and non-polluting energy, and has been a focus of most attentions in terms of solving current problems of pollution and shortage faced by fossil energy. Wherein, a solar cell with its ability of directly converting the solar energy into electricity has become one of the most important options of alternative energies. Generally, such product is used in an outdoor environment for a long time, and a tolerability thereof is often greatly influenced by the environment and climate. For instance, both the solar cell itself or a packaging material thereof, under a condition of operating under the sun light for a long time, a material degradation thereof causing by ultraviolet (UV) light is most likely to happen. Therefore, in order to enhance a service life of the product, and to obtain tolerability parameters of the product within a short amount of time, an accelerated aging test is often performed on the product. The conventional light aging equipments mostly use solar simulator light sources, xenon lamps or UV lamps as light sources for the accelerated aging test, but there are still the following problems in need to be solved; limited by operation modes of the conventional light sources, the product is mostly irradiated by a large area light source, and thus an illuminance on the product in a unit area is also weakened; and herein, in order to enhance a rate of aging, a means to enhance an intensity of the light source is often used, but this may easily result in a temperature increase of the product; namely, the conventional light sources may easily cause the product to be heated, and thereby influencing aging factors of the product. Furthermore, the abovementioned aging light sources are mostly adopted with large scale area aging, and thus partial irradiations on the products and projections of different illuminance light sources on partial regions are not easily. Moreover, the abovementioned light sources all perform aging irradiations on the product with a certain range of wavelengths, and are not able to provide a single wavelength light source or to fine-tune a wavelength range thereof according to the conditions. According to the foregoing, performing a light aging test on a product with precise conditions is still a subject in need of further study. One of exemplary embodiments provides a material aging apparatus using a pulse laser to enhance a light aging efficiency in an object. One of exemplary embodiments provides a material aging apparatus including a pulse laser, a beam expansion assembly and a platform. The platform is configured to carry an object. The pulse laser transmits a first beam to the beam expansion assembly. The beam expansion assembly expands the first beam to a second beam and projects the second beam onto the object. Several exemplary embodiments accompanied with figures are described in detail below to further describe the disclosure in details. FIG. 1 is a schematic diagram illustrating a material aging apparatus according to an embodiment. FIG. 2 is an exploded view illustrating the material aging apparatus of FIG. 1. Referring to FIG. 1 and FIG. 2 at the same time, in the present embodiment, the material aging apparatus 100 is configured to perform a light aging test to an object 200 so as to acquire a material weatherability. The material aging apparatus 100 includes a base 110, a platform 120, a pulse laser 130 and a beam expansion assembly 140. The base 110 is, for example, a granite base, and the platform 120, the pulse laser 130 and the beam expansion assembly 140 are mounted thereon. The platform 120 is, for example, an X axis-Y axis moving platform configured to carry and drive the object 200 to move. The beam expansion assembly 140 is located above the platform 120 and the object 200. A beam emitted by the pulse laser 130 is projected onto the object 200 on the platform 120 through the beam expansion assembly 140. In addition, the material aging apparatus 100 further includes a carrier 150 and a temperature control module 160 disposed on the platform 120, wherein the carrier 150, for example, fixes the object 200 thereon by a means of vacuum suction, and the temperature control module 160 is disposed within the carrier 150 and has a water cooling circuit (not shown) configured to provide a cooling effect to the object 200 on the carrier 150. FIG. 3 is a simplified schematic diagram illustrating the material aging apparatus 100 of FIG. 1. Referring to FIG. 3, the pulse laser 130 is configured to transmit the first beam B1 to the beam expansion assembly 140, and the beam expansion assembly 140 expands the first beam B1 to a second beam B2 and projects the second beam B2 onto a surface of the object 200. Herein, the first beam B1 and the second beam B2 with a wavelength range between 280 nm to 400 nm are projected from the pulse laser 130 of the present embodiment, which is namely a UV light having a significant influence on a material degradation, as an irradiation light source for the light aging. In the present embodiment, the pulse laser 130 is a short-pulse laser, wherein pulse duration thereof is less than 1 μs, and pulse repetition rate thereof is greater than or equal to 10 Hz, so that the pulse laser 130 is characterized as projecting beam with periodic and instantaneous high energy, and the object 200 illuminated by the beam of the pulse laser 130 with periodic and instantaneous highly intensified but low accumulated energy aging effect. For instance, when the object 200 is a solar cell module, taking the pulse laser 130 with energy of 100 mJ, pulse duration of 5 ns and pulse repetition rate of 10 Hz as an example, average power density of the second beam B2, which is projected on the object 200 through the beam expansion assembly 140, can be adjusted between 10 kw/m2 to 0.1 kw/m2 according to different beam expansion areas; however, due to characteristics of the pulse laser 130, instant irruption subjected to the object 200 (converted with 5 ns) is up to 20 Mkw/m2 to 0.2 Mkw/m2, and under the effect of this moderate average power density, the light aging effecting on the solar cell module may be accelerated without burning and melting the module or causing a recrystallization (tempering). Similarly, when the object 200 is a polymer material, the average power density of the second beam B2 is between 5 kw/m2 to 0.1 kw/m2 and the light aging effecting on the object 200 may be accelerated without damaging the polymer material. In addition to providing instantaneous high power density aging characteristics, the pulse laser 130, with its characteristics, is different from the conventional persistent irradiation light source, and thus energy irradiated on the object 200 is not being accumulated; namely, the temperature of the object 200 is not going to be gradually increased due to persistent irradiation, and therefore capable of effectively reducing influences on the light aging test of the object 200 due to thermal effect. In other word, the material aging apparatus 100 of the present embodiment may control the temperature of object 200 merely via the temperature control module 160 disposed within the carrier 150. In the present embodiment, the temperature control module 160 may maintain the temperature of the object 200 on the carrier 150 between 10° C. to 60° C., such that, in addition to avoiding heat accumulation to affect the light aging, it is also capable of effectively avoiding the material from having a recrystallization (annealing) phenomenon and even a damage due to burning and melting. FIG. 4 is a schematic diagram illustrating a moving path of a second beam of FIG. 3 for forming spot on an object. Referring to FIG. 3 and FIG. 4 at the same time, in the present embodiment, by using the platform 120 to actuate the object 200 to move, users are enable to control the moving path of the second beam B2 on the object 200, and further adjust the moving path, power and wavelength range of the second beam B2 according to testing requirements. For instance, the surface of the object 200 is separated into four areas A1 to A4, and enables the second beam B2 to respectively pass through the surface of the object 200 with different paths in areas A1 to A4, so that the power, wavelength and even irradiation area of the second beam B2 are also changed along the paths at the same time. As a result, in the area A1, a spot SP1 formed by the second beam B2 on the object 200 is scanned via a path L1. Similarly, in the areas A2 to A4, spots SP2 to SP4 are also capable of being respectively scanned via paths L2 to L4, wherein the paths L1 to L4 respectively have different density degrees. As a result, the users can perform irradiations on the same object 200 via the spots formed with multiple requirements, and therefore a tolerability parameter of the object 200 against the light aging may be obtained in a more efficient way. Moreover, via the pulse laser 130 with adjustable wavelength, it is also able to produce selective ageing effects to the object 200 by targeting some specific wavelengths, and the users are therefore able to figure out a relative relationship of the material characteristics of the object 200 in correspondence to specific absorption wavelengths. Moreover, in order to successfully achieve the above mentioned effects, the first beam B1 projected from the pulse laser 130 has to be firstly expanded by the beam expansion assembly 140 and then projected onto the object 200 as the second beam B2 to form the required spots. FIG. 5 is a schematic diagram illustrating the beam expansion assembly of FIG. 1. Referring to FIG. 5, the beam expansion assembly 140 is composed of a plurality of optical elements (lens group) 141, 142, so as to expand a spot area of the first beam B1 and shape the spot area to a desired contour; namely, in the present embodiment, the speck area of the second beam B2 is larger than a spot area of the first beam B1 spot area. As illustrated in FIG. 4, the spots SP1 to SP4 with areas larger than 1 cm2 are formed on the object 200 of 20 cm2, and then the object 200 is scanned by the spots SP1 to SP4 via the different paths illustrated in FIG. 4 in order to accelerate the light aging effect on the object 200. In addition, the energy per unit area of the second beam B2, after the beam expansion, is smaller than the energy per unit area of the first beam B1, and therefore the energy subjected to the object 200 is also reduced at the same time in order to prevent the temperature from being too high. Compositions of the beam expansion assembly are not limited herein, any element capable of beam expanding and shaping to the first beam B1 emitted by the pulse laser 130 may be adapted to the present embodiment. FIG. 6 is a schematic diagram illustrating a beam expansion assembly according to another embodiment, wherein the beam expansion assembly 140A of the present embodiment is a Galilean beam expander, which in terms of the structure is different form a Keplerian beam expander illustrated in FIG. 5 but can achieve an effect similar to beam expansion. In addition, FIG. 7 is a schematic diagram illustrating a beam expansion assembly according to still another embodiment, wherein the beam expansion assembly 140B, targeting the two beams respectively, can perform the beam expansion and shaping effects at the same time. Accordingly, the users, targeting the area size and appearance of the object, can further configure the contours and sizes of the required spots via the beam expansion assemblies. In summary, in the embodiments of the disclosure, the material aging apparatus reduces the average illumination per unit area thereof via using the short pulse laser as the light source and adjusting the beam area thereof through the beam expansion assembly, and therefore is capable of using the light source with a lower accumulated energy yet a most intensified instant energy to irradiate the object, thereby effectively improving problems caused by the conventional light-tube type or light-box type light aging equipments. Furthermore, since the pulse laser can provide the light aging irradiation to partial areas of the object, and in coordination with the movable platform, the pulse laser irradiates the object via different powers, different paths and different wavelengths, and completes light aging irradiations of different requirements on the same object. As such, the material aging apparatus is able to figure out the light aging parameters of the object in a more efficient way. It will be apparent to those skilled in the art that various modifications and variations can be made to the structure of the disclosed embodiments without departing from the scope or spirit of the disclosure. In view of the foregoing, it is intended that the disclosure cover modifications and variations of this disclosure provided they fall within the scope of the following claims and their equivalents. |
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description | FIG. 1 shows an apparatus 1 in accordance with the present invention. A support vessel 3 such as a floating drilling rig or other floating vessel is located substantially vertically above a hydrocarbon borehole (not shown) and is typically either dynamically positioned over the particular well by means of conventional thrusters for example or is moored by conventional means over the required well. The support vessel 3 comprises a conventional crane or derrick 5 and is further provided with a make up/breakout unit generally shown at 7. The make up/breakout unit 7 comprises an upper tong 9 and a lower tong 11, which may preferably be a lower rotary back-up 11. The reader is directed towards our co-pending British Patent Application No 0004354.7 and PCT Application No GB 00/04241 for further details of a suitable make up/break out unit 7. In summary, the upper tong 9 provides means to make up and breakout tubing, casing or drillpipe 16 during tripping and snubbing operations, and is hydraulically powered. The upper tong 9 comprises three sliding jaws. (not shown) which virtually encircle the tubular 16 to maximize torque whilst minimizing marking and damage and is provided with a cam operated jaw system (not shown) which can be operated to allow passage of workstring tool joints as well as tubing casing couplings. The upper tong 9 is powered by hydraulic motors (not shown) which provide speeds and torque capable of spinning and making/breaking high torque connections. The lower tong or rotary back-up 11 generally has two functions. During drilling operations, the rotary back-up 11 generates the torque required for high speed milling and drilling. This torque is transferred to the outer diameter of the string 16 by means of three sliding jaws (not shown). During tripping operations, the jaws of the rotary back-up 11 are activated to grip the lower tubular 16 in the make up/breakout unit 7 (i.e. the tubular that is already included in the string 16) and resist the torque generated by the upper tong 9 when making up or breaking out the tubular connections. The make up/breakout unit 7 is mounted on the support vessel 3 by means of tong table compensator 13 which moves the make up/breakout unit 7 with respect to the support vessel 3, such that the make up/breakout unit 7 is substantially stationary with respect to the subsea surface. Thus, the tong table compensator 13 compensates for the action of the wave motion upon the support vessel 3. The heave of the support vessel 3 due to the ocean waves may be in the. region of 10 to 15 feet, and the stroke of the compensator table 13 substantially matches the heave of the support vessel 3, such that the make up/breakout unit is stationary with respect to the subsea equipment 17. A riser 15 extends downwardly from the support vessel 3 wherein the upper end of the riser 15 is located substantially immediately below the tong table compensator 13, although the lower end of the riser 15 could be offset from the upper end of the riser 15 as is known in the art. The riser 15 extends downwardly towards subsea equipment 17 of the apparatus 1. The subsea equipment 17 comprises an outer frame 19 which is suitably provided with sufficient strength bearing members in the form of struts, braces, etc. The subsea equipment 17 comprises a subsea annular seal 21 or lower riser seal 21 which is provided at the upper, in use, end of the frame 19. The lower end of the riser 15 is located within the lower riser seal 21, such that the outer surface of the lower end of the riser 15 is sealed to the sea water by the lower riser seal 21. A set of travelling slips 23 are mounted upon a passive rotary bearing 25. The rotating portion of the passive rotary bearing 25 supports the traveling slips 23, and allows the travelling slips 23 to rotate with respect to the non-rotating portion of the passive rotary bearing 25. The purpose of the travelling slips 23 will be described subsequently. The lower end of the passive rotary bearing 25 is mounted to the upper end of a hydraulic jack system 27, as can be more clearly seen in FIG. 2 and particularly FIG. 3. There are at least a pair of hydraulic jacks 27, and more preferably four hydraulic jacks 27 mounted equi-distantly spaced around a central region through which a tubing string 16 can pass. Each hydraulic jack 27 comprises a piston 29 and cylinder 31 arrangement. The piston 29 comprises a piston head 33 at its lower end, and the outer radial surface of the piston head 33 is sealed with respect to the inner surface of the cylinder 31 by means of a set of xe2x80x9cVxe2x80x9d seal packings 35. The packings 35 shown in FIGS. 4(a) to 4(c) comprise an upper set of three seals arranged to seal in the direction from above the piston head 33, and hence seal against the subsea hydrostatic pressure, and a lower set of three seals arranged to seal in the direction from below the piston head 33 and hence seal against the hydraulic fluid pressure within the cylinder 31; however, it should be noted that more than, or less than, three seals can be provided in the upper and/or lower sets of seals. A second set of xe2x80x9cVxe2x80x9d seal packing 37 acts between the longitudinal body of the piston 29 and the upper end of the cylinder 31, and acts to prevent the hydrostatic pressure of sea water from outside of the cylinder 31 from entering the interior of the cylinder 31. An upper hydraulic port 39 is provided within a side wall of the cylinder 31 toward the upper end thereof, and a lower hydraulic port 41 is provided in the side wall toward the lower end of the cylinder 31, such that the hydraulic ports 39, 41 provide access to the interior of the cylinder 31. A hydraulic fluid control system for providing pressurised hydraulic fluid to the cylinders 31 in a controlled and selective manner is also provided as will now be described. A hydraulic fluid charge pump 43 is provided on the support vessel 3 and is capable of providing/maintaining pressurised hydraulic fluid to an arrangement of hydraulic fluid accumulators 47, mounted on the frame 19, via hydraulic fluid line 45. The accumulators 47 act as high pressure hydraulic fluid reservoirs subsea, and are interconnected via upper 49 and lower 51 valve manifolds and a network of hydraulic lines 50. The valve manifolds 49, 51 are operated by an operator on the support vessel by means of a control console 53 via a control line 55. The control console 53 and hence control line 55 may be electrically or hydraulically powered. Accordingly, operation of the control console 53 in the prescribed manner by the operator can raise or lower the hydraulic jacks 27 by injection of pressurised hydraulic fluid into the respective upper or lower hydraulic ports 39, 41. The lower end of the cylinder 31 of hydraulic jacks 27 is mounted to the upper end of a set of stationary slips 57 which can be operated from the support vessel 3 to selectively grip and hence support the tubing string 16. The set of stationary slips 57 are substantially the same as a conventional set of stationary slips (previously in the prior art only used above the sea surface), although they will likely require modification for use subsea as in the present invention. The stationary slips 57 are mounted to the upper end of a subsea stack 59 which is mounted directly above a blowout prevent or (BOP) 61, which as is conventional, comprises an upper set of rams 62 for emergency sealing about the tubing string 16, a middle set of shear rams 63 which are capable of cutting through the tubing string 16, and a lower set of pipe rams 64 which can be operated in an emergency to seal around the tubing string 16; the BOP is only operated in the event of an emergency in order to seal off the wellbore located below the SOP 61. Conventional choke 67 and kill 69 lines run from the support vessel 3 down to the SOP 61, and are strapped to the riser 15. It should be noted that with the apparatus 1 as shown in FIGS. 1 to 5, there is a further fluid tank (not shown) located on the support vessel 3, and due to the presence of the lower riser seal 21, it is possible to fill the annulus between the inner circumference of the riser and the outer circumference of the tubing/drillstring 16 with fluid supplied from the fluid tank which provides the advantage that the riser is prevented from collapsing due to the hydrostatic pressure of the ocean. In addition, communication lines 71 connect the wellbore, via the SOP 61, to the inner circumference of the lower riser seal 21 and hence the communication lines 71 provide communication between the wellbore and the riser 15. This provides the advantage that if a severe gas kick is experienced from the wellbore, the fluid in the tank will rise, and an operator or a sensor on the support vessel 3 can observe this rise in fluid level in the tank and operate the SOP stack 61 to close the wellbore. In addition, the advantage is provided that if a drilling operation is being conducted through a formation which is relatively low in pressure, the fluid in the riser will quickly drain into the formation and hence the fluid level in the tank will sharply fall and in this scenario the operator can also operate the SOP stack 61 to close the wellbore whilst remedial work can be conducted. Alternatively, and as shown in the second embodiment of apparatus 100 disclosed in FIG. 6 and FIG. 7, it is possible to omit the lower riser seal 21, such that the annulus between the outer circumference of the tubing/drillstring 16 and the inner circumference of the riser 15 is filled with sea water in order to prevent the riser 15 from collapsing. In the apparatus 100, the lower end of the riser 15 secures to a flange located at the uppermost portion of the frame 19, and will thus be open to the sea water. However, the embodiment as shown in FIGS. 1 to 5 is preferred since this provides the advantage that the operator can view the result of the wellbore pressure. In all other respects, the apparatus 100 as shown in FIGS. 6 and 7 is identical with the apparatus 1 as shown in FIGS. 1 to 5. Operation of the apparatus of FIG. 1 will now be described in relation to a drilling operation, but those experienced in the art will appreciate that apparatus 1 can be operated for a host of other operations, such as well intervention, for example. In particular, as those skilled in the art will appreciate, if a through (production) tubing operation is to be performed, the production tree (not shown) is left in place at the mouth of the wellbore, and the frame 19 including the subsea stack 59 and SOP 61 are lowered from the vessel 3 and coupled to the production tree. However, if a full workover operation is to be conducted, the production tubing (not shown) is removed from the wellbore, with the appropriate plugs (not shown) having been placed into the wellbore, and thereafter the frame 19 including the subsea stack 59 and SOP 61 are lowered from the vessel 3 and coupled to the mouth of the wellbore. It should also be noted that the apparatus 100 is operated in a similar manner to the apparatus 1, with the exception that the lower seal riser 21 is not present in the apparatus 100. The tubing string 16, in this case the drillstring 16, is progressively made up in the upper tong 9 and lower rotary back-up 11 and is lowered into the well by the crane 5 and associated winch, with successive drillpipe being included in the drillstring 16. The drillstring 16 first enters the riser 15 and continues downwards until it enters the lower riser seal 21. The lower end of the drillstring 16 then enters the bore of the travelling slips 23, at which point the travelling slips 23 are actuated to securely grip the lower end of the drillstring 16. The hydraulic jack 27 is then operated by an operator at the control console 53, such that the hydraulic jacks 27 are lowered. Hence, the travelling slips 23 take the weight of the drillstring 16. The jacks 27 continue to be lowered until the lower end of the drillstring passes through the stationary slips 57, at which point movement of the hydraulic jacks 27 is halted and the stationary slips 57 are actuated to securely grip the lower end of the drillstring 16. The travelling slips 23 are then disengaged from gripping the drillstring 16 and hence the weight of the drillstring 16 is held by the stationary slips 57. The jacks 27 can then be actuated in the reverse direction, that is to lift the travelling slips 23 upwards until the jacks 27 have reached their, full stroke and the travelling slips 23 can then be actuated once again to grip the next section of drillstring 16; thereafter the stationary slips 57 engagement of the drill string 16 is released. By repeating this operation, the drillstring 16 can be inserted into the wellbore in a much faster manner than achieved using conventional methods. In order to retrieve the drillstring 16 from the wellbore, the apparatus is operated in the reverse direction to that described above, and again, the drillstring 16 can be removed from the wellbore in a manner much quicker than is capable of being achieved with conventional methods. The subsea equipment 17, including the frame 19, subsea stack 59 and SOP 61 can then be retrieved to the vessel 3, after the wellbore has been suitably sealed. Alternatively, if a through (production) tubing operation has been performed, the production tree is left in place at the mouth of the wellbore, and the subsea equipment 17 is de-coupled from the production tree and is retrieved to the vessel. However, if a full workover operation has been conducted, the subsea equipment 17 is de-coupled from the wellbore and is retrieved to the vessel; the production tree is then reinserted in the wellbore mouth, and the appropriate plugs can also then be removed. The embodiments described herein provide the great advantage that most support vessels currently used in the offshore oil and gas industry can be used for such operations with relatively minimal conversion being required, which results in substantial cost and manpower savings. In addition, there is only a very limited supply of jack up rigs available, which means that such operations must wait until a suitable vessel is available; the embodiments described herein substantially increase the number of vessels available in the world capable of performing such operations. Modifications and improvements may be made to the embodiments herein described, without departing from the scope of the invention. |
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abstract | A clamper (23) includes a clamp mechanism (231) which inserts an insertion portion (2311a) into a tube member and clamps the tube member and a lifting and lowering mechanism (232) which lifts and lowers the clamp mechanism (231) in the insertion direction of the insertion portion (2311a). The clamp mechanism (231) includes a clamp body (2311) which has the insertion portion (2311a), a cotter (2312) which protrudes from the insertion portion (2311a) to come into friction-contact with the tube member and is movable in a reciprocating manner in the insertion direction of the insertion portion (2311a), a piston rod (2313) which presses the cotter (2312) so that the cotter protrudes from the insertion portion (2311a) when being pulled toward the opposite side to the insertion direction of the insertion portion (2311a), and a rod cylinder (2315) which is integrated with the clamp body (2311) and pulls the piston rod (2313). |
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abstract | A fuel rack apparatus includes: a base plate having an upper surface and a lower surface; and a plurality of storage tubes coupled to and extending upward from the upper surface of the base plate, the storage tubes arranged in a side-by-side arrangement to form an array of the storage tubes. Each of the storage tubes extend along a longitudinal axis and include: an outer tube having an inner surface defining an inner cavity; and an inner plate-assemblage positioned within the outer tube that divides the inner cavity into a plurality of interior flux trap chambers and a fuel storage cell. |
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