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Sample records for breeder reactor ii

  1. Shutdown and Closure of the Experimental Breeder Reactor - II

    SciTech Connect

    Michelbacher, John A.; Baily, Carl E.; Baird, Daniel K.; Henslee, S. Paul; Knight, Collin J.; Rosenberg, Kenneth E.

    2002-07-01

    The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor - II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m{sup 3} (86,000 gallons) of sodium and the secondary system contained 50 m{sup 3} (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated lay-up plan defining the system end state, as well as instructions for achieving the lay-up condition. A goal of system-by-system lay-up is to minimize

  2. Shutdown and closure of the experimental breeder reactor - II.

    SciTech Connect

    Michelbacher, J. A.; Baily, C. E.; Baird, D. K.; Henslee, S. P.; Knight, C. J.; Rosenberg, K. E.

    2002-09-26

    The Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to maintain the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The EBR-II is a pool-type reactor. The primary system contained approximately 325 m{sup 3} (86,000 gallons) of sodium and the secondary system contained 50 m{sub 3} (13,000 gallons). In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility was built to react the sodium to a solid sodium hydroxide monolith for burial as a low level waste in a land disposal facility. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in circuits and components must be passivated, inerted, or removed to preclude future concerns with sodium-air reactions that could generate potentially explosive mixtures of hydrogen and leave corrosive compounds. The passivation process being implemented utilizes a moist carbon dioxide gas that generates a passive layer of sodium carbonate/sodium bicarbonate over any quantities of residual sodium. Tests being conducted will determine the maximum depths of sodium that can be reacted using this method, defining the amount that must be dealt with later to achieve RCRA clean closure. Deactivation of the EBR-II complex is on schedule for a March, 2002, completion. Each system associated with EBR-II has an associated layup plan defining the system end state, as well as instructions for achieving the layup condition. A goal of system-by-system layup is to minimize surveillance

  3. Decommissioning of Experimental Breeder Reactor - II Complex, Post Sodium Draining

    SciTech Connect

    J. A. Michelbacher; S. Paul Henslee; Collin J. Knight; Steven R. sherman

    2005-09-01

    The Experimental Breeder Reactor - II (EBR-II) was shutdown in September 1994 as mandated by the United States Department of Energy. This sodium-cooled reactor had been in service since 1964. The bulk sodium was drained from the primary and secondary systems and processed. Residual sodium remaining in the systems after draining was converted into sodium bicarbonate using humid carbon dioxide. This technique was tested at Argonne National Laboratory in Illinois under controlled conditions, then demonstrated on a larger scale by treating residual sodium within the EBR-II secondary cooling system, followed by the primary tank. This process, terminated in 2002, was used to place a layer of sodium bicarbonate over all exposed surfaces of sodium. Treatment of the remaining EBR-II sodium is governed by the Resource Conservation and Recovery Act (RCRA). The Idaho Department of Environmental Quality issued a RCRA Operating Permit in 2002, mandating that all hazardous materials be removed from EBR-II within a 10 year period, with the ability to extend the permit and treatment period for another 10 years. A preliminary plan has been formulated to remove the remaining sodium and NaK from the primary and secondary systems using moist carbon dioxide, steam and nitrogen, and a water flush. The moist carbon dioxide treatment was resumed in May 2004. As of August 2005, approximately 60% of the residual sodium within the EBR-II primary tank had been treated. This process will continue through the end of 2005, when it is forecast that the process will become increasingly ineffective. At that time, subsequent treatment processes will be planned and initiated. It should be noted that the processes and anticipated costs associated with these processes are preliminary. Detailed engineering has not been performed, and approval for these methods has not been obtained from the regulator or the sponsors.

  4. Instrumentation and control improvements at Experimental Breeder Reactor II

    SciTech Connect

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I C systems of the next generation of liquid metal reactor (LMR) plants.

  5. Instrumentation and control improvements at Experimental Breeder Reactor II

    SciTech Connect

    Christensen, L.J.; Planchon, H.P.

    1993-03-01

    The purpose of this paper is to describe instrumentation and control (I&C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I&C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I&C systems of the next generation of liquid metal reactor (LMR) plants.

  6. Data handling at EBR-II (Experimental Breeder Reactor II) for advanced diagnostics and control work

    SciTech Connect

    Lindsay, R.W.; Schorzman, L.W.

    1988-01-01

    Improved control and diagnostics systems are being developed for nuclear and other applications. The Experimental Breeder Reactor II (EBR-II) Division of Argonne National Laboratory has embarked on a project to upgrade the EBR-II control and data handling systems. The nature of the work at EBR-II requires that reactor plant data be readily available for experimenters, and that the plant control systems be flexible to accommodate testing and development needs. In addition, operational concerns require that improved operator interfaces and computerized diagnostics be included in the reactor plant control system. The EBR-II systems have been upgraded to incorporate new data handling computers, new digital plant process controllers, and new displays and diagnostics are being developed and tested for permanent use. In addition, improved engineering surveillance will be possible with the new systems.

  7. Scram reliability under seismic conditions at the Experimental Breeder Reactor II

    SciTech Connect

    Roglans, J.; Wang, C.Y.; Hill, D.J.

    1993-08-01

    A Probabilistic Risk Assessment of the Experimental Breeder Reactor II has recently been completed. Seismic events are among the external initiating events included in the assessment. As part of the seismic PRA a detailed study has been performed of the ability to shutdown the reactor under seismic conditions. A comprehensive finite element model of the EBR-II control rod drive system has been used to analyze the control rod system response when subjected to input seismic accelerators. The results indicate the control rod drive system has a high seismic capacity. The estimated seismic fragility for the overall reactor shutdown system is dominated by the primary tank failure.

  8. Fast Breeder Reactor studies

    SciTech Connect

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  9. Low-order dynamic modeling of the Experimental Breeder Reactor II

    SciTech Connect

    Berkan, R.C. . Dept. of Nuclear Engineering); Upadhyaya, B.R.; Kisner, R.A. )

    1990-07-01

    This report describes the development of a low-order, linear model of the Experimental Breeder Reactor II (EBR-II), including the primary system, intermediate heat exchanger, and steam generator subsystems. The linear model is developed to represent full-power steady state dynamics for low-level perturbations. Transient simulations are performed using model building and simulation capabilities of the computer software Matrix{sub x}. The inherently safe characteristics of the EBR-II are verified through the simulation studies. The results presented in this report also indicate an agreement between the linear model and the actual dynamics of the plant for several transients. Such models play a major role in the learning and in the improvement of nuclear reactor dynamics for control and signal validation studies. This research and development is sponsored by the Advanced Controls Program in the Instrumentation and Controls Division of the Oak Ridge National Laboratory. 17 refs., 67 figs., 15 tabs.

  10. Breeder reactors in France

    SciTech Connect

    Zaleski, C.P.

    1980-04-11

    France relies on nuclear power as an important part of her energy program. Anticipating problems with the availability of natural uranium before the year 2020, the French have been pursuing a three-stage program of development of breeder reactors. The third reactor in this program, the near-commercial plant Super Phenix Mark I, is expected to reach power operation in 1983. Although there are still some uncertainties, particularly about the date when the breeder will become competitive with other energy sources, the outlook is considered favorable and preliminary designs for commercial plants are under way.

  11. Statistical analysis of duplex-tube performance in Experimental Breeder Reactor II superheater SU-712

    SciTech Connect

    Gross, K.C.; Seidel, B.R.

    1986-06-01

    A detailed investigation was made of historical data recorded at Experimental Breeder Reactor II during operation with superheater SU-712. The objective of this study was to analyze and characterize the performance of 72 duplex steam tubes that became unbonded during a long period of operation. The information processing system ANALYZE was developed to perform the required numerical manipulations, statistical analyses, and correlation analyses with a large data base containing some five million data values. The ANALYZE system was successfully employed (a) to characterize the performance of all the steam tubes in terms of frequency and relative severity of unbonding, and (b) to establish a correlation between the observed anomalous behavior of the superheater and its operating parameters. Results from this investigation were used to select sections for materials examinations and physical tests that were performed after SU-712 was removed from operation.

  12. Experimental Breeder Reactor II (EBR-II): Instrumentation for core surveillance

    SciTech Connect

    Christensen, L.J.

    1989-01-01

    EBR-II has operated for 25 years in support of several major programs. During this time period, several of the original, non-replaceable, flow sensors, RDT sensors and thermocouples have failed in the primary system. This has led to the development of new sensors and the use of calculated values using computer models of the plant. It is important for the next generation of LMR reactors to minimize or eliminate the use of non-replaceable sensors. EBR-II is perhaps the best modeled reactor in the world, thanks to a dedicated T-H analysis program. The success of this program relied on excellent measurements of temperature and flow in subassemblies in the core. The instrumented subassemblies of the XX series provided that measurement capability. From this test series, EBR-II calculations showed that the core could withstand a loss-of-flow without scram accident and a loss-of-heat sink without scram accident from full reactor power without core damage. From this, reactor designers can now design with confidence, inherently safe reactors. 11 refs., 8 figs.

  13. Seventeen years of LMFBR experience: Experimental Breeder Reactor II (EBR-II)

    SciTech Connect

    Perry, W.H.; Lentz, G.L.; Richardson, W.J.; Wolz, G.C.

    1982-05-01

    Operating experience at EBR-II over the past 17 years has shown that a sodium-cooled pool-type reactor can be safely and efficiently operated and maintained. The reactor has performed predictably and benignly during normal operation and during both unplanned and planned plant upsets. The duplex-tube evaporators and superheaters have never experienced a sodium/water leak, and the rest of the steam-generating system has operated without incident. There has been no noticeable degradation of the heat transfer efficiency of the evaporators and superheaters, except for the one superheater replaced in 1981. There has been no need to perform any chemical cleaning of steam-system components. Operation of EBR-II has produced a wealth of information. As an irradiation facility, EBR-II has generated specific information on the behavior of oxide, carbide, and metallic fuels. As an LMFBR power plant, EBR-II has produced general information related to plant-systems and equipment design, plant safety, plant availability, and plant maintenance.

  14. Seventeen years of LMFBR experience: Experimental Breeder Reactor II (EBR-II)

    SciTech Connect

    Perry, W.H.; Lentz, G.L.; Richardson, W.J.; Wolz, G.C.

    1982-01-01

    Operating experience at EBR-II over the past 17 years has shown that a sodium-cooled pool-type reactor can be safely and efficiently operated and maintained. The reactor has performed predictably and benignly during normal operation and during both unplanned and planned plant upsets. The duplex-tube evaporators and superheaters have never experienced a sodium/water leak, and the rest of the steam-generating system has operated without incident. There has been no noticeable degradation of the heat transfer efficiency of the evaporators and superheaters, except for the one superheater replaced in 1981. There has been no need to perform any chemical cleaning of steam-system components.

  15. Experimental Breeder Reactor II (EBR-II) Fuel-Performance Test Facility (FPTF)

    SciTech Connect

    Pardini, J.A.; Brubaker, R.C.; Veith, D.J.; Giorgis, G.C.; Walker, D.E.; Seim, O.S.

    1982-01-01

    The Fuel-Performance Test Facility (FPTF) is the latest in a series of special EBR-II instrumented in-core test facilities. A flow control valve in the facility is programmed to vary the coolant flow, and thus the temperature, in an experimental-irradiation subassembly beneath it and coupled to it. In this way, thermal transients can be simulated in that subassembly without changing the temperatures in surrounding subassemblies. The FPTF also monitors sodium flow and temperature, and detects delayed neutrons in the sodium effluent from the experimental-irradiation subassembly beneath it. This facility also has an acoustical detector (high-temperature microphone) for detecting sodium boiling.

  16. Breeder Reactors, Understanding the Atom Series.

    ERIC Educational Resources Information Center

    Mitchell, Walter, III; Turner, Stanley E.

    The theory of breeder reactors in relationship to a discussion of fission is presented. Different kinds of reactors are characterized by the cooling fluids used, such as liquid metal, gas, and molten salt. The historical development of breeder reactors over the past twenty-five years includes specific examples of reactors. The location and a brief…

  17. POWER BREEDER REACTOR

    DOEpatents

    Monson, H.O.

    1960-11-22

    An arrangement is offered for preventing or minimizing the contraction due to temperature rise, of a reactor core comprising vertical fuel rods in sodium. Temperature rise of the fuel rods would normally make them move closer together by inward bowing, with a resultant undesired increase in reactivity. According to the present invention, assemblies of the fuel rods are laterally restrained at the lower ends of their lower blanket sections and just above the middle of the fuel sections proper of the rods, and thus the fuel sections move apart, rather than together, with increase in temperature.

  18. An integrated plant-life extension program for EBR-II (Experimental Breeder Reactor)

    SciTech Connect

    King, R.W.

    1986-01-01

    An integrated plant-life extension program is being developed and implemented at EBR-II. The program plan has five primary areas of focus, and is structured to take advantage of inherent features of the liquid-metal-cooled reactor that give it potential for very long life. The program is centered around development and increased use of computer-based software systems for surveillance, diagnostics, prognostics, data handling, and knowledge transfer. Even though the program is only partially implemented, benefits are already being realized in the form of increased understanding of plant system status and performance due to development of diagnostic data-handling software for manipulation of plant sensor data, and improved force monitoring and protection of the remotely operated fuel handling system. The eventual integration of the elements of the program is a key feature that is expected to enhance the overall effectiveness of the program.

  19. NUCLEAR REACTOR FUEL-BREEDER FUEL ELEMENT

    DOEpatents

    Currier, E.L. Jr.; Nicklas, J.H.

    1962-08-14

    A fuel-breeder fuel element was developed for a nuclear reactor wherein discrete particles of fissionable material are dispersed in a matrix of fertile breeder material. The fuel element combines the advantages of a dispersion type and a breeder-type. (AEC)

  20. Fast breeder reactor protection system

    DOEpatents

    van Erp, J.B.

    1973-10-01

    Reactor protection is provided for a liquid-metal-fast breeder reactor core by measuring the coolant outflow temperature from each of the subassemblies of the core. The outputs of the temperature sensors from a subassembly region of the core containing a plurality of subassemblies are combined in a logic circuit which develops a scram alarm if a predetermined number of the sensors indicate an over temperature condition. The coolant outflow from a single subassembly can be mixed with the coolant outflow from adjacent subassemblies prior to the temperature sensing to increase the sensitivity of the protection system to a single subassembly failure. Coherence between the sensors can be required to discriminate against noise signals. (Official Gazette)

  1. Off-normal performance of EBR-II (Experimental Breeder Reactor) driver fuel

    SciTech Connect

    Seidel, B.R.; Batte, G.L.; Lahm, C.E.; Fryer, R.M.; Koenig, J.F.; Hofman, G.L.

    1986-09-01

    The off-normal performance of EBR-II Mark-II driver fuel has been more than satisfactory as demonstrated by robust reliability under repeated transient overpower and undercooled loss-of-flow tests, by benign run-beyond-cladding-breach behavior, and by forgiving response to fabrication defects including lack of bond. Test results have verified that the metallic driver fuel is very tolerant of off-normal events. This behavior has allowed EBR-II to operate in a combined steady-state and transient mode to provide test capability without limitation from the metallic driver fuel.

  2. Light-Water Breeder Reactor

    DOEpatents

    Beaudoin, B. R.; Cohen, J. D.; Jones, D. H.; Marier, Jr, L. J.; Raab, H. F.

    1972-06-20

    Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

  3. Advanced absorber assembly design for breeder reactors

    SciTech Connect

    Pitner, A.L.; Birney, K.R.

    1980-01-01

    An advanced absorber assembly design has been developed for breeder reactor control rod applications that provides for improved in-reactor performance, longer lifetimes, and reduced fabrication costs. The design comprises 19 vented pins arranged in a circular array inside of round duct tubes. The absorber material is boron carbide; cladding and duct components are constructed from the modified Type 316 stainless steel alloy. Analyses indicate that this design will scram 30 to 40% faster than the reference FFTF absorber assembly. The basic design characteristics of this advanced FFTF absorber assembly are applicable to large core breeder reactor design concepts.

  4. Deployment Scenario of Heavy Water Cooled Thorium Breeder Reactor

    SciTech Connect

    Mardiansah, Deby; Takaki, Naoyuki

    2010-06-22

    Deployment scenario of heavy water cooled thorium breeder reactor has been studied. We have assumed to use plutonium and thorium oxide fuel in water cooled reactor to produce {sup 233}U which will be used in thorium breeder reactor. The objective is to analysis the potential of water cooled Th-Pu reactor for replacing all of current LWRs especially in Japan. In this paper, the standard Pressurize Water Reactor (PWR) has been designed to produce 3423 MWt; (i) Th-Pu PWR, (ii) Th-Pu HWR (MFR = 1.0) and (iii) Th-Pu HWR (MFR 1.2). The properties and performance of the core were investigated by using cell and core calculation code. Th-Pu PWR or HWR produces {sup 233}U to introduce thorium breeder reactor. The result showed that to replace all (60 GWe) LWR by thorium breeder reactor within a period of one century, Th-Pu oxide fueled PWR has insufficient capability to produce necessary amount of {sup 233}U and Th-Pu oxide fueled HWR has almost enough potential to produce {sup 233}U but shows positive void reactivity coefficient.

  5. Experimental Breeder Reactor I Preservation Plan

    SciTech Connect

    Julie Braun

    2006-10-01

    Experimental Breeder Reactor I (EBR I) is a National Historic Landmark located at the Idaho National Laboratory, a Department of Energy laboratory in southeastern Idaho. The facility is significant for its association and contributions to the development of nuclear reactor testing and development. This Plan includes a structural assessment of the interior and exterior of the EBR I Reactor Building from a preservation, rather than an engineering stand point and recommendations for maintenance to ensure its continued protection.

  6. BREEDER: a microcomputer program for financial analysis of a large-scale prototype breeder reactor

    SciTech Connect

    Giese, R.F.

    1984-04-01

    This report describes a microcomputer-based, single-project financial analysis program: BREEDER. BREEDER is a user-friendly model designed to facilitate frequent and rapid analyses of the financial implications associated with alternative design and financing strategies for electric generating plants and large-scale prototype breeder (LSPB) reactors in particular. The model has proved to be a useful tool in establishing cost goals for LSPB reactors. The program is available on floppy disks for use on an IBM personal computer (or IBM look-a-like) running under PC-DOS or a Kaypro II transportable computer running under CP/M (and many other CP/M machines). The report documents version 1.5 of BREEDER and contains a user's guide. The report also includes a general overview of BREEDER, a summary of hardware requirements, a definition of all required program inputs, a description of all algorithms used in performing the construction-period and operation-period analyses, and a summary of all available reports. The appendixes contain a complete source-code listing, a cross-reference table, a sample interactive session, several sample runs, and additional documentation of the net-equity program option.

  7. Fast Breeder Reactors in Sweden: Vision and Reality.

    PubMed

    Fjaestad, Maja

    2015-01-01

    The fast breeder is a type of nuclear reactor that aroused much attention in the 1950s and '60s. Its ability to produce more nuclear fuel than it consumes offered promises of cheap and reliable energy. Sweden had advanced plans for a nuclear breeder program, but canceled them in the middle of the 1970s with the rise of nuclear skepticism. The article investigates the nuclear breeder as a technological vision. The nuclear breeder reactor is an example of a technological future that did not meet its industrial expectations. But that does not change the fact that the breeder was an influential technology. Decisions about the contemporary reactors were taken with the idea that in a foreseeable future they would be replaced with the efficient breeder. The article argues that general themes in the history of the breeder reactor can deepen our understanding of the mechanisms behind technological change. PMID:26334698

  8. Fusion reactor breeder material safety compatibility studies

    SciTech Connect

    Jeppson, D.W.; Cohen, S.; Muhlestein, L.D.

    1983-09-01

    Tritium breeder material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Breeder material safety compatibility studies are being conducted to identify and characterize breeder-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate the following. 1. Ternary oxides (LiAlO/sub 2/, Li/sub 2/ZrO/sub 3/, Li/sub 2/SiO/sub 3/, Li/sub 4/SiO/sub 4/, and LiTiO/sub 3/) at postulated blanket operating temperatures are chemically compatible with water coolant, while liquid lithium and Li/sub 7/Pb/sub 2/ reactions with water generate heat, aerosol, and hydrogen. 2. Lithium oxide and 17Li-83Pb alloy react mildly with water requiring special precautions to control hydrogen release. 3. Liquid lithium reacts substantially, while 17Li83Pb alloy reacts mildly with concrete to produce hydrogen. 4. Liquid lithium-air reactions may present some major safety concerns. Additional scoping tests are needed, but the ternary oxides, lithium oxide, and 17Li-83Pb have definite safety advantages over liquid lithium and Li/sub 7/Pb/sub 2/. The ternary oxides present minimal safetyrelated problems when used with water as coolant, air or concrete; but they do require neutron multipliers, which may have safety compatibility concerns with surrounding materials. The combined favorable neutronics and minor safety compatibility concerns of lithium oxide and 17Li-83Pb make them prime candidates as breeder materials. Current safety efforts are directed toward assessing the compatibility of lithium oxide and the lithium-lead alloy with coolants and other materials.

  9. Prediction of stainless steel activation in experimental breeder reactor 2 (EBR-II) reflector and blanket subassemblies

    SciTech Connect

    Bunde, K.A.

    1996-12-31

    Stainless steel structural components in nuclear reactors become radioactive wastes when no longer useful. Prior to disposal, certain physical attributes must be analyzed. These attributes include structural integrity, chemical stability, and the radioactive material content among others. The focus of this work is the estimation of the radioactive material content of stainless steel wastes from a research reactor operated by Argonne National Laboratory.

  10. Analysis of UF6 breeder reactor power plants

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.

    1976-01-01

    Gaseous UF6 fueled breeder reactor design and technical applications of such concepts are summarized. Special attention was given to application in nuclear power plants and to reactor efficiency and safety factors.

  11. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    SciTech Connect

    Not Available

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  12. Atoms in Appalachia. Historical report on the Clinch River Breeder Reactor site

    SciTech Connect

    Schaffer, D

    1982-01-01

    The background information concerning the acquisition of the land for siting the Clinch River Breeder Reactor is presented. Historical information is also presented concerning the land acquisition for the Oak Ridge facilities known as the Manhattan Project during World War II.

  13. Nuclear breeder reactor fuel element with silicon carbide getter

    DOEpatents

    Christiansen, David W.; Karnesky, Richard A.

    1987-01-01

    An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

  14. Clinch River Breeder Reactor Plant Project: construction schedule

    SciTech Connect

    Purcell, W.J.; Martin, E.M.; Shivley, J.M.

    1982-01-01

    The construction schedule for the Clinch River Breeder Reactor Plant and its evolution are described. The initial schedule basis, changes necessitated by the evaluation of the overall plant design, and constructability improvements that have been effected to assure adherence to the schedule are presented. The schedule structure and hierarchy are discussed, as are tools used to define, develop, and evaluate the schedule.

  15. Fission-suppressed hybrid reactor: the fusion breeder

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Coops, M.S.

    1982-12-01

    Results of a conceptual design study of a /sup 233/U-producing fusion breeder are presented. The majority of the study was devoted to conceptual design and evaluation of a fission-suppressed blanket and to fuel cycle issues such as fuel reprocessing, fuel handling, and fuel management. Studies in the areas of fusion engineering, reactor safety, and economics were also performed.

  16. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    SciTech Connect

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  17. Feasibility study on the thorium fueled boiling water breeder reactor

    SciTech Connect

    PetrusTakaki, N.

    2012-07-01

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  18. COUPLED FAST-THERMAL POWER BREEDER REACTOR

    DOEpatents

    Avery, R.

    1961-07-18

    A nuclear reactor having a region operating predominantly on fast neutrons and another region operating predominantly on slow neutrons is described. The fast region is a plutonium core and the slow region is a natural uranium blanket around the core. Both of these regions are free of moderator. A moderating reflector surrounds the uranium blanket. The moderating material and thickness of the reflector are selected so that fissions in the uranium blanket make a substantial contribution to the reactivity of the reactor.

  19. Gas core reactors for actinide transmutation and breeder applications

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.

    1978-01-01

    This work consists of design power plant studies for four types of reactor systems: uranium plasma core breeder, uranium plasma core actinide transmuter, UF6 breeder and UF6 actinide transmuter. The plasma core systems can be coupled to MHD generators to obtain high efficiency electrical power generation. A 1074 MWt UF6 breeder reactor was designed with a breeding ratio of 1.002 to guard against diversion of fuel. Using molten salt technology and a superheated steam cycle, an efficiency of 39.2% was obtained for the plant and the U233 inventory in the core and heat exchangers was limited to 105 Kg. It was found that the UF6 reactor can produce high fluxes (10 to the 14th power n/sq cm-sec) necessary for efficient burnup of actinide. However, the buildup of fissile isotopes posed severe heat transfer problems. Therefore, the flux in the actinide region must be decreased with time. Consequently, only beginning-of-life conditions were considered for the power plant design. A 577 MWt UF6 actinide transmutation reactor power plant was designed to operate with 39.3% efficiency and 102 Kg of U233 in the core and heat exchanger for beginning-of-life conditions.

  20. MOLTEN PLUTONIUM FUELED FAST BREEDER REACTOR

    DOEpatents

    Kiehn, R.M.; King, L.D.P.; Peterson, R.E.; Swickard, E.O. Jr.

    1962-06-26

    A description is given of a nuclear fast reactor fueled with molten plutonium containing about 20 kg of plutonium in a tantalum container, cooled by circulating liquid sodium at about 600 to 650 deg C, having a large negative temperature coefficient of reactivity, and control rods and movable reflector for criticality control. (AEC)

  1. Steam generator for liquid metal fast breeder reactor

    DOEpatents

    Gillett, James E.; Garner, Daniel C.; Wineman, Arthur L.; Robey, Robert M.

    1985-01-01

    Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

  2. Mechanical design of a light water breeder reactor

    DOEpatents

    Fauth, Jr., William L.; Jones, Daniel S.; Kolsun, George J.; Erbes, John G.; Brennan, John J.; Weissburg, James A.; Sharbaugh, John E.

    1976-01-01

    In a light water reactor system using the thorium-232 -- uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements.

  3. Breeding nuclear fuels with accelerators: replacement for breeder reactors

    SciTech Connect

    Grand, P.; Takahashi, H.

    1984-01-01

    One application of high energy particle accelerators has been, and still is, the production of nuclear fuel for the nuclear energy industry; tantalizing because it would create a whole new industry. This approach to producing fissile from fertile material was first considered in the early 1950's in the context of the nuclear weapons program. A considerable development effort was expended before discovery of uranium ore in New Mexico put an end to the project. Later, US commitment to the Liquid Metal Fast Breeder Reactors (LMFBR) killed any further interest in pursuing accelerator breeder technology. Interest in the application of accelerators to breed nuclear fuels, and possibly burn nuclear wastes, revived in the late 1970's, when the LMFBR came under attack during the Carter administration. This period gave the opportunity to revisit the concept in view of the present state of the technology. This evaluation and the extensive calculational modeling of target designs that have been carried out are promising. In fact, a nuclear fuel cycle of Light Water Reactors and Accelerator Breeders is competitive to that of the LMFBR. At this time, however, the relative abundance of uranium reserves vs electricity demand and projected growth rate render this study purely academic. It will be for the next generation of accelerator builders to demonstate the competitiveness of this technology versus that of other nuclear fuel cycles, such as LMFBR's or Fusion Hybrid systems. 22 references, 1 figure, 5 tables.

  4. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    SciTech Connect

    Illum, D.B.; Olson, G.L.; McCardell, R.K.

    1999-01-01

    The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

  5. Steam generator tubing development for commercial fast breeder reactors

    SciTech Connect

    Sessions, C.E.; Uber, C.F.

    1981-11-01

    The development work to design, manufacture, and evaluate pre-stressed double-wall 2/one quarter/ Cr-1 Mo steel tubing for commercial fast breeder reactor steam generator application is discussed. The Westinghouse plan for qualifying tubing vendors to produce this tubing is described. The results achieved to date show that a long length pre-stressed double-wall tube is both feasible and commercially available. The evaluation included structural analysis and experimental measurement of the pre-stress within tubes, as well as dimensional, metallurgical, and interface wear tests of tube samples produced. This work is summarized and found to meet the steam generator design requirements. 10 refs.

  6. Secure automated fabrication: remote fabrication of breeder-reactor fuel

    SciTech Connect

    Gerber, E.W.; Rice, L.H.; Horgos, R.M.; Nagamoto, T.T.; Graham, R.A.

    1981-05-01

    The Secure Automated Fabrication (SAF) Program was initiated at the Hanford Engineering Development Laboratory (HEDL) to develop and demonstrate an advanced manufacturing line (SAF line) for plutonium oxide breeder reactor fuel pins. The SAF line is to be installed in the Fuels and Materials Examination Facility (FMEF) at Hanford and will utilize technology that focuses on improved safety features for plant operating personnel, the public, and the environment. Equipment and process improvements incorporated by the SAF line will yield significant gains in nuclear materials safeguards, product quality and productivity.

  7. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    SciTech Connect

    Seifritz, W.

    1983-11-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase.

  8. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    SciTech Connect

    Adams, S.R.

    1985-10-01

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

  9. UF6 breeder reactor power plants for electric power generation

    NASA Technical Reports Server (NTRS)

    Rust, J. H.; Clement, J. D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

  10. Light-water breeder reactor (LWBR Development Program)

    DOEpatents

    Beaudoin, B.R.; Cohen, J.D.; Jones, D.H.; Marier, L.J. Jr.; Raab, H.F.

    1972-06-20

    Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

  11. Designing a SCADA system simulator for fast breeder reactor

    NASA Astrophysics Data System (ADS)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

    2016-04-01

    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  12. Accident analysis of heavy water cooled thorium breeder reactor

    NASA Astrophysics Data System (ADS)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  13. Accident analysis of heavy water cooled thorium breeder reactor

    SciTech Connect

    Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

    2015-04-16

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of < 0.2% dk/k, and negative coolant reactivity coefficient. One of the nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The

  14. Comparative assessment of nuclear fuel cycles. Light-water reactor once-through, classical fast breeder reactor, and symbiotic fast breeder reactor cycles

    SciTech Connect

    Hardie, R.W.; Barrett, R.J.; Freiwald, J.G.

    1980-06-01

    The object of the Alternative Nuclear Fuel Cycle Study is to perform comparative assessments of nuclear power systems. There are two important features of this study. First, this evaluation attempts to encompass the complete, integrated fuel cycle from mining of uranium ore to disposal of waste rather than isolated components. Second, it compares several aspects of each cycle - energy use, economics, technological status, proliferation, public safety, and commercial potential - instead of concentrating on one or two assessment areas. This report presents assessment results for three fuel cycles. These are the light-water reactor once-through cycle, the fast breeder reactor on the classical plutonium cycle, and the fast breeder reactor on a symbiotic cycle using plutonium and /sup 233/U as fissile fuels. The report also contains a description of the methodology used in this assessment. Subsequent reports will present results for additional fuel cycles.

  15. The Case Against the Fast Breeder Reactor: An Anti-Nuclear Establishment View.

    ERIC Educational Resources Information Center

    Lovins, Amory B.

    1973-01-01

    Environmentalists lobby points out that hazards which may result from mistakes in proposed fast breeder reactor for additional energy can be detrimental for mankind. Such projects must be carefully planned and cautiously executed. (PS)

  16. Alloys for a liquid metal fast breeder reactor

    DOEpatents

    Rowcliffe, Arthur F.; Bleiberg, Melvin L.; Diamond, Sidney; Bajaj, Ram

    1979-01-01

    An essentially gamma-prime precipitation-hardened iron-chromium-nickel alloy has been designed with emphasis on minimum nickel and chromium contents to reduce the swelling tendencies of these alloys when used in liquid metal fast breeder reactors. The precipitation-hardening components have been designed for phase stability and such residual elements as silicon and boron, also have been selected to minimize swelling. Using the properties of these alloys in one design would result in an increased breeding ratio over 20% cold worked stainless steel, a reference material, of 1.239 to 1.310 and a reduced doubling time from 15.8 to 11.4 years. The gross stoichiometry of the alloying composition comprises from about 0.04% to about 0.06% carbon, from about 0.05% to about 1.0% silicon, up to about 0.1% zirconium, up to about 0.5% vanadium, from about 24% to about 31% nickel, from 8% to about 11% chromium, from about 1.7% to about 3.5% titanium, from about 1.0% to about 1.8% aluminum, from about 0.9% to about 3.7% molybdenum, from about 0.04% to about 0.8% boron, and the balance iron with incidental impurities.

  17. Ultrasonic decontamination of prototype fast breeder reactor fuel pins.

    PubMed

    Kumar, Aniruddha; Bhatt, R B; Behere, P G; Afzal, Mohd

    2014-04-01

    Fuel pin decontamination is the process of removing particulates of radioactive material from its exterior surface. It is an important process step in nuclear fuel fabrication. It assumes more significance with plutonium bearing fuel known to be highly radio-toxic owing to its relatively longer biological half life and shorter radiological half life. Release of even minute quantity of plutonium oxide powder in the atmosphere during its handling can cause alarming air borne activity and may pose a severe health hazard to personnel working in the vicinity. Decontamination of fuel pins post pellet loading operation is thus mandatory before they are removed from the glove box for further processing and assembly. This paper describes the setting up of ultrasonic decontamination process, installed inside a custom built fume-hood in the production line, comprising of a cleaning tank with transducers, heaters, pin handling device and water filtration system and its application in cleaning of fuel pins for prototype fast breeder reactor. The cleaning process yielded a typical decontamination efficiency of more than 99%. PMID:24405906

  18. Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

    SciTech Connect

    Massimino, R.J.; Williams, D.A.

    1983-05-01

    This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

  19. Analysis of Sodium Fire in the Containment Building of Prototype Fast Breeder Reactor Under the Scenario of Core Disruptive Accident

    SciTech Connect

    Rao, P.M.; Kasinathan, N.; Kannan, S.E.

    2006-07-01

    The potential for sodium release to reactor containment building from reactor assembly during Core Disruptive Accident (CDA) in Fast Breeder Reactors (FBR) is an important safety issue with reference to the structural integrity of Reactor Containment Building (RCB). For Prototype Fast Breeder Reactor (PFBR), the estimated sodium release under a CDA of 100 MJ energy release is 350 kg. The ejected sodium reacts easily with air in RCB and causes temperature and pressure rise in the RCB. For estimating the severe thermal consequences in RCB, different modes of sodium fires like pool and spray fires were analyzed by using SOFIRE -- II and NACOM sodium fire computer codes. Effects of important parameters like amount of sodium, area of pool, containment air volume and oxygen concentration have been investigated. A peak pressure rise of 7.32 kPa is predicted by SOFIRE II code for 350 kg sodium pool fire in 86,000 m{sup 3} RCB volume. Under sodium release as spray followed by unburnt sodium as pool fire mode analysis, the estimated pressure rise is 5.85 kPa in the RCB. In the mode of instantaneous combustion of sodium, the estimated peak pressure rise is 13 kPa. (authors)

  20. Nuclear breeder reactor fuel element with axial tandem stacking and getter

    DOEpatents

    Gibby, Ronald L.; Lawrence, Leo A.; Woodley, Robert E.; Wilson, Charles N.; Weber, Edward T.; Johnson, Carl E.

    1981-01-01

    A breeder reactor fuel element having a tandem arrangement of fissile and fertile fuel with a getter for fission product cesium disposed between the fissile and fertile sections. The getter is effective at reactor operating temperatures to isolate the cesium generated by the fissile material from reacting with the fertile fuel section.

  1. Method of locating a leaking fuel element in a fast breeder power reactor

    DOEpatents

    Honekamp, John R.; Fryer, Richard M.

    1978-01-01

    Leaking fuel elements in a fast reactor are identified by measuring the ratio of .sup.134 Xe to .sup.133 Xe in the reactor cover gas following detection of a fuel element leak, this ratio being indicative of the power and burnup of the failed fuel element. This procedure can be used to identify leaking fuel elements in a power breeder reactor while continuing operation of the reactor since the ratio measured is that of the gases stored in the plenum of the failed fuel element. Thus, use of a cleanup system for the cover gas makes it possible to identify sequentially a multiplicity of leaking fuel elements without shutting the reactor down.

  2. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-2: Liquid Metal Fast Breeder Reactors.

    ERIC Educational Resources Information Center

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical liquid metal fast breeder reactor (LMFBR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating the use with a simplified model. The heart of the module is…

  3. Recommendations concerning models and parameters best suited to breeder reactor environmental radiological assessments

    SciTech Connect

    Miller, C.W.; Baes, C.F. III; Dunning, D.E. Jr.

    1980-05-01

    Recommendations are presented concerning the models and parameters best suited for assessing the impact of radionuclide releases to the environment by breeder reactor facilities. These recommendations are based on the model and parameter evaluations performed during this project to date. Seven different areas are covered in separate sections.

  4. Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

    DOEpatents

    Brehm, Jr., William F.; Colburn, Richard P.

    1982-01-01

    An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.

  5. ORIGEN2 model and results for the Clinch River Breeder Reactor

    SciTech Connect

    Croff, A G; Bjerke, M A

    1982-06-01

    Reactor physics calculations and literature information acquisition have led to the development of a Clinch River Breeder Reactor (CRBR) model for the ORIGEN2 computer code. The model is based on cross sections taken directly from physics codes. Details are presented concerning the physical description of the fuel assemblies, the fuel management scheme, irradiation parameters, and initial material compositions. The ORIGEN2 model for the CRBR has been implemented, resulting in the production of graphical and tabular characteristics (radioactivity, thermal power, and toxicity) of CRBR spent fuel, high-level waste, and fuel-assembly structural material waste as a function of decay time. Characteristics for pressurized water reactors (PWRs), commercial liquid-metal fast breeder reactors (LMFBRs), and the Fast Flux Test Facility (FFTF) have also been included in this report for comparison with the CRBR data.

  6. Feasibility of Water Cooled Thorium Breeder Reactor Based on LWR Technology

    SciTech Connect

    Takaki, Naoyuki; Permana, Sidik; Sekimoto, Hiroshi

    2007-07-01

    The feasibility of Th-{sup 233}U fueled, homogenous breeder reactor based on matured conventional LWR technology was studied. The famous demonstration at Shipping-port showed that the Th-{sup 233}U fueled, heterogeneous PWR with four different lattice fuels was possible to breed fissile but its low averaged burn-up including blanket fuel and the complicated core configuration were not suitable for economically competitive reactor. The authors investigated the wide design range in terms of fuel cell design, power density, averaged discharge burn-up, etc. to determine the potential of water-cooled Th reactor as a competitive breeder. It is found that a low moderated (MFR=0.3) H{sub 2}O-cooled reactor with comparable burn-up with current LWR is feasible to breed fissile fuel but the core size is too large to be economical because of the low pellet power density. On the other hand, D{sub 2}O-cooled reactor shows relatively wider feasible design window, therefore it is possible to design a core having better neutronic and economic performance than H{sub 2}O-cooled. Both coolant-type cores show negative void reactivity coefficient while achieving breeding capability which is a distinguished characteristics of thorium based fuel breeder reactor. (authors)

  7. Helium Leak Detection of Vessels in Fuel Transfer Cell (FTC) of Prototype Fast Breeder Reactor (PFBR)

    NASA Astrophysics Data System (ADS)

    Dutta, N. G.

    2012-11-01

    Bharatiya Nabhikiya Vidyut Nigam (BHAVINI) is engaged in construction of 500MW Prototype Fast Breeder Reactor (PFBR) at Kalpak am, Chennai. In this very important and prestigious national programme Special Product Division (SPD) of M/s Kay Bouvet Engg.pvt. ltd. (M/s KBEPL) Satara is contributing in a major way by supplying many important sub-assemblies like- Under Water trolley (UWT), Airlocks (PAL, EAL) Container and Storage Rack (CSR) Vessels in Fuel Transfer Cell (FTC) etc for PFBR. SPD of KBEPL caters to the requirements of Government departments like - Department of Atomic Energy (DAE), BARC, Defense, and Government undertakings like NPCIL, BHAVINI, BHEL etc. and other precision Heavy Engg. Industries. SPD is equipped with large size Horizontal Boring Machines, Vertical Boring Machines, Planno milling, Vertical Turret Lathe (VTL) & Radial drilling Machine, different types of welding machines etc. PFBR is 500 MWE sodium cooled pool type reactor in which energy is produced by fissions of mixed oxides of Uranium and Plutonium pellets by fast neutrons and it also breeds uranium by conversion of thorium, put along with fuel rod in the reactor. In the long run, the breeder reactor produces more fuel then it consumes. India has taken the lead to go ahead with Fast Breeder Reactor Programme to produce electricity primarily because India has large reserve of Thorium. To use Thorium as further fuel in future, thorium has to be converted in Uranium by PFBR Technology.

  8. A FAST BREEDER REACTOR SPENT FUEL MEASUREMENTS PROGRAM FOR BN-350 REACTOR

    SciTech Connect

    P. STAPLES; J. HALBIG; ET AL

    1999-04-01

    A project to verify the fissile content of fast breeder reactor spent nuclear fuel is underway in the Republic of Kasakhstan. There are a variety of assembly types with different irradiation histories and profiles in the reactor that require a variety of measurement and analysis procedures. These procedures will be discussed and compared as will the general process that has been designed to resolve any potential measurement discrepancies. The underwater counter is part of a system that is designed to assist the International Atomic Energy Agency (IAEA) in maintaining continuity of knowledge from the time of measurement until the measured item is placed in a welded container with a unique identification. In addition to satisfying IAEA requirements for the spent nuclear fuel, this measurement program is able to satisfy some of the measurement requirements for the Kasakhstan Atomic Energy Agency concerning the repackaging of the spent nuclear fuel into a standard canister. The project is currently operational in a mode requiring the IAEA's continuous presence.

  9. Safety and core design of large liquid-metal cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  10. Use of ferritic steels in breeder reactors worldwide

    SciTech Connect

    Patriarca, P.

    1983-01-01

    The performance of LMFBR reactor steam generator materials is reviewed. Tensile properties of stainless steel-304, stainless steel-316, chromium-molybdenum steels, and Incoloy 800H are presented for elevated temperatures.

  11. End-of-life nondestructive examination of Light Water Breeder Reactor fuel rods (LWBR Development Program)

    SciTech Connect

    Gorscak, D.A.; Campbell, W.R.; Clayton, J.C.

    1987-10-01

    In-bundle and out-of-bundle (single rod) nondestructive examinations of Light Water Breeder Reactor fuel rods were performed. In-bundle examinations included visual examination and measurement of rod bow, rod-to-rod gaps, and rod removal forces. Out-of-bundle examinations included rod visuals and measurement of fuel rod length, diameter and ovality, cladding oxide and crud thickness, support grid induced cladding wear mark depth and volume, and fuel rod free hanging bow. The out-of-bundle examination also included ultrasonic inspection for cladding defects, neutron radiography for pellet integrity and plenum gap measurements, and gamma scans for instack axial gap screening and binary fuel stack length measurements. The measurements confirmed design predictions of fuel rod performance and provided evidence of excellent fuel rod performance for operation of Light Water Breeder Reactor to 29,047 effective full power hours (EFPH).

  12. FUEL-BREEDER FUEL ELEMENT FOR NUCLEAR REACTOR

    DOEpatents

    Abbott, W.E.; Balent, R.

    1958-09-16

    A fuel element design to facilitate breeding reactor fuel is described. The fuel element is comprised of a coatainer, a central core of fertile material in the container, a first bonding material surrounding the core, a sheet of fissionable material immediately surrounding the first bonding material, and a second bonding material surrounding the fissionable material and being in coniact with said container.

  13. Development of DIPRES feed for the fabrication of mixed-oxide fuels for fast breeder reactors

    SciTech Connect

    Griffin, C W; Rasmussen, D E; Lloyd, M H

    1983-01-01

    The DIrect PREss Spheroidized feed process combines the conversion of uranium-plutonium solutions into spheres by internal gelation with conventional pellet fabrication techniques. In this manner, gel spheres could replace conventional powders as the feed material for pellet fabrication of nuclear fuels. Objective of the DIPRES feed program is to develop and qualify a process to produce mixed-oxide fuel pellets from gel spheres for fast breeder reactors. This process development includes both conversion and fabrication activities.

  14. Measurements of thermal-hydraulic parameters in liquid-metal-cooled fast-breeder reactors

    SciTech Connect

    Sackett, J.I.

    1983-01-01

    This paper discusses instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, sodium purity, and leakage. The paper identifies the overall instrumentation requirements for LMFBR's and those aspects of instrumentation which are unique or of special concern to LMFBR systems. It also gives an overview of the status of instrument design and performance.

  15. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    SciTech Connect

    Pope, R B; Diggs, J M

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  16. Packaging and shipment of U. S. breeder reactor experiments

    SciTech Connect

    Berger, J.D.

    1980-01-01

    Irradiation testing of fuels and materials in the Fast Test Reactor (FTR) required development of a shipping cask (designated T-3) and associated hardware for loading and shipping of these experiments to postirradiation examination facilities. The T-3 shipping-cask program included design, fabrication, and testing of internal cask packages to protect the experiments during loading, shipping, and unloading. The cask was designed for loading in both the vertical and horizontal attitudes.

  17. Development of fast breeder reactor fuel reprocessing technology at the Power Reactor and Nuclear Fuel Development Corporation

    SciTech Connect

    Kawata, T.; Takeda, H.; Togashi, A.; Hayashi, S. . Tokai Works); Stradley, J.G. )

    1991-01-01

    For the past two decades, a broad range of research development (R D) programs to establish fast breeder reactor (FBR) system and its associated fuel cycle technology have been pursued by the Power Reactor and Nuclear Fuel Development Corporation (PNC). Developmental activities for FBR fuel reprocessing technology have been primarily conducted at PNC Tokai Works where many important R D facilities for nuclear fuel cycle are located. These include cold and uranium tests for process equipment development in the Engineering Demonstration Facilities (EDF)-I and II, and laboratory-scale hot tests in the Chemical Processing Facility (CPF) where fuel dissolution and solvent extraction characteristics are being investigated with irradiated FBR fuel pins whose burn-up ranges up to 100,000 MWd/t. An extensive effort has also been made at EDF-III to develop advanced remote technology which enables to increase plant availability and to decrease radiation exposures to the workers in future reprocessing plants. The PNC and the United States Department of Energy (USDOE) entered into the joint collaboration in which the US shares the R Ds to support FBR fuel reprocessing program at the PNC. Several important R Ds on advanced process equipment such as a rotary dissolver and a centrifugal contactor system are in progress in a joint effort with the Oak Ridge National Laboratory (ORNL) Consolidated Fuel Reprocessing Program (CFRP). In order to facilitate hot testing on advanced processes and equipment, the design of a new engineering-scale hot test facility is now in progress aiming at the start of hot operation in late 90's. 31 refs., 2 tabs.

  18. Adaptive robust control of the EBR-II reactor

    SciTech Connect

    Power, M.A.; Edwards, R.M.

    1996-05-01

    Simulation results are presented for an adaptive H{sub {infinity}} controller, a fixed H{sub {infinity}} controller, and a classical controller. The controllers are applied to a simulation of the Experimental Breeder Reactor II primary system. The controllers are tested for the best robustness and performance by step-changing the demanded reactor power and by varying the combined uncertainty in initial reactor power and control rod worth. The adaptive H{sub {infinity}} controller shows the fastest settling time, fastest rise time and smallest peak overshoot when compared to the fixed H{sub {infinity}} and classical controllers. This makes for a superior and more robust controller.

  19. Materials development for a fast breeder reactor steam generator concept

    SciTech Connect

    Sessions, C.E.; Reynolds, S.D. Jr.; Hebbar, M.A.; Lewis, J.F.; Kiefer, J.H.

    1981-11-01

    The progress achieved since 1977 in the important area of materials and processes development of fast reactor steam generator development is summarized. The two distinguishing features of the proposed Westinghouse-Tampa steam generator concept are the convoluted shell expansion joint (CSEJ) and the double-wall tubing with a third fluid leak detection capability. A 2/one quarter/ Cr-1 Mo low alloy steel will be used for all important parts of the generator including the CSEJ and the tubes. Other areas in which progress was made include tube-to-tubesheet (T/TS) welding, post-weld heat treatment (PWHT), tube expansion, and development of materials specifications for prototype and future plant materials. 8 refs.

  20. Activation characteristics of a solid breeder blanket for a fusion power demonstration reactor

    NASA Astrophysics Data System (ADS)

    Fischer, Ulrich; Tsige-Tamirat, Haileyesus

    2002-12-01

    Activation characteristics have been assessed for a helium cooled solid breeder blanket on the basis of three-dimensional activation calculations for a 2200 MW fusion power demonstration reactor. FISPACT inventory calculations were performed for the beryllium neutron multiplier, the Li 4SiO 4 breeder ceramics and the Eurofer low activation steel. Neutron flux spectra distributions were provided by a previous MCNP calculation. Detailed spatial distributions have been obtained for the nuclide inventories and related quantities such as activity, decay heat and contact dose rate. These data are available form the authors upon request. On the basis of the calculated contact gamma dose rates, the waste quality was assessed with regard to a possible re-use of the activated materials following the remote or the hands-on handling recycling options.

  1. Special topics reports for the reference tandem mirror fusion breeder. Volume 2. Reactor safety assessment

    SciTech Connect

    Maya, I.; Hoot, C.G.; Wong, C.P.C.; Schultz, K.R.; Garner, J.K.; Bradbury, S.J.; Steele, W.G.; Berwald, D.H.

    1984-09-01

    The safety features of the reference fission suppressed fusion breeder reactor are presented. These include redundancy and overcapacity in primary coolant system components to minimize failure probability, an improved valve location logic to provide for failed component isolation, and double-walled coolant piping and steel guard vessel protection to further limit the extent of any leak. In addition to the primary coolant and decay heat removal system, reactor safety systems also include an independent shield cooling system, the module safety/fuel transfer coolant system, an auxiliary first wall cooling system, a psssive dump tank cooling system based on the use of heat pipes, and several lithium fire suppression systems. Safety system specifications are justified based on the results of thermal analysis, event tree construction, consequence calculations, and risk analysis. The result is a reactor design concept with an acceptably low probability of a major radioactivity release. Dose consequences of maximum credible accidents appear to be below 10CFR100 regulatory limits.

  2. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    SciTech Connect

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  3. Conjugate heat transfer analysis of multiple enclosures in prototype fast breeder reactor

    SciTech Connect

    Velusamy, K.; Balaubramanian, V.; Vaidyanathan, G.; Chetal, S.C.

    1995-09-01

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled reactor under design. The main vessel of the reactor serves as the primary boundary. It is surrounded by a safety vessel which in turn is surrounded by biological shield. The gaps between them are filled with nitrogen. Knowledge of temperature distribution prevailing under various operating conditions is essential for the assessment of structural integrity. Due to the presence of cover gas over sodium free level within the main vessel, there are sharp gradients in temperatures. Also cover gas height reduces during station blackout conditions due to sodium level rise in main vessel caused by temperature rise. This paper describes the model used to analyse the natural convection in nitrogen, conduction in structures and radiation interaction among them. Results obtained from parametric studies for PFBR are also presented.

  4. Design Feasible Area on Water Cooled Thorium Breeder Reactor in Equilibrium States

    SciTech Connect

    Sidik Permana; Naoyuki Takaki; Hiroshi Sekimoto

    2006-07-01

    Thorium as supplied fuel has good candidate for fuel material if it is converted into fissile material {sup 233}U which shows superior characteristics in the thermal region. The Shippingport reactor used {sup 233}U-Th fuel system, and the molten salt breeder reactor (MSBR) project showed that breeding is possible in a thermal spectrum. In the present study, feasibility of water cooled thorium breeder reactor is investigated. The key properties such as flux, {eta} value, criticality and breeding performances are evaluated for different moderator to fuel ratios (MFR) and burn-ups. The results show the feasibility of breeding for different MFR and burn-ups. The required {sup 233}U enrichment is about 2% - 9% as charge fuel. The lower MFR and the higher enrichment of {sup 233}U are preferable to improve the average burn-up; however the design feasible window is shrunk. This core shows the design feasible window especially in relation to MFR with negative void reactivity coefficient. (authors)

  5. Wireless, in-vessel neutron monitor for initial core-loading of advanced breeder reactors

    NASA Technical Reports Server (NTRS)

    Delorenzo, J. T.; Kennedy, E. J.; Blalock, T. V.; Rochelle, J. M.; Chiles, M. M.; Valentine, K. H.

    1981-01-01

    An experimental wireless, in-vessel neutron monitor was developed to measure the reactivity of an advanced breeder reactor as the core is loaded for the first time to preclude an accidental critically incident. The environment is liquid sodium at a temperature of approx. 220 C, with negligible gamma or neutron radiation. With ultrasonic transmission of neutron data, no fundamental limitation was observed after tests at 230 C for 2000 h. The neutron sensitivity was approx. 1 count/s-nv, and the potential data transmission rate was approx. 10,000 counts/s.

  6. Pressure drop considerations of a lithium cooled fusion breeder tokamak reactor blanket

    SciTech Connect

    Wong, C.P.C.

    1983-12-06

    Liquid lithium was selected as one of the coolants for the 1983 fusion breeder blanket used on the magnetically confined tokamak fusion reactor, and as a result, the thermal-hydraulic calculations were dominated by magnetohydrodynamic (MHD) considerations. The applicable sets of MHD equations for the engineering thermal-hydraulic design were reviewed and compared. Special attention was given to the MHD calculations for the fertile material zone, a packed bed of composite beryllium and thorium balls, since this region can dominate the thermal-hydraulic behavior of this blanket module. To keep the pressure drops acceptable, fertile fuel balls were omitted in the inboard blanket.

  7. Review of uncertainty estimates associated with models for assessing the impact of breeder reactor radioactivity releases

    SciTech Connect

    Miller, C.; Little, C.A.

    1982-08-01

    The purpose is to summarize estimates based on currently available data of the uncertainty associated with radiological assessment models. The models being examined herein are those recommended previously for use in breeder reactor assessments. Uncertainty estimates are presented for models of atmospheric and hydrologic transport, terrestrial and aquatic food-chain bioaccumulation, and internal and external dosimetry. Both long-term and short-term release conditions are discussed. The uncertainty estimates presented in this report indicate that, for many sites, generic models and representative parameter values may be used to calculate doses from annual average radionuclide releases when these calculated doses are on the order of one-tenth or less of a relevant dose limit. For short-term, accidental releases, especially those from breeder reactors located in sites dominated by complex terrain and/or coastal meteorology, the uncertainty in the dose calculations may be much larger than an order of magnitude. As a result, it may be necessary to incorporate site-specific information into the dose calculation under these circumstances to reduce this uncertainty. However, even using site-specific information, natural variability and the uncertainties in the dose conversion factor will likely result in an overall uncertainty of greater than an order of magnitude for predictions of dose or concentration in environmental media following shortterm releases.

  8. Cold Trap Dismantling and Sodium Removal at a Fast Breeder Reactor - 12327

    SciTech Connect

    Graf, A.; Petrick, H.; Stutz, U.; Hosking, P.

    2012-07-01

    The first German prototype Fast Breeder Nuclear Reactor (KNK) is currently being dismantled after being the only operating Fast Breeder-type reactor in Germany. As this reactor type used sodium as a coolant in its primary and secondary circuit, seven cold traps containing various amounts of partially activated sodium needed to be disposed of as part of the dismantling. The resulting combined difficulties of radioactive contamination and high chemical reactivity were handled by treating the cold traps differently depending on their size and the amount of sodium contained inside. Six small cold traps were processed onsite by cutting them up into small parts using a band saw under a protective atmosphere. The sodium was then converted to sodium hydroxide by using water. The remaining large cold trap could not be handled in the same way due to its dimensions (2.9 m x 1.1 m) and the declared amount of sodium inside (1,700 kg). It was therefore manually dismantled inside a large box filled with a protective atmosphere, while the resulting pieces were packaged for later burning in a special facility. The experiences gained by KNK during this process may be advantageous for future dismantling projects in similar sodium-cooled reactors worldwide. The dismantling of a prototype fast breeder reactor provides the challenge not only to dismantle radioactive materials but also to handle sodium-contaminated or sodium-containing components. The treatment of sodium requires additional equipment and installations to ensure a safe handling. Since it is not permitted to bring sodium into a repository, all sodium has to be neutralized either through a controlled reaction with water or by incinerating. The resulting components can be disposed of as normal radioactive waste with no further conditions. The handling of sodium needs skilled and experienced workers to minimize the inherent risks. And the example of the disposal of the large KNK cold trap shows the interaction with others and

  9. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    SciTech Connect

    Budd, W.A.

    1986-03-01

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  10. Development of advanced blanket materials for a solid breeder blanket of a fusion reactor

    NASA Astrophysics Data System (ADS)

    Kawamura, H.; Ishitsuka, E.; Tsuchiya, K.; Nakamichi, M.; Uchida, M.; Yamada, H.; Nakamura, K.; Ito, H.; Nakazawa, T.; Takahashi, H.; Tanaka, S.; Yoshida, N.; Kato, S.; Ito, Y.

    2003-08-01

    The design of an advanced solid breeding blanket in a DEMO reactor requires a tritium breeder and a neutron multiplier that can withstand high temperatures and high neutron fluences, and the development of such advanced blanket materials has been carried out by collaboration between JAERI, universities and industries in Japan. The Li2TiO3 pebble fabricated by a wet process is a reference material as a tritium breeder, but its stability at high temperatures has to be improved for its application in a DEMO blanket. One of these improved materials, TiO2-doped Li2TiO3 pebbles, was successfully fabricated and studied. For the advanced neutron multiplier, beryllides that have a high melting point and good chemical stability have been studied. Some characterization of Be12Ti was conducted, and it became clear that it had lower swelling and tritium inventory than beryllium metal. Pebble fabrication study for Be12Ti was also performed and Be12Ti pebbles were successfully fabricated. These activities have shown that there is a bright prospect in realizing a DEMO blanket by the application of TiO2-doped Li2TiO3 and beryllides.

  11. Tritium trapping in silicon carbide in contact with solid breeder under high flux isotope reactor irradiation

    SciTech Connect

    H. Katsui; Y. Katoh; A. Hasegawa; M. Shimada; Y. Hatano; T. Hinoki; S. Nogami; T. Tanaka; S. Nagata; T. Shikama

    2013-11-01

    The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C ß-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ~3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation.

  12. Fusion breeder

    SciTech Connect

    Moir, R.W.

    1982-04-20

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the US fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the US fusion program and the US nuclear energy program. The purpose of this paper is to suggest this policy change be made and tell why it should be made, and to outline specific research and development goals so that the fusion breeder will be developed in time to meet fissile fuel needs.

  13. Statement of the Executive Committee of the Scientists' Institute for Public Information--Comments on the Breeder Reactor

    ERIC Educational Resources Information Center

    Environment, 1975

    1975-01-01

    Inheritance of the Atomic Energy Commission's Draft Environmental Impact Statement for the Liquid Metal Fast Breeder Reactor Program by the Energy Research and Development Administration has caused much concern among members of the Scientists' Institute for Public Information (SIPI). SIPI members are concerned about the inadequacy and economic…

  14. Recommendations concerning research and model evaluation needs to support breeder reactor environmental radiological assessments

    SciTech Connect

    Miller, C. W.; Dunning, Jr., D. E.; Etnier, E. L.; Kocher, D. C.; McDowell-Boyer, L. M.; Meyer, H. R.; Rohwer, P. S.

    1980-12-01

    Purpose of this report is to present recommendations concerning needs for model evaluations, environmental research, and biomedical research to support breeder reactor environmental radiological assessments. More data are needed to specify dry deposition velocities and to validate plume depletion models. More atmospheric dispersion data are required to characterize flow near buildings, in complex terrain, and for travel distances at 100 km or more. Field data are needed for terrestrial food chain transport models, especially those used to assess the impact of acute radionuclide releases. Efforts are needed to develop models for the estimation of dose from external exposure to photons from a finite, elevated plume resulting from an acute radionuclide release to the atmosphere. Estimates of doses to man from internally deposited radionuclides require scrutiny. Further study of tritium is needed to determine its dependence on dose and dose rate and to specify the relative toxicity of various physiochemical forms of tritium in the environment.

  15. Clinch River Breeder Reactor Plant Steam Generator Few Tube Test model post-test examination

    SciTech Connect

    Impellezzeri, J.R.; Camaret, T.L.; Friske, W.H.

    1981-03-11

    The Steam Generator Few Tube Test (FTT) was part of an extensive testing program carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. This paper describes the problems encountered with the mechanical features of the FTT model design which led to premature test termination, and the results of the post-test examination. Conditions of tube bowing and significant tube and tube support gouging was observed. An interpretation of the visual and metallurgical observations is also presented. The CRBRP steam generator has undergone design evaluations to resolve observed deficiences found in the FFTM.

  16. Recovery of tritium dissolved in sodium at the steam generator of fast breeder reactor

    SciTech Connect

    Oya, Y.; Oda, T.; Tanaka, S.; Okuno, K.

    2008-07-15

    The tritium recovery technique in steam generators for fast breeder reactors using the double pipe concept was proposed. The experimental system for developing an effective tritium recovery technique was developed and tritium recovery experiments using Ar gas or Ar gas with 10-10000 ppm oxygen gas were performed using D{sub 2} gas instead of tritium gas. It was found that deuterium permeation through two membranes decreased by installing the double pipe concept with Ar gas. By introducing Ar gas with 10000 ppm oxygen gas, the concentration of deuterium permeation through two membranes decreased by more than 1/200, compared with the one pipe concept, indicating that most of the deuterium was scavenged by Ar gas or reacted with oxygen to form a hydroxide. However, most of the hydroxide was trapped at the surface of the membranes because of the short duration of the experiment. (authors)

  17. Uncertainty estimates for predictions of the impact of breeder-reactor radionuclide releases

    SciTech Connect

    Miller, C.W.; Little, C.A.

    1982-01-01

    This paper summarizes estimates, compiled in a larger report, of the uncertainty associated with models and parameters used to assess the impact on man radionuclide releases to the environment by breeder reactor facilities. These estimates indicate that, for many sites, generic models and representative parameter values may reasonably be used to calculate doses from annual average radionuclide releases when these calculated doses are on the order of one-tenth or less of a relevant dose limit. For short-term, accidental releases, the uncertainty in the dose calculations may be much larger than an order of magnitude. As a result, it may be necessary to incorporate site-specific information into the dose calculation under such circumstances. However, even using site-specific information, inherent natural variability within human receptors, and the uncertainties in the dose conversion factor will likely result in an overall uncertainty of greater than an order of magnitude for predictions of dose following short-term releases.

  18. Coated ceramic breeder materials

    DOEpatents

    Tam, Shiu-Wing; Johnson, Carl E.

    1987-04-07

    A breeder material for use in a breeder blanket of a nuclear reactor is disclosed. The breeder material comprises a core material of lithium containing ceramic particles which has been coated with a neutron multiplier such as Be or BeO, which coating has a higher thermal conductivity than the core material.

  19. Recommended practices in elevated temperature design: A compendium of breeder reactor experiences (1970-1986): An overview

    SciTech Connect

    Wei, B.C.; Cooper, W.L. Jr.; Dhalla, A.K.

    1987-09-01

    Significant experiences have been accumulated in the establishment of design methods and criteria applicable to the design of Liquid Metal Fast Breeder Reactor (LMFBR) components. The Subcommittee of the Elevated Temperature Design under the Pressure Vessel Research Council (PVRC) has undertaken to collect, on an international basis, design experience gained, and the lessons learned, to provide guidelines for next generation advanced reactor designs. This paper shall present an overview and describe the highlights of the work.

  20. Primary disassembly of Light Water Breeder Reactor modules for core evaluation (LWBR Development Program)

    SciTech Connect

    Greenberger, R.J.; Miller, E.L.

    1987-10-01

    After successfully operating for 29,047 effective full power hours, the Light Water Breeder Reactor (LWBR) core was defueled prior to total decommissioning of the Shippingport Atomic Power Station. All nuclear fuel and much of the reactor internal hardware was removed from the reactor vessel. Non-fuel components were prepared for shipment to disposal sites, and the fuel assemblies were partially disassembled and shipped to the Expended Core Facility (ECF) in Idaho. At ECF, the fuel modules underwent further disassembly to provide fuel rods for nondestructive testing to establish the core's breeding efficiency and to provide core components for examinations to assess their performance characteristics. This report presents a basic description of the processes and equipment used to disassemble LWBR fuel modules for subsequent proof-of-breeding (POB) and core examination operations. Included are discussions of module handling fixtures and equipment, the underwater milling machine and bandsaw assemblies, and the associated design and operation of this equipment for LWBR fuel module disassembly.

  1. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    NASA Astrophysics Data System (ADS)

    Harto, Andang Widi

    2012-06-01

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  2. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    SciTech Connect

    Harto, Andang Widi

    2012-06-06

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  3. Protected air-cooled condenser for the Clinch River Breeder Reactor Plant

    SciTech Connect

    Louison, R.; Boardman, C.E.

    1981-05-29

    The long term residual heat removal for the Clinch River Breeder Reactor Plant (CRBRP) is accomplished through the use of three protected air-cooled condensers (PACC's) each rated at 15M/sub t/ following a normal or emergency shutdown of the reactor. Steam is condensed by forcing air over the finned and coiled condenser tubes located above the steam drums. The steam flow is by natural convection. It is drawn to the PACC tube bundle for the steam drum by the lower pressure region in the tube bundle created from the condensing action. The concept of the tube bundle employs a unique patented configuration which has been commercially available through CONSECO Inc. of Medfore, Wisconsin. The concept provides semi-parallel flow that minimizes subcooling and reduces steam/condensate flow instabilities that have been observed on other similar heat transfer equipment such as moisture separator reheaters (MSRS). The improved flow stability will reduce temperature cycling and associated mechanical fatigue. The PACC is being designed to operate during and following the design basis earthquake, depressurization from the design basis tornado and is housed in protective building enclosure which is also designed to withstand the above mentioned events.

  4. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    SciTech Connect

    Slater, C.O.; Reed, D.A.; Cramer, S.N.; Emmett, M.B.; Tomlinson, E.T.

    1981-01-01

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design.

  5. Effect of Lithium Enrichment on the Tritium Breeding Characteristics of Various Breeders in a Fusion Driven Hybrid Reactor

    NASA Astrophysics Data System (ADS)

    Übeyli, Mustafa

    2009-09-01

    Selection of lithium containing materials is very important in the design of a deuterium-tritium (DT) fusion driven hybrid reactor in order to supply its tritium self-sufficiency. Tritium, an artificial isotope of hydrogen, can be produced in the blanket by using the neutron capture reactions of lithium in the coolants and/or blanket materials which consist of lithium. This study presents the effect of lithium-6 enrichment in the coolant of the reactor on the tritium breeding of the hybrid blanket. Various liquid-solid breeder couples were investigated to determine the effective breeders. Numerical results pointed out that the tritium production increased with increasing lithium-6 enrichment for all cases.

  6. Breeder Reprocessing Engineering Test

    SciTech Connect

    Burgess, C.A.; Meacham, S.A.

    1984-01-01

    The Breeder Reprocessing Engineering Test (BRET) is a developmental activity of the US Department of Energy to demonstrate breeder fuel reprocessing technology while closing the fuel cycle for the Fast Flux Test Facility (FFTF). It will be installed in the existing Fuels and Materials Examination Facility (FMEF) at the Hanford Site near Richland, Washington, The major objectives of BRET are: (1) close the US breeder fuel cycle; (2) develop and demonstrate reprocessing technology and systems for breeder fuel; (3) provide an integrated test of breeder reactor fuel cycle technology - rprocessing, safeguards, and waste management. BRET is a joint effort between the Westinghouse Hanford Company and Oak Ridge National Laboratory. 3 references, 2 figures.

  7. Multiple lead seal assembly for a liquid-metal-cooled fast-breeder nuclear reactor

    DOEpatents

    Hutter, Ernest; Pardini, John A.

    1977-03-15

    A reusable multiple lead seal assembly provides leak-free passage of stainless-steel-clad instrument leads through the cover on the primary tank of a liquid-metal-cooled fast-breeder nuclear reactor. The seal isolates radioactive argon cover gas and sodium vapor within the primary tank from the exterior atmosphere and permits reuse of the assembly and the stainless-steel-clad instrument leads. Leads are placed in flutes in a seal body, and a seal shell is then placed around the seal body. Circumferential channels in the body and inner surface of the shell are contiguous and together form a conduit which intersects each of the flutes, placing them in communication with a port through the wall of the seal shell. Liquid silicone rubber sealant is injected into the flutes through the port and conduit; the sealant fills the space in the flutes not occupied by the leads themselves and dries to a rubbery hardness. A nut, threaded onto a portion of the seal body not covered by the seal shell, jacks the body out of the shell and shears the sealant without damage to the body, shell, or leads. The leads may then be removed from the body. The sheared sealant is cleaned from the body, leads, and shell and the assembly may then be reused with the same or different leads.

  8. Ceramics for fusion reactors: The role of the lithium orthosilicate as breeder

    NASA Astrophysics Data System (ADS)

    Carella, Elisabetta; Hernández, Teresa

    2012-11-01

    Lithium-based oxide ceramics are studied as breeder blanket materials for the controlled thermonuclear reactors (CTR). Lithium orthosilicate (Li4SiO4) is one of the most promising candidates because of its lithium concentration (0.54 g/cm3), its high melting temperature (1523 K) and its excellent tritium release behavior. It is reported that the diffusion of tritium is closely related to that of lithium, so it is possible to find an indirect measure of the trend of tritium studying the diffusivity of Li+. In the present work, the synthesis of the Li4SiO4 is carried out by Spray drying followed by pyrolysis. The study of the Li+ ion diffusion on the sintered bodies, is investigated by means of electrical conductivity measurements. The effect of the γ-ray irradiation is evaluated by the impedance spectroscopy method (EIS) from room temperature to 1173 K. The results indicate that the síntesis process employed can produce Li4SiO4 in the form of pebbles, finally the best ion species for the electrical conduction is the Li+ and is shown that the g-irradiation to a dose of 5MGy, facilitate its mobility through the creation of defects, without change in its conduction process.

  9. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    SciTech Connect

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  10. Decontamination of liquid-metal fast breeder reactor components for reuse; The French experience

    SciTech Connect

    Michaille, P. ); Moroni, J.C. ); Lambert, I. )

    1991-02-01

    Decontamination of stainless steel liquid-metal fast breeder reactor components for reuse in France began with the decontamination of Rapsodie components. At that time, dilute phosphoric acid was used. To cope with additional irradiated components after Phenix came into operation, an extensive study was performed, which led to the selection of a procedure involving two baths. The first bath, alkaline permanganate (AP), is applied for 3 h; the second bath, sulfo-phosphoric acid (SP), is applied for 6 h, both at 60{degrees}C. Up to three cycles are repeated until the residual dose rate is sufficiently low. Eight intermediate heat exchangers (IHXs) and two primary pumps from Phenix were decontaminated using this method. This paper reports that because SP can pickle only a limited depth ({approximately} 3{mu}m), due to the passivation effect of phosphoric acid, and because of the waste treatment problems associated with phosphates, new solutions were explored. One possibility involves improvement of the AP-SP procedure: In the SPm procedure, the AP bath is omitted and the phosphoric concentration is reduced by a factor of 4. A second approach is the use of a new formula, called SECA, a mixture of maleic and citric acid used in reducing conditions (imposed by hydrazine). Since the Phenix and Superphenix waste treatment facilities are not designed to reprocess maleic-citric acid, only the SPm procedure has been used on reactor components. A low-contaminated IHX from Rapsodie served as a test benchmark, not only for the decontamination procedure, but also for the requalification criteria, before the SPm procedure was applied to a highly contaminated IHX from Phenix. Recent results are presented.

  11. Atmospheric Dispersion of Sodium Aerosol due to a Sodium Leak in a Fast Breeder Reactor Complex

    NASA Astrophysics Data System (ADS)

    Punitha, G.; Sudha, A. Jasmin; Kasinathan, N.; Rajan, M.

    Liquid sodium at high temperatures (470 K to 825 K) is used as the primary and secondary coolant in Liquid Metal cooled Fast Breeder Reactors (LMFBR). In the event of a postulated sodium leak in the Steam Generator Building (SGB) of a LMFBR, sodium readily combusts in the ambient air, especially at temperatures above 523 K. Intense sodium fire results and sodium oxide fumes are released as sodium aerosols. Sodium oxides are readily converted to sodium hydroxide in air due to the presence of moisture in it. Hence, sodium aerosols are invariably in the form of particulate sodium hydroxide. These aerosols damage not only the equipment and instruments due to their corrosive nature but also pose health hazard to humans. Hence, it is essential to estimate the concentration of sodium aerosols within the plant boundary for a sodium leak event. The Gaussian Plume Dispersion Model can obtain the atmospheric dispersion of sodium aerosols in an open terrain. However, this model does not give accurate results for dispersion in spaces close to the point of release and with buildings in between. The velocity field due to the wind is altered to a large extent by the intervening buildings and structures. Therefore, a detailed 3-D estimation of the velocity field and concentration has to be obtained through rigorous computational fluid dynamics (CFD) approach. PHOENICS code has been employed to determine concentration of sodium aerosols at various distances from the point of release. The dispersion studies have been carried out for the release of sodium aerosols at different elevations from the ground and for different wind directions.

  12. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    SciTech Connect

    Malathi, N.; Sahoo, P. Ananthanarayanan, R.; Murali, N.

    2015-02-15

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control and Safety Rod Drive Mechanism during reactor operation.

  13. Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

    NASA Astrophysics Data System (ADS)

    Malathi, N.; Sahoo, P.; Ananthanarayanan, R.; Murali, N.

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ˜100 Hz/mm, ˜1 s, and ˜0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation.

  14. Level monitoring system with pulsating sensor--application to online level monitoring of dashpots in a fast breeder reactor.

    PubMed

    Malathi, N; Sahoo, P; Ananthanarayanan, R; Murali, N

    2015-02-01

    An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control & Safety Rod Drive Mechanism during reactor operation. PMID:25725884

  15. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    SciTech Connect

    Not Available

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  16. Verification of the Plant Dynamics Analytical Code CERES Using the Results of the Plant Trip Test of the Prototype Fast Breeder Reactor MONJU

    SciTech Connect

    Yoshihisa Nishi; Nobuyuki Ueda; Izumi Kinoshita; Akira Miyakawa; Mitsuya Kato

    2006-07-01

    CERES is plant system analysis code for LMRs (liquid metal cooled reactors) developed by the Central Research Institute of Electric Power Industry (CRIEPI). To verify the CERES code, analyses were performed by using the result of the plant trip test of the prototype FBR (fast breeder reactor) 'MONJU' at 40% rated power. The verification work was performed as a joint research of CRIEPI and JAEA (Japan Atomic Energy Agency). Following three verification analyses were performed mainly. (I) Analysis concerning the primary/ secondary/auxiliary cooling system (the plenum in the reactor vessel (R/V) was modeled in R-Z 2-dimension). (II) Analysis concerning the thermal-hydraulic characteristics in the plenum of R/V (the plenum was modeled in 3-dimension). (III) Analysis concerning the flow characteristics inside the intermediate heat exchanger (IHX) (the plenum in the IHX was modeled in 3-dimension). Analytical results by the CERES code showed good agreement with the results of the test of the 'MONJU'. Fundamental abilities of the CERES as a plant dynamics calculation code had been verified through these analyses. Additionally, some characteristic flows in plenums of 'MONJU' became clear by these analyses. (authors)

  17. The long-term future for civilian nuclear power generation in France: The case for breeder reactors. Breeder reactors: The physical and physical chemistry parameters, associate material thermodynamics and mechanical engineering: Novelties and issues

    NASA Astrophysics Data System (ADS)

    Dautray, Robert

    2011-06-01

    The author firstly gives a summary overview of the knowledge base acquired since the first breeder reactors became operational in the 1950s. "Neutronics", thermal phenomena, reactor core cooling, various coolants used and envisioned for this function, fuel fabrication from separated materials, main equipment (pumps, valves, taps, waste cock, safety circuits, heat exchange units, etc.) have now attained maturity, sufficient to implement sodium cooling circuits. Notwithstanding, the use of metallic sodium still raises certain severe questions in terms of safe handling (i.e. inflammability) and other important security considerations. The structural components, both inside the reactor core and outside (i.e. heat exchange devices) are undergoing in-depth research so as to last longer. The fuel cycle, notably the refabrication of fuel elements and fertile elements, the case of transuranic elements, etc., call for studies into radiation induced phenomena, chemistry separation, separate or otherwise treatments for materials that have different radioactive, physical, thermodynamical, chemical and biological properties. The concerns that surround the definitive disposal of certain radioactive wastes could be qualitatively improved with respect to the pressurized water reactors (PWRs) in service today. Lastly, the author notes that breeder reactors eliminate the need for an isotope separation facility, and this constitutes a significant contribution to contain nuclear proliferation. Among the priorities for a fully operational system (power station - the fuel cycle - operation-maintenance - the spent fuel pool and its cooling system-emergency cooling system-emergency electric power-transportation movements-equipment handling - final disposal of radioactive matter, independent safety barriers), the author includes materials (fabrication of targets, an irradiation and inspection instrument), the chemistry of all sorting processes, equipment "refabrication" or rehabilitation

  18. Study of safeguards system on dry reprocessing for fast breeder reactor

    SciTech Connect

    Li, T. K.; Burr, Tom; Menlove, Howard O.; Thomas, K. E.; Fukushima, M.; Hori, M.

    2002-01-01

    A 'Feasibility Study on the Commercialized Fast Breeder Reactor (FBR) Cycle System' is underway at Japan Nuclear Cycle Development Institute (JNC). Concepts to commercialize the FBR fuel cycle are being created together with their necessary research and development (R&D) tasks. 'Dry,' non-aqueous, processes are candidates for FBR fuel reprocessing. Dry reprocessing technology takes advantage of proliferation barriers, due to the lower decontamination factors achievable by the simple pyrochemical processes proposed. The concentration o f highly radioactive impurities and non-fissile materials in products from a dry reprocess is generally significantly larger than the normal aqueous (Purex) process. However, the safeguards of dry reprocesses have not been widely analyzed. In 2000, JNC and Los Alamos National Laboratoiy (LANL) initiated a joint research program to study the safeguards aspects of dry reprocessing. In this study, the safeguardability of the three options: metal electrorefining, oxide electrowinning, and fluoride volatility processes, are assessed. FBR spent fuels are decladded and powdered into mixed oxides (MOX) at the Head-End process either by oxidation-reduction reactions (metal electrorefining and fluoride volatility) or mechanically (oxide electrowinning). At the oxide electrowinning process, the spent MOX he1 powder is transferred to chloride in molten salt and nuclear materials are extracted onto cathode as oxides. For metal electrorefining process, on the other hand, the MOX fuel is converted to chloride in molten salt, and nuclear materials are extracted onto cathode as a metal fomi. At lhe fluoride volatility process, the MOX fuel powder is converted to U&/PuF6 (gaseous form) in a fluidized bed; plutonium and uranium fluorides are separated by volatilization properties and then are converted to oxides. Since the conceptual design of a dry reprocessing plant is incomplete, the operational mode, vessel capacities, residence times, and campaigns

  19. Ceramic breeder materials

    SciTech Connect

    Johnson, C.E.

    1990-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 32 refs., 1 fig., 1 tab.

  20. Topaz-II reactor control unit development

    SciTech Connect

    Wyant, F.J.; Jensen, D.; Logothetis, J.

    1994-12-31

    The development for a new digital reactor control unit for the Topaz-II reactor is described. The unit is expected to provide the means for automated control during a possible Topaz flight experiment. The breadboard design and development is discussed.

  1. Modeling and analysis of the unprotected loss-of-flow accident in the Clinch River Breeder Reactor

    SciTech Connect

    Morris, E.E.; Dunn, F.E.; Simms, R.; Gruber, E.E.

    1985-01-01

    The influence of fission-gas-driven fuel compaction on the energetics resulting from a loss-of-flow accident was estimated with the aid of the SAS3D accident analysis code. The analysis was carried out as part of the Clinch River Breeder Reactor licensing process. The TREAT tests L6, L7, and R8 were analyzed to assist in the modeling of fuel motion and the effects of plenum fission-gas release on coolant and clad dynamics. Special, conservative modeling was introduced to evaluate the effect of fission-gas pressure on the motion of the upper fuel pin segment following disruption. For the nominal sodium-void worth, fission-gas-driven fuel compaction did not adversely affect the outcome of the transient. When uncertainties in the sodium-void worth were considered, however, it was found that if fuel compaction occurs, loss-of-flow driven transient overpower phenomenology could not be precluded.

  2. Computerized operating procedures for shearing and dissolution of segments from LWBR (Light Water Breeder Reactor) fuel rods

    SciTech Connect

    Osudar, J.; Deeken, P.G.; Graczyk, D.G.; Fagan, J.E.; Martino, F.J.; Parks, J.E.; Levitz, N.M.; Kessie, R.W.; Leddin, J.M.

    1987-05-01

    This report presents two detailed computerized operating procedures developed to assist and control the shearing and dissolution of irradiated fuel rods. The procedures were employed in the destructive analysis of end-of-life fuel rods from the Light Water Breeder Reactor (LWBR) that was designed by the Westinghouse Electric Corporation Bettis Atomic Power Laboratory. Seventeen entire fuel rods from the end-of-life core of the LWBR were sheared into 169 precisely characterized segments, and more than 150 of these segments were dissolved during execution of the LWBR Proof-of-Breeding (LWBR-POB) Analytical Support Project at Argonne National Laboratory. The procedures illustrate our approaches to process monitoring, data reduction, and quality assurance during the LWBR-POB work.

  3. Carbon transport in a bimetallic sodium loop simulating the intermediate heat transport system of a liquid metal fast breeder reactor

    SciTech Connect

    Hampton, L.V.; Spalaris, C.N.; Roy, P.

    1980-04-01

    Carbon transport data from a bimetallic sodium loop simulating the intermediate heat transport system of a Liquid Metal Fast Breeder Reactor are discussed. The results of bulk carbon analyses after 15,000 hours' exposure indicate a pattern of carburization of Type 304 stainless steel foils which is independent of loop sodium temperature. A model based on carbon activity gradients accounting for this behavior is proposed. Data also indicate that carburization of Type 304 stainless steel is a diffusion-controlled process; however, decarburization of the ferritic 2 1/4 Cr-1Mo steel is not. It is proposed that the decarburization of the ferritic steel is controlled by the dissolution of carbides in the steel matrix. The differences in the sodium decarburization behavior of electroslag remelted and vacuum-arc remelted 2 1/4 Cr-1Mo steel are also highlighted.

  4. Water Flow Simulation Test on Flow-Induced Oscillation of Thermowell in Prototype Fast Breeder Reactor “MONJU”

    NASA Astrophysics Data System (ADS)

    Kondo, Masaya; Anoda, Yoshinari

    Water flow simulation tests were performed on the flow-induced oscillations of the thermowell in the prototype fast breeder reactor (FBR), MONJU. The displacements of the target cylinder were measured, and the oscillation amplitudes, the frequency characteristics, and the phase relationships were estimated. The estimations showed that the oscillations of the target cylinder had a one-dimensional oscillation region in the in-line direction with symmetric vortices shedding and a two-dimensional oscillation region induced by alternative vortices. The phase estimation, carried out by a methodology using wavelet analysis and statistical analysis, indicated that the effect of the alternative vortices on the in-line oscillation was changed with the flow velocity.

  5. The effect of coolant orificing on the core performance of a heterogeneous liquid-metal fast breeder reactor

    SciTech Connect

    Mamoru, K.; Shigehiro, A.; Yoshiaki, O.

    1983-04-01

    The effect of orificing on the core performance of a commercial-size heterogeneous liquid-metal fast breeder reactor was studied analytically. The thermal power output was flattened at beginning of life, and the coolant flow rate was chosen such that the maximum inner cladding temperature of a driver fuel and a blanket fuel was less than or equal to 620/sup 0/C at both beginning of equilibrium life (BOEL) and end of equilibrium life (EOEL). The difference between reactor outlet temperatures at BOEL and EOEL was then calculated for six core configurations: one homogeneous core configuration and five heterogeneous ones. The results showed that the core outlet temperature variation due to the change of the power profile of the radial heterogeneous core configurations is similar to that of the homogeneous one, even when a single type of orificing is used in each core zone, and it will not be necessary to use the more detailed orificing in each zone of a heterogeneous core configuration. The study concludes that for the present design, especially the thermal design, of some heterogeneous core configurations, it is feasible to control the change of the reactor outlet temperature with burnup, even when a single type of orificing is used in each core zone.

  6. Comparison of oxide- and metal-core behavior during CRBRP (Clinch River Breeder Reactor Plant) station blackout

    SciTech Connect

    Polkinghorne, S T; Atkinson, S A

    1986-01-01

    A resurrected concept that could significantly improve the inherently safe response of Liquid-Metal cooled Reactors (LMRs) during severe undercooling transients is the use of metallic fuel. Analytical studies have been reported on for the transient behavior of metal-fuel cores in innovative, inherently safe LMR designs. This paper reports on an analysis done, instead, for the Clinch River Breeder Reactor Plant (CRBRP) design with the only innovative change being the incorporation of a metal-fuel core. The SSC-L code was used to simulate a protected station blackout accident in the CRBRP with a 943 MWt Integral Fast Reactor (IFR) metal-fuel core. The results, compared with those for the oxide-fueled CRBRP, show that the margin to boiling is greater for the IFR core. However, the cooldown transient is more severe due to the faster thermal response time of metallic fuel. Some additional calculations to assess possible LMR design improvements (reduced primary system pressure losses, extended flow coastdown) are also discussed. 8 refs., 13 figs., 2 tabs.

  7. Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

    SciTech Connect

    McCulloch, R.W.; Post, D.W.; Lovell, R.T.; Snyder, S.D.

    1981-04-01

    Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin simulators used in test loops by the Breeder Reactor Program Thermal-Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor-Core Flow Test Loop. Thermal, mechanical, and electrical design considerations are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relate this profile to that generated by the coils in completed fuel pin simulators.

  8. Advanced In-Service Inspection Approaches Applied to the Phenix Fast Breeder Reactor

    SciTech Connect

    Guidez, J.; Martin, L.; Dupraz, R.

    2006-07-01

    The safety upgrading of the Phenix plant undertaken between 1994 and 1997 involved a vast inspection programme of the reactor, the external storage drum and the secondary sodium circuits in order to meet the requirements of the defence-in-depth safety approach. The three lines of defence were analysed for every safety related component: demonstration of the quality of design and construction, appropriate in-service inspection and controlling the consequences of an accident. The in-service reactor block inspection programme consisted in controlling the core support structures and the high-temperature elements. Despite the fact that limited consideration had been given to inspection constraints during the design stage of the reactor in the 1960's, as compared to more recent reactor projects such as the European Fast Reactor (EFR), all the core support line elements were able to be inspected. The three following main operations are described: Ultrasonic inspection of the upper hangers of the main vessel, using small transducers able to withstand temperatures of 130 deg. C, Inspection of the conical shell supporting the core dia-grid. A specific ultrasonic method and a special implementation technique were used to control the under sodium structure welds, located up to several meters away from the scan surface. Remote inspection of the hot pool structures, particularly the core cover plug after partial sodium drainage of the reactor vessel. Other inspections are also summarized: control of secondary sodium circuit piping, intermediate heat exchangers, primary sodium pumps, steam generator units and external storage drum. The pool type reactor concept, developed in France since the 1960's, presents several favourable safety and operational features. The feedback from the Phenix plant also shows real potential for in-service inspection. The design of future generation IV sodium fast reactors will benefit from the experience acquired from the Phenix plant. (authors)

  9. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    SciTech Connect

    Not Available

    1980-09-01

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases.

  10. Simulation of Radioactive Corrosion Product in Primary Cooling System of Japanese Sodium-Cooled Fast Breeder Reactor

    NASA Astrophysics Data System (ADS)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    Radioactive Corrosion Product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. The most important CP is 54Mn and 60Co. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE incorporating the Particle Model. Moreover, among the percentage of total radioactive deposition accounted for by CP in particle form, 54Mn was estimated to constitute approximately 20 % and 60Co approximately 40 % in the cold-leg region. These calculation results are consistent with the measured results for the actual cold-leg piping in the JOYO.

  11. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    SciTech Connect

    John D. Bess

    2010-03-01

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these

  12. Automated operator procedure prompting for startup of Experimental Breeder Reactor-2

    SciTech Connect

    Renshaw, A.W.; Ball, S.J.; Ford, C.E.

    1990-11-01

    This report describes the development of an operator procedure prompting aid for startup of a nuclear reactor. This operator aid is a preliminary design for a similar aid that eventually will be used with the Advanced Liquid Metal Reactor (ALMR) presently in the design stage. Two approaches were used to develop this operator procedure prompting aid. One method uses an expert system software shell, and the other method uses database software. The preliminary requirements strongly pointed toward features traditionally associated with both database and expert systems software. Database software usually provides data manipulation flexibility and user interface tools, and expert systems tools offer sophisticated data representation and reasoning capabilities. Both methods, including software and associated hardware, are described in this report. Proposals for future enhancements to improve the expert system approach to procedure prompting and for developing other operator aids are also offered. 25 refs., 14 figs.

  13. Optimization of a heterogeneous fast breeder reactor core with improved behavior during unprotected transients

    SciTech Connect

    Poumerouly, S.; Schmitt, D.; Massara, S.; Maliverney, B.

    2012-07-01

    Innovative Sodium-cooled Fast Reactors (SFRs) are currently being investigated by CEA, AREVA and EDF in the framework of a joint French collaboration, and the construction of a GEN IV prototype, ASTRID (Advanced Sodium Technical Reactor for Industrial Demonstration), is scheduled in the years 2020. Significant improvements are expected so as to improve the reactor safety: the goal is to achieve a robust safety demonstration of the mastering of the consequences of a Core Disruptive Accident (CDA), whether by means of prevention or mitigation features. In this framework, an innovative design was proposed by CEA in 2010. It aims at strongly reducing the sodium void effect, thereby improving the core behavior during unprotected loss of coolant transients. This design is strongly heterogeneous and includes, amongst others, a fertile plate, a sodium plenum associated with a B{sub 4}C upper blanket and a stepwise modulation of the fissile height of the core (onwards referred to as the 'diabolo shape'). In this paper, studies which were entirely carried out at EDF are presented: the full potential of this heterogeneous concept is thoroughly investigated using the SDDS methodology. (authors)

  14. The fusion breeder

    NASA Astrophysics Data System (ADS)

    Moir, Ralph W.

    1982-10-01

    The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the U.S. fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the U.S. fusion program and the U.S. nuclear energy program. There is wide agreement that many approaches will work and will produce fuel for five equal-sized LWRs, and some approach as many as 20 LWRs at electricity costs within 20% of those at today's price of uranium (30/lb of U3O8). The blankets designed to suppress fissioning, called symbiotes, fusion fuel factories, or just fusion breeders, will have safety characteristics more like pure fusion reactors and will support as many as 15 equal power LWRs. The blankets designed to maximize fast fission of fertile material will have safety characteristics more like fission reactors and will support 5 LWRs. This author strongly recommends development of the fission suppressed blanket type, a point of view not agreed upon by everyone. There is, however, wide agreement that, to meet the market price for uranium which would result in LWR electricity within 20% of today's cost with either blanket type, fusion components can cost severalfold more than would be allowed for pure fusion to meet the goal of making electricity alone at 20% over today's fission costs. Also widely agreed is that the critical-path-item for the fusion breeder is fusion development itself; however, development of fusion breeder specific items (blankets, fuel cycle) should be started now in order to have the fusion breeder by the time the rise in uranium prices forces other more costly choices.

  15. TOKOPS: Tokamak Reactor Operations Study: The influence of reactor operations on the design and performance of tokamaks with solid-breeder blankets: Final report

    SciTech Connect

    Conn, R.W.; Ghoniem, N.M.; Firestone, M.A.

    1986-09-01

    Reactor system operation and procedures have a profound impact on the conception and design of power plants. These issues are studied here using a model tokamak system employing a solid-breeder blanket. The model blanket is one which has evolved from the STARFIRE and BCSS studies. The reactor parameters are similar to those characterizing near-term fusion engineering reactors such as INTOR or NET (Next European Tokamak). Plasma startup, burn analysis, and methods for operation at various levels of output power are studied. A critical, and complicating, element is found to be the self-consistent electromagnetic response of the system, including the presence of the blanket and the resulting forces and loadings. Fractional power operation, and the strategy for burn control, is found to vary depending on the scaling law for energy confinement, and an extensive study is reported. Full-power reactor operation is at a neutron wall loading pf 5 MW/m/sup 2/ and a surface heat flux of 1 MW/m/sup 2/. The blanket is a pressurized steel module with bare beryllium rods and low-activation HT-9-(9-C-) clad LiAlO/sub 2/ rods. The helium coolant pressure is 5 MPa, entering the module at 297/sup 0/C and exiting at 550/sup 0/C. The system power output is rated at 1000 MW(e). In this report, we present our findings on various operational scenarios and their impact on system design. We first start with the salient aspects of operational physics. Time-dependent analyses of the blanket and balance of plant are then presented. Separate abstracts are included for each chapter.

  16. Microstructure analysis for chemical interaction between cesium and SUS 316 steel in fast breeder reactor application

    SciTech Connect

    Sasaki, K.; Fukumoto, K. I.; Oshima, T.; Tanigaki, T.; Masayoshi, U.

    2012-07-01

    In this study the corrosion products on a surface after cesium corrosion examination at 650 deg. C for 100 hrs were characterized by TEM observation around the corroded area on the surface in order to understand the corrosion mechanism of cesium fission product for cladding materials in fast reactor. The experimental results suggest the main corrosion mechanism occurred in the process of the separation of cesium chromate and metal (Fe, Ni). The main reaction of corrosion process was considered to be equation, 2Cs + 7/2 O{sub 2} + 2Cr {yields} Cs{sub 2}Cr{sub 2}O{sub 7}(L). (authors)

  17. Core loading pattern optimization of thorium fueled heavy water breeder reactor using genetic algorithm

    SciTech Connect

    Soewono, C. N.; Takaki, N.

    2012-07-01

    In this work genetic algorithm was proposed to solve fuel loading pattern optimization problem in thorium fueled heavy water reactor. The objective function of optimization was to maximize the conversion ratio and minimize power peaking factor. Those objectives were simultaneously optimized using non-dominated Pareto-based population ranking optimal method. Members of non-dominated population were assigned selection probabilities based on their rankings in a manner similar to Baker's single criterion ranking selection procedure. A selected non-dominated member was bred through simple mutation or one-point crossover process to produce a new member. The genetic algorithm program was developed in FORTRAN 90 while neutronic calculation and analysis was done by COREBN code, a module of core burn-up calculation for SRAC. (authors)

  18. Fault tree analysis of the EBR-II reactor shutdown system

    SciTech Connect

    Kamal, S.A.; Hill, D.J.

    1992-01-01

    As part of the level I Probabilistic Risk Assessment of the Experimental Breeder Reactor II (EBR-II), detailed fault trees for the reactor shutdown system are developed. Fault tree analysis is performed for two classes of transient events that are of particular importance to EBR-II operation: loss-of-flow and transient-overpower. In all parts of EBR-II reactor shutdown system, redundancy has been utilized in order to reduce scram failure probability. Therefore, heavy emphasis is placed in the fault trees on the common cause failures (CCFs) among similar mechanical components of the control and safety rods and among similar electrical components in redundant detection channels and shutdown strings. Generic beta-factors that cover all types of similar components and reflect redundancy level are used to model the CCFs. Human errors are addressed in the fault trees in two major areas: errors that would prevent the automatic scram channels from detecting the abnormal events and errors that would prevent utilization of the manual scram capability. The fault tree analysis of the EBR-II shutdown system has provided not only a systematic process for calculating the probabilities of system failures but also useful insights into the system and how its elements interact during transient events that require shutdown.

  19. Fault tree analysis of the EBR-II reactor shutdown system

    SciTech Connect

    Kamal, S.A.; Hill, D.J.

    1992-12-01

    As part of the level I Probabilistic Risk Assessment of the Experimental Breeder Reactor II (EBR-II), detailed fault trees for the reactor shutdown system are developed. Fault tree analysis is performed for two classes of transient events that are of particular importance to EBR-II operation: loss-of-flow and transient-overpower. In all parts of EBR-II reactor shutdown system, redundancy has been utilized in order to reduce scram failure probability. Therefore, heavy emphasis is placed in the fault trees on the common cause failures (CCFs) among similar mechanical components of the control and safety rods and among similar electrical components in redundant detection channels and shutdown strings. Generic beta-factors that cover all types of similar components and reflect redundancy level are used to model the CCFs. Human errors are addressed in the fault trees in two major areas: errors that would prevent the automatic scram channels from detecting the abnormal events and errors that would prevent utilization of the manual scram capability. The fault tree analysis of the EBR-II shutdown system has provided not only a systematic process for calculating the probabilities of system failures but also useful insights into the system and how its elements interact during transient events that require shutdown.

  20. Kalkar nuclear power plant (SNR-300) - A sodium-cooled fast breeder reactor prototype

    SciTech Connect

    Morgenstern, F.H.

    1987-09-01

    The status of the Kalkar nuclear power plant in early summer 1986 is that, apart from later alterations to the workshop building, the assembly and non-nuclear commissioning work has practically been completed. From a technical point of view, nuclear commissioning of the plant can begin, but vital factors for this are the necessary nuclear licenses. The most important licensing prerequisites have been fulfilled;all essential appraisals have been available since January/February 1986. At the beginning of April 1986, the Reactor Safety Commission and the Radiation Protection Commission cast a positive vote for initial fuel loading. Before the accident in Chernobyl, but particularly since then, the issuing of the licenses has come under the political pressure of the commencing election campaign phase for the federal elections in January 1987. The initial project definition phase, the organizational boundary conditions, and the major requirements for the construction of the plant are summarized in chronological form. To provide the total picture, references dealing with general and technical aspects of the project are listed.

  1. Materials accounting in a fast-breeder-reactor fuels-reprocessing facility: optimal allocation of measurement uncertainties

    SciTech Connect

    Dayem, H.A.; Ostenak, C.A.; Gutmacher, R.G.; Kern, E.A.; Markin, J.T.; Martinez, D.P.; Thomas, C.C. Jr.

    1982-07-01

    This report describes the conceptual design of a materials accounting system for the feed preparation and chemical separations processes of a fast breeder reactor spent-fuel reprocessing facility. For the proposed accounting system, optimization techniques are used to calculate instrument measurement uncertainties that meet four different accounting performance goals while minimizing the total development cost of instrument systems. We identify instruments that require development to meet performance goals and measurement uncertainty components that dominate the materials balance variance. Materials accounting in the feed preparation process is complicated by large in-process inventories and spent-fuel assembly inputs that are difficult to measure. To meet 8 kg of plutonium abrupt and 40 kg of plutonium protracted loss-detection goals, materials accounting in the chemical separations process requires: process tank volume and concentration measurements having a precision less than or equal to 1%; accountability and plutonium sample tank volume measurements having a precision less than or equal to 0.3%, a shortterm correlated error less than or equal to 0.04%, and a long-term correlated error less than or equal to 0.04%; and accountability and plutonium sample tank concentration measurements having a precision less than or equal to 0.4%, a short-term correlated error less than or equal to 0.1%, and a long-term correlated error less than or equal to 0.05%. The effects of process design on materials accounting are identified. Major areas of concern include the voloxidizer, the continuous dissolver, and the accountability tank.

  2. Cost/performance comparison between pulse columns and centrifugal contactors designed to process Clinch River Breeder Reactor fuel

    SciTech Connect

    Ciucci, J.A. Jr.

    1983-12-01

    A comparison between pulse columns and centrifugal contactors was made to determine which type of equipment was more advantageous for use in the primary decontamination cycle of a remotely operated fuel reprocessing plant. Clinch River Breeder Reactor (CRBR) fuel was chosen as the fuel to be processed in the proposed 1 metric tonne/day reprocessing facility. The pulse columns and centrifugal contactors were compared on a performance and total cost basis. From this comparison, either the pulse columns or the centrifugal contactors will be recommended for use in a fuel reprocessing plant built to reprocess CRBR fuel. The reliability, solvent exposure to radiation, required time to reach steady state, and the total costs were the primary areas of concern for the comparison. The pulse column units were determined to be more reliable than the centrifugal contactors. When a centrifugal contactor motor fails, it can be remotely changed in less than one eight hour shift. Pulse columns expose the solvent to approximately five times as much radiation dose as the centrifugal contactor units; however, the proposed solvent recovery system adequately cleans the solvent for either case. The time required for pulse columns to reach steady state is many times longer than the time required for centrifugal contactors to reach steady state. The cost comparison between the two types of contacting equipment resulted in centrifugal contactors costing 85% of the total cost of pulse columns when the contactors were stacked on three levels in the module. If the centrifugal contactors were all positioned on the top level of a module with the unoccupied volume in the module occupied by other equipment, the centrifugal contactors cost is 66% of the total cost of pulse columns. Based on these results, centrifugal contactors are recommended for use in a remotely operated reprocessing plant built to reprocess CRBR fuel.

  3. Investigations on natural circulation in reactor models and shutdown heat removal systems for LMFBRs (liquid metal fast breeder reactors)

    SciTech Connect

    Hoffmann, H.; Weinberg, D.; Marten, K. ); Ieda, Yoshiaki )

    1989-11-01

    For sodium-cooled pool-type reactors, studies have been undertaken to remove the decay heat by natural convection alone, as in the case of failure of all power supplies. For this purpose, four immersion coolers (ICs), two each installed at a 180-deg circumferential position with respect to the others, are arranged within the reactor tank. They are connected with natural-drift air coolers through independent intermediate circuits. The primary sodium in the tank as well as the secondary sodium in the intermediate loop circulate by natural convection. The general functioning of this passive shutdown decay heat removal (DHR) system is demonstrated in 1:20 and 1:5 scale test models using water as a simulant fluid for sodium. The model design is based on the thermohydraulics similarity criteria. In the RAMONA three-dimensional 1:20 scale model, experiments were carried out to clarify the steady-state in-vessel thermohydraulics for different parameter combinations (core power, radial power distribution across the core, DHR by 2 or 4 ICs in operation, above-core structure geometry and position, different IC designs). For all mentioned parameters, temperatures and their fluctuations were measured and used to indicate isotherms and lines of identical temperature fluctuations. The flow patterns were observed visually. The experiments were recalculated by an updated version of the single-phase three-dimensional thermohydraulics code COMMIX.

  4. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    SciTech Connect

    Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.; Werner, James E.

    2006-01-20

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  5. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    NASA Astrophysics Data System (ADS)

    Morton, Sheryl L.; Baily, Carl E.; Hill, Thomas J.; Werner, James E.

    2006-01-01

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a low-temperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  6. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    SciTech Connect

    Sheryl Morton; Carl Baily; Tom Hill; Jim Werner

    2006-02-01

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a lowtemperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  7. Development of a Fast Breeder Reactor Fuel Bundle-Duct Interaction Analysis Code - BAMBOO: Analysis Model and Validation by the Out-of-Pile Compression Test

    SciTech Connect

    Uwaba, Tomoyuki; Tanaka, Kosuke

    2001-10-15

    To analyze the wire-wrapped fast breeder reactor (FBR) fuel pin bundle deformation under bundle-duct interaction (BDI) conditions, the Japan Nuclear Cycle Development Institute has developed the BAMBOO computer code. A three-dimensional beam element model is used in this code to calculate fuel pin bowing and cladding oval distortion, which are the dominant deformation mechanisms in a fuel pin bundle. In this work, the property of the cladding oval distortion considering the wire-pitch was evaluated experimentally and introduced in the code analysis.The BAMBOO code was validated in this study by using an out-of-pile bundle compression testing apparatus and comparing these results with the code results. It is concluded that BAMBOO reasonably predicts the pin-to-duct clearances in the compression tests by treating the cladding oval distortion as the suppression mechanism to BDI.

  8. Study on laser welding of fuel clad tubes and end plugs made of modified 9Cr-1Mo steel for metallic fuel of Fast Breeder Reactors

    NASA Astrophysics Data System (ADS)

    Harinath, Y. V.; Gopal, K. A.; Murugan, S.; Albert, S. K.

    2013-04-01

    A procedure for Pulsed Laser Beam Welding (PLBW) has been developed for fabrication of fuel pins made of modified 9Cr-1Mo steel for metallic fuel proposed to be used in future in India's Fast Breeder Reactor (FBR) programme. Initial welding trials of the samples were carried out with different average power using Nd-YAG based PLBW process. After analyzing the welds, average power for the weld was optimized for the required depth of penetration and weld quality. Subsequently, keeping the average power constant, the effect of various other welding parameters like laser peak power, pulse frequency, pulse duration and energy per pulse on weld joint integrity were studied and a procedure that would ensure welds of acceptable quality with required depth of penetration, minimum size of fusion zone and Heat Affected Zone (HAZ) were finalized. This procedure is also found to reduce the volume fraction delta-ferrite in the fusion zone.

  9. Progress in the R and D Project on Oxide Dispersion Strengthened and Precipitation Hardened Ferritic Steels for Sodium Cooled Fast Breeder Reactor Fuels

    SciTech Connect

    Kaito, Takeji; Ohtsuka, Satoshi; Inoue, Masaki

    2007-07-01

    High burnup capability of sodium cooled fast breeder reactor (SFR) fuels depends significantly on irradiation performance of their component materials. Japan Atomic Energy Agency (JAEA) has been developing oxide dispersion strengthened (ODS) ferritic steels and a precipitation hardened (PH) ferritic steel as the most prospective materials for fuel pin cladding and duct tubes, respectively. Technology for small-scale manufacturing is already established, and several hundreds of ODS steel cladding tubes and dozens of PH steel duct tubes were successfully produced. We will step forward to develop manufacturing technology for mass production to supply these steels for future SFR fuels. Mechanical properties of the products were examined by out-of-pile and in-pile tests including material irradiation tests in the experimental fast reactor JOYO and foreign fast reactors. The material strength standards (MSSs) were tentatively compiled in 2005 for ODS steels and in 1993 for PH steel. In order to upgrade the MSSs and to demonstrate high burnup capability of the materials, we will perform a series of irradiation tests in BOR-60 and JOYO until 2015 and contribute to design study for a demonstration SFR of which operation is expected after 2025. (authors)

  10. Ceramic breeder materials

    SciTech Connect

    Johnson, C.E.; Kummerer, K.R.; Roth, E.

    1987-01-01

    Ceramic materials are under investigation as potential breeder material in fusion reactors. This paper will review candidate materials with respect to fabrication routes and characterization, properties in as-fabricated and irradiated condition, and experimental results from laboratory and inpile investigations on tritium transport and release. Also discussed are the resources of beryllium, which is being considered as a neutron multiplier. The comparison of ceramic properties that is attempted here aims at the identification of the most-promising material for use in a tritium breeding blanket. 82 refs., 12 figs., 5 tabs.

  11. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  12. Preparation of pyrolytic carbon coating on graphite for inhibiting liquid fluoride salt and Xe135 penetration for molten salt breeder reactor

    NASA Astrophysics Data System (ADS)

    Song, Jinliang; Zhao, Yanling; He, Xiujie; Zhang, Baoliang; Xu, Li; He, Zhoutong; Zhang, DongSheng; Gao, Lina; Xia, Huihao; Zhou, Xingtai; Huai, Ping; Bai, Shuo

    2015-01-01

    A fixed-bed deposition method was used to prepare rough laminar pyrolytic carbon coating (RLPyC) on graphite for inhibiting liquid fluoride salt and Xe135 penetration during use in molten salt breeder reactor. The RLPyC coating possessed a graphitization degree of 44% and had good contact with graphite substrate. A high-pressure reactor was constructed to evaluate the molten salt infiltration in the isostatic graphite (IG-110, TOYO TANSO CO., LTD.) and RLPyC coated graphite under 1.01, 1.52, 3.04, 5.07 and 10.13 × 105 Pa for 12 h. Mercury injection and molten-salt infiltration experiments indicated the porosity and the salt-infiltration amount of 18.4% and 13.5 wt% under 1.52 × 105 Pa of IG-110, which was much less than 1.2% and 0.06 wt% under 10.13 × 105 Pa of the RLPyC, respectively. A vacuum device was constructed to evaluate the Xe135 penetration in the graphite. The helium diffusion coefficient of RLPyC coated graphite was 2.16 × 10-12 m2/s, much less than 1.21 × 10-6 m2/s of the graphite. Thermal cycle experiment indicated the coatings possessed excellent thermal stability. The coated graphite could effectively inhibit the liquid fluoride salt and Xe135 penetration.

  13. Validation Work to Support the Idaho National Engineering and Environmental Laboratory Calculational Burnup Methodology Using Shippingport Light Water Breeder Reactor (LWBR) Spent Fuel Assay Data

    SciTech Connect

    J. W. Sterbentz

    1999-08-01

    Six uranium isotopes and fourteen fission product isotopes were calculated on a mass basis at end-of-life (EOL) conditions for three fuel rods from different Light Water Breeder Reactor (LWBR) measurements. The three fuel rods evaluated here were taken from an LWBR seed module, a standard blanket module, and a reflector (Type IV) module. The calculated results were derived using a depletion methodology previously employed to evaluate many of the radionuclide inventories for spent nuclear fuels at the Idaho National Engineering and Environmental Laboratory. The primary goal of the calculational task was to further support the validation of this particular calculational methodology and its application to diverse reactor types and fuels. Result comparisons between the calculated and measured mass concentrations in the three rods indicate good agreement for the three major uranium isotopes (U-233, U-234, U-235) with differences of less than 20%. For the seed and standard blanket rod, the U-233 and U-234 differences were within 5% of the measured values (these two isotopes alone represent greater than 97% of the EOL total uranium mass). For the major krypton and xenon fission product isotopes, differences of less than 20% and less than 30% were observed, respectively. In general, good agreement was obtained for nearly all the measured isotopes. For these isotopes exhibiting significant differences, possible explanations are discussed in terms of measurement uncertainty, complex transmutations, etc.

  14. Fusion Breeder Program interim report

    SciTech Connect

    Moir, R.; Lee, J.D.; Neef, W.

    1982-06-11

    This interim report for the FY82 Fusion Breeder Program covers work performed during the scoping phase of the study, December, 1981-February 1982. The goals for the FY82 study are the identification and development of a reference blanket concept using the fission suppression concept and the definition of a development plan to further the fusion breeder application. The context of the study is the tandem mirror reactor, but emphasis is placed upon blanket engineering. A tokamak driver and blanket concept will be selected and studied in more detail during FY83.

  15. Development of a Fast Breeder Reactor Fuel Bundle Deformation Analysis Code - BAMBOO: Development of a Pin Dispersion Model and Verification by the Out-of-Pile Compression Test

    SciTech Connect

    Uwaba, Tomoyuki; Ito, Masahiro; Ukai, Shigeharu

    2004-02-15

    To analyze the wire-wrapped fast breeder reactor fuel pin bundle deformation under bundle/duct interaction conditions, the Japan Nuclear Cycle Development Institute has developed the BAMBOO computer code. This code uses the three-dimensional beam element to calculate fuel pin bowing and cladding oval distortion as the primary deformation mechanisms in a fuel pin bundle. The pin dispersion, which is disarrangement of pins in a bundle and would occur during irradiation, was modeled in this code to evaluate its effect on bundle deformation. By applying the contact analysis method commonly used in the finite element method, this model considers the contact conditions at various axial positions as well as the nodal points and can analyze the irregular arrangement of fuel pins with the deviation of the wire configuration.The dispersion model was introduced in the BAMBOO code and verified by using the results of the out-of-pile compression test of the bundle, where the dispersion was caused by the deviation of the wire position. And the effect of the dispersion on the bundle deformation was evaluated based on the analysis results of the code.

  16. Assessment of potential impact of the Clinch River Breeder Reactor Plant thermal effluent on the Watts Bar Reservoir striped bass population

    SciTech Connect

    Heuer, J H; McIntosh, D; Ostrowski, P; Tomljanovich, D A

    1983-11-01

    This report is an assessment of potential adverse impact to striped bass (Morone saxatilis) in Watts Bar Reservoir caused by thermal effluent from operation of the Clinch River Breeder Reactor Plant (CRBRP). The Clinch River arm of Watts Bar Reservoir is occupied by adult striped bass during the warmest months of the year. Concern was raised that operation of the CRBRP, specifically thermal discharges, could conflict with management of striped bass. In all cases examined the thermal plume becomes nearly imperceptible within a short distance from the discharge pipe (about 30 ft (10 m)) compared to river width (about 630 ft (190 m)). Under worst case conditions any presence of the plume in the main channel (opposite side of the river from the discharge) will be confined to the surface layer of the water. An ample portion of river cross sections containing ambient temperature water for passage or residence of adult striped bass will always be available in the vicinity of this thermal effluent. Although a small portion of river cross section would exceed the thermal tolerance of striped bass, the fish would naturally avoid this area and seek out adjacent cooler water. Therefore, it is concluded the CRBRP thermal effluent will not significantly affect the integrity of the striped bass thermal refuge in the Clinch River arm of Watts Bar Reservoir. At this time there is no need to consider alternative diffuser designs and thermal modeling. 8 references, 3 figures, 2 tables.

  17. RELAP5/MOD3 Analysis of Transient Steam-Generator Behavior During Turbine Trip Test of a Prototype Fast Breeder Reactor MONJU

    SciTech Connect

    Yoshihisa Shindo; Hiroshi Endo; Tomoko Ishizu; Kazuo Haga

    2006-07-01

    In order to develop a thermal-hydraulic model of the steam-generator (SG) to simulate transient phenomena in the sodium cooled fast breeder reactor (FBR) MONJU, Japan Nuclear Energy Safety Organization (JNES) verified the SG model using the RELAP5/MOD3 code against the results of the turbine trip test at a 40% power load of MONJU. The modeling by using RELAP5 was considered to explain the significant observed behaviors of the pressure and the temperature of the EV steam outlet, and the temperature of water supply distributing piping till 600 seconds after the turbine trip. The analysis results of these behaviors showed good agreement with the test results based on results of parameter study as the blow efficiency (release coef.) and heat transferred from the helical coil region to the down-comer (temperature heating down-comer tubes). It was found that the RELAP5/MOD3 code with a two-fluids model can predict well the physical situation: the gas-phase of steam generated by the decompression boiling moves upward in the down-comer tubes accompanied by the enthalpy increase of the water supply chambers; and that the pressure change of a 'shoulder' like shape is induced by the mass balance between the steam mass generated in the down-comer tubes and the steam mass blown from the SG. The applicability of RELAP5/MOD3 to SG modeling was confirmed by simulating the actual FBR system. (authors)

  18. Preliminary study on nano- and micro-composite sol-gel based alumina coatings on structural components of lead-bismuth eutectic cooled fast breeder reactors

    NASA Astrophysics Data System (ADS)

    Dou, Peng; Kasada, Ryuta

    2011-02-01

    In order to protect the structural components of lead-bismuth eutectic cooled fast breeder reactors from liquid metal corrosion, Al 2O 3 nano- and micro-composite coatings were developed using an improved sol-gel process, which includes dipping specimens in a sol-gel solution dispersed with fine α-Al 2O 3 powders prepared by mechanical milling. Accelerated corrosion tests were conducted on coated specimens in liquid lead-bismuth eutectic at 500 °C under dynamic conditions. Scanning electron microscopy (SEM) and X-ray diffraction (XRD) analyses revealed that the coatings are composed of α-Al 2O 3 and they are about 10 μm thick. After the corrosion tests, no spallation occurred on the coatings, and neither Pb nor Bi penetrated into the coatings, which indicates that the coatings possess an enhanced dynamic LBE corrosion resistance to lead-bismuth eutectic corrosion. The nano-structured composite particles integrated into the coatings play an important role in achieving such superior lead-bismuth eutectic corrosion resistance.

  19. Post-scram Liquid Metal cooled Fast Breeder Reactor (LMFBR) neat transport system dynamics and steam generator control

    NASA Astrophysics Data System (ADS)

    Brukx, J. F. L. M.

    1982-06-01

    Loop type LMFBR heat transport system dynamics after reactor shutdown and during subsequent decay heat removal are considered with emphasis on steam generator dynamics including the development and evaluation of various post-scram steam generator control systems, and natural circulation of the sodium coolant, including the influence of superimposed free convection on forced convection heat transfer and pressure drop. The normal operating and decay heat removal functions of the overall heat transport system are described.

  20. Superphenix: Is the fast breeder dream over -- or over yonder?

    SciTech Connect

    1997-03-01

    A detailed history of France`s Superphenix commercial fast breeder reactor project is presented. Important project milestones are discussed from the project`s conception in 1971 to its current status. Recommendations of the Castaing Commission on the project and future plans for use of the reactor are outlined. In addition, world wide fast breeder projects are listed and discussed.

  1. The TOPAZ II space reactor response under accident conditions

    SciTech Connect

    Voss, S.S.

    1993-12-31

    The TOPAZ II is a single-cell thermionic space reactor power system developed by the Russians during the period of time from {approximately}1969 to 1989. The TOPAZ II has never been flight demonstrated, but the system was extensively tested on the ground. As part of the development and test program, the response of the TOPAZ II under accident conditions was analyzed and characterized. The US TOPAZ II team has been working closely with the Russian specialists to understand the TOPAZ II system, its operational characteristics, and its response under potential accident conditions. The purpose of the technical exchange is to enable a potential launch of a TOPAZ II by the US. The information is required to integrate the system with a US spacecraft and to support the safety review process. The purpose of this paper is to provide a brief overview of the system and its response under actual and postulated accident conditions.

  2. Sodium fast reactor safety and licensing research plan. Volume II.

    SciTech Connect

    Ludewig, H.; Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A.; Phillips, J.; Zeyen, R.; Clement, B.; Garner, Frank; Walters, Leon; Wright, Steve; Ott, Larry J.; Suo-Anttila, Ahti Jorma; Denning, Richard; Ohshima, Hiroyuki; Ohno, S.; Miyhara, S.; Yacout, Abdellatif; Farmer, M.; Wade, D.; Grandy, C.; Schmidt, R.; Cahalen, J.; Olivier, Tara Jean; Budnitz, R.; Tobita, Yoshiharu; Serre, Frederic; Natesan, Ken; Carbajo, Juan J.; Jeong, Hae-Yong; Wigeland, Roald; Corradini, Michael; Thomas, Justin; Wei, Tom; Sofu, Tanju; Flanagan, George F.; Bari, R.; Porter D.; Lambert, J.; Hayes, S.; Sackett, J.; Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  3. Summary of the radiological assessment of the fuel cycle for a thorium-uranium carbide-fueled fast breeder reactor

    SciTech Connect

    Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Meyer, H.R.; Morse, L.E.; Till, J.E.; Yalcintas, M.G.

    1980-01-01

    A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

  4. Automated breeder fuel fabrication

    SciTech Connect

    Goldmann, L.H.; Frederickson, J.R.

    1983-09-01

    The objective of the Secure Automated Fabrication (SAF) Project is to develop remotely operated equipment for the processing and manufacturing of breeder reactor fuel pins. The SAF line will be installed in the Fuels and Materials Examination Facility (FMEF). The FMEF is presently under construction at the Department of Energy's (DOE) Hanford site near Richland, Washington, and is operated by the Westinghouse Hanford Company (WHC). The fabrication and support systems of the SAF line are designed for computer-controlled operation from a centralized control room. Remote and automated fuel fabriction operations will result in: reduced radiation exposure to workers; enhanced safeguards; improved product quality; near real-time accountability, and increased productivity. The present schedule calls for installation of SAF line equipment in the FMEF beginning in 1984, with qualifying runs starting in 1986 and production commencing in 1987. 5 figures.

  5. US LMFBR (Liquid Metal Fast Breeder Reactor): flow induced vibration program (1977-1986): A summary and overview

    SciTech Connect

    Wambsganss, M.W.; Chen, S.S.; Mulcahy, T.M.; Jendrzejczyk, J.A.

    1986-09-01

    This paper summarizes the activities and accomplishments under the US LMFBR Flow Induced Vibration Program for the period 1977-1986. Since 1977 represents the date of the last IAEA IWGFR Specialists Meeting on LMFBR Flow Induced Vibration, this paper thus provides an update to the results presented at that meeting. This period also represents a period of substantial change for the US LMFBR program. A major reactor project, the FFTF, was completed and a second major project, the CRBR plant, was terminated. This change adversely impacted the US flow induced vibration program. Nevertheless, base technology activities have continued. In this paper, research in the following areas is summarized: Vibration characteristics and scaling, Turbulent buffeting and vortex shedding, Fluidelastic instabilities of tube bundles in crossflow, and Instabilities induced by leakage flows.

  6. Embedded computer systems for control applications in EBR-II

    SciTech Connect

    Carlson, R.B.; Start, S.E.

    1993-03-01

    The purpose of this paper is to describe the embedded computer systems approach taken at Experimental Breeder Reactor II (EBR-II) for non-safety related systems. The hardware and software structures for typical embedded systems are presented The embedded systems development process is described. Three examples are given which illustrate typical embedded computer applications in EBR-II.

  7. Embedded computer systems for control applications in EBR-II

    SciTech Connect

    Carlson, R.B.; Start, S.E.

    1993-01-01

    The purpose of this paper is to describe the embedded computer systems approach taken at Experimental Breeder Reactor II (EBR-II) for non-safety related systems. The hardware and software structures for typical embedded systems are presented The embedded systems development process is described. Three examples are given which illustrate typical embedded computer applications in EBR-II.

  8. Thermal breeder fuel enrichment zoning

    DOEpatents

    Capossela, Harry J.; Dwyer, Joseph R.; Luce, Robert G.; McCoy, Daniel F.; Merriman, Floyd C.

    1992-01-01

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  9. Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 2: Application to EBR-II Primary Sodium System and Related Systems

    SciTech Connect

    Steven R. Sherman; Collin J. Knight

    2006-03-01

    Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/or to comply with decontamination and decomissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidifed carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, USA. This report is Part 2 of a two-part report. This second report provides a supplement to the first report and describes the application of the humdidified carbon dioxide technique ("carbonation") to the EBR-II primary tank, primary cover gas systems, and the intermediate heat exchanger. Future treatment plans are also provided.

  10. Neutron flux spectra and radiation damage parameters for the Russian Bor-60 and SM-2 reactors

    SciTech Connect

    Karasiov, A.V.; Greenwood, L.R.

    1995-04-01

    The objective is to compare neutron irradiation conditions in Russian reactors and similar US facilities. Neutron fluence and spectral information and calculated radiation damage parameters are presented for the BOR-60 (Fast Experimental Reactor - 60 MW) and SM-2 reactors in Russia. Their neutron exposure characteristics are comparable with those of the Experimental Breeder Reactor (ERB-II), the Fast Flux Test Facility (FFTF), and the High Flux Isotope Reactor (HFIR) in the United States.

  11. Second generation Research Reactor Fuel Container (RRFC-II).

    SciTech Connect

    Abhold, M. E.; Baker, M. C.; Bourret, S. C.; Harker, W. C.; Pelowitz, D. G.; Polk, P. J.

    2001-01-01

    The second generation Research Reactor Fuel Counter (RRFC-II) has been developed to measure the remaining {sup 235}U content in foreign spent Material Test Reactor (MTR)-type fuel being returned to the Westinghouse Savannah River Site (WSRS) for interim storage and subsequent disposal. The fuel to be measured started as fresh fuel nominally with 93% enriched Uraniuin alloyed with A1 clad in Al. The fuel was irradiated to levels of up to 65% burnup. The RRFC-II, which will be located in the L-Basin spent fuel pool, is intended to assay the {sup 235}U content using a combination of passive neutron coincidence counting, active neutron coincidence counting, and active-multiplicity analysis. Measurements will be done underwater, eliminating the need for costly and hazardous handling operations of spent fuel out of water. The underwater portion of the RRFC-II consists of a watertight stainless steel housing containing neutron and gamma detectors and a scanning active neutron source. The portion of the system that resides above water consists of data-processing electronics; electromechanical drive electronics; a computer to control the operation of the counter, to collect, and to analyze data; and a touch screen interface located at the equipment rack. The RRFC-II is an improved version of the Los Alamos-designed RRFC already installed in the SRS Receipts Basin for Offsite Fuel. The RRFC-II has been fabricated and is scheduled for installation in late FY 2001 pending acceptance testing by Savannah River Site personnel.

  12. BESAFE II: Accident safety analysis code for MFE reactor designs

    NASA Astrophysics Data System (ADS)

    Sevigny, Lawrence Michael

    The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications

  13. Fabrication, properties, and tritium recovery from solid breeder materials

    SciTech Connect

    Johnson, C.E. ); Kondo, T. ); Roux, N. ); Tanaka, S. ); Vollath, D. )

    1991-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Experimental Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 133 refs., 1 fig.

  14. Technical Information on the Carbonation of the EBR-II Reactor, Summary Report Part 1: Laboratory Experiments and Application to EBR-II Secondary Sodium System

    SciTech Connect

    Steven R. Sherman

    2005-04-01

    Residual sodium is defined as sodium metal that remains behind in pipes, vessels, and tanks after the bulk sodium metal has been melted and drained from such components. The residual sodium has the same chemical properties as bulk sodium, and differs from bulk sodium only in the thickness of the sodium deposit. Typically, sodium is considered residual when the thickness of the deposit is less than 5-6 cm. This residual sodium must be removed or deactivated when a pipe, vessel, system, or entire reactor is permanently taken out of service, in order to make the component or system safer and/or to comply with decommissioning regulations. As an alternative to the established residual sodium deactivation techniques (steam-and-nitrogen, wet vapor nitrogen, etc.), a technique involving the use of moisture and carbon dioxide has been developed. With this technique, sodium metal is converted into sodium bicarbonate by reacting it with humid carbon dioxide. Hydrogen is emitted as a by-product. This technique was first developed in the laboratory by exposing sodium samples to humidified carbon dioxide under controlled conditions, and then demonstrated on a larger scale by treating residual sodium within the Experimental Breeder Reactor II (EBR-II) secondary cooling system, followed by the primary cooling system, respectively. The EBR-II facility is located at the Idaho National Laboratory (INL) in southeastern Idaho, U.S.A. This report is Part 1 of a two-part report. It is divided into three sections. The first section describes the chemistry of carbon dioxide-water-sodium reactions. The second section covers the laboratory experiments that were conducted in order to develop the residual sodium deactivation process. The third section discusses the application of the deactivation process to the treatment of residual sodium within the EBR-II secondary sodium cooling system. Part 2 of the report, under separate cover, describes the application of the technique to residual sodium

  15. Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    SciTech Connect

    Shott, Gregory

    2014-08-31

    The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

  16. A preliminary investigation of the Topaz II reactor as a lunar surface power supply

    SciTech Connect

    Polansky, G.F.; Houts, M.G.

    1995-12-31

    Reactor power supplies offer many attractive characteristics for lunar surface applications. The Topaz II reactor resulted from an extensive development program in the former Soviet Union. Flight quality reactor units remain from this program and are currently under evaluation in the United States. This paper examines the potential for applying the Topaz II, originally developed to provide spacecraft power, as a lunar surface power supply.

  17. Code System to Calculate Mixed Cores in TRIGA Mark II Research Reactor.

    Energy Science and Technology Software Center (ESTSC)

    2001-08-29

    Version 00 TRIGLAV is a computer program for reactor calculations of mixed cores in a TRIGA Mark II research reactor. It can be applied for fuel element burn-up calculations, for power and flux distributions calculations and for reactivity predictions. The TRIGLAV program requires the WIMS-D4 program with the original WIMS cross-section library extended for TRIGA reactor specific nuclides. This package includes the code TRIGAC, which is a new version of TRIGAP.

  18. Activation analysis of the PULSAR-II fusion power reactor

    SciTech Connect

    Khater, H.Y.

    1995-12-31

    The PULSAR-II pulsed tokamak power plant design utilizes a blanket made of the vanadium alloy, V-5Cr-5Ti, and cooled with liquid lithium. The shield is made of a mixture of the low activation austenitic steel (Tenelon) and vanadium. The blanket is assumed to be replaced every 5.6 full power years (FPY) and the shield is assumed to stay in place for 30 FPY. The activity induced in the blanket at the end of its lifetime is higher than the activity induced in the shield after 30 FPY. At shutdown, the blanket and shield activities are 2678 MCi and 1747 MCi, respectively. One year after shutdown the shield activity drops to 18 MCi compared to 84 MCi for the blanket. The total decay heat generated in the blanket at the end of its lifetime is 34.7 MW and drops to 17.6 MW within an hour. At shutdown, 25.3 MW of decay heat are generated in the shield, dropping to only 0.1 MW within the first year. One week after shutdown, the values of the integrated decay heat are 1770 GJ for the blanket and 469 GJ for the shield. The radwaste classification of the reactor structure is evaluated according to both the NRC 10CFR61 and Fetter waste disposal concentration limits. After 5.6 years of irradiation, the blanket will only qualify for Class C low level waste. After 30 years of operation, the shield will also qualify for disposal as Class C waste. Only remote maintenance will be allowed inside the containment building.

  19. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    SciTech Connect

    Not Available

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

  20. Laser fusion driven breeder design study. Final report

    SciTech Connect

    Berwald, D.H.; Massey, J.V.

    1980-12-01

    The results of the Laser Fusion Breeder Design Study are given. This information primarily relates to the conceptual design of an inertial confinement fusion (ICF) breeder reactor (or fusion-fission hybrid) based upon the HYLIFE liquid metal wall protection concept developed at Lawrence Livermore National Laboratory. The blanket design for this breeder is optimized to both reduce fissions and maximize the production of fissile fuel for subsequent use in conventional light water reactors (LWRs). When the suppressed fission blanket is compared with its fast fission counterparts, a minimal fission rate in the blanket results in a unique reactor safety advantage for this concept with respect to reduced radioactive inventory and reduced fission product decay afterheat in the event of a loss-of-coolant-accident.

  1. Homopolar Gun for Pulsed Spheromak Fusion Reactors II

    SciTech Connect

    Fowler, T

    2004-06-14

    A homopolar gun is discussed that could produce the high currents required for pulsed spheromak fusion reactors even with unit current amplification and open field lines during injection, possible because close coupling between the gun and flux conserver reduces gun losses to acceptable levels. Example parameters are given for a gun compatible with low cost pulsed reactors and for experiments to develop the concept.

  2. Analysis of Critical Reactor Response for TOPAZ-II Water Immersion Scenarios

    NASA Astrophysics Data System (ADS)

    Ponomarev-Stepnoi, Nikolai N.; Glushkov, Yevgeny S.; Yermoshin, Mikhail Y.; Skorlygin, Vladimir V.

    1994-07-01

    The unmodified TOPAZ-II water immersion event leading to surrounding the reactor with water and filling with water all internal core cavities (including TFE NaK channels) may hypothetically result in criticality. This paper presents results of preliminary studies of such an accident. Possible scenarios have been analyzed as well as reactivity effects involving the water presence in internal core cavities. A preliminary coupled model has been developed to describe accident transients in the reactor and TFE. The model is based on assumptions that result in overestimating possible consequences. The numerical simulations results point at the TOPAZ-II reactor capability to quench effectively possible power bursts and predict stable periodic oscillations as a final system state, wherein steaming and then refilling up some internal core cavities occurs. That may be considered to be demonstration of the TOPAZ-II reactor self-control capability if its criticality involves water immersion event.

  3. Validating the Serpent Model of FiR 1 Triga Mk-II Reactor by Means of Reactor Dosimetry

    NASA Astrophysics Data System (ADS)

    Viitanen, Tuomas; Leppänen, Jaakko

    2016-02-01

    A model of the FiR 1 Triga Mk-II reactor has been previously generated for the Serpent Monte Carlo reactor physics and burnup calculation code. In the current article, this model is validated by comparing the predicted reaction rates of nickel and manganese at 9 different positions in the reactor to measurements. In addition, track-length estimators are implemented in Serpent 2.1.18 to increase its performance in dosimetry calculations. The usage of the track-length estimators is found to decrease the reaction rate calculation times by a factor of 7-8 compared to the standard estimator type in Serpent, the collision estimators. The differences in the reaction rates between the calculation and the measurement are below 20%.

  4. Sliding mode control of the space nuclear reactor system TOPAZ II

    SciTech Connect

    Shtessel, Y.B.; Wyant, F.J.

    1996-03-01

    The Automatic Control System (ACS) of the space nuclear reactor power system TOPAZ II that generates electricity from nuclear heat using in-core thermionic converters is considered. Sliding Mode Control Technique was applied to the reactor system controller design in order to provide the robust high accuracy following of a neutron (thermal) power reference profile in a start up regime and a payload electric power (current) reference profile following in an operation regime. Extensive simulations of the TOPAZ II reactor system with the designed sliding mode controllers showed improved accuracy and robustness of the reactor system performances in a start up regime and in an electric power supply regime as well. {copyright} {ital 1996 American Institute of Physics.}

  5. Studies on sodium boiling phenomena in out of pile rod bundles for various accidental situations in Liquid Metal Fast Breeder Reactors (LMFBR) experiments and interpretations

    NASA Astrophysics Data System (ADS)

    Seiler, J. M.; Rameau, B.

    Bundle sodium boiling in nominal geometry for different accident conditions is reviewed. Voiding of a subassembly is controlled by not only hydrodynamic effects but mainly by thermal effects. There is a strong influence of the thermal inertia of the bundle material compared to the sodium thermal inertia. Flow instability, during a slow transient, can be analyzed with numerical tools and estimated using simplified approximations. Stable boiling operational conditions under bundle mixed convection (natural convection in the reactor) can be predicted. Voiding during a fast transient can be approximated from single channel calculations. The phenomenology of boiling behavior for a subassembly with inlet completely blocked, submitted to decay heat and lateral cooling; two-phase sodium flow pressure drop in a tube of large hydraulic diameter under adiabatic conditions; critical flow phenomena and voiding rate under high power, slow transient conditions; and onset of dry out under local boiling remains problematical.

  6. Thermal baffle for fast-breeder reacton

    DOEpatents

    Rylatt, John A.

    1977-01-01

    A liquid-metal-cooled fast-breeder reactor includes a bridge structure for separating hot outlet coolant from relatively cool inlet coolant consisting of an annular stainless steel baffle plate extending between the core barrel surrounding the core and the thermal liner associated with the reactor vessel and resting on ledges thereon, there being inner and outer circumferential webs on the lower surface of the baffle plate and radial webs extending between the circumferential webs, a stainless steel insulating plate completely covering the upper surface of the baffle plate and flex seals between the baffle plate and the ledges on which the baffle plate rests to prevent coolant from washing through the gaps therebetween. The baffle plate is keyed to the core barrel for movement therewith and floating with respect to the thermal liner and reactor vessel.

  7. Tritium system design for the mirror reactors FPD-I, FPD-II, and FPD-III

    SciTech Connect

    Finn, P.A.

    1985-01-01

    The tritium system design for the Fusion Power Demonstration Reactor (FPD-I, II, and III) is described. The device operates at 25% availability. For FPD-II, an engineering mode using tritium neutral beams is part of the design.

  8. Fusion-reactor plasmas with polarized nuclei. II

    SciTech Connect

    Kulsrud, R.M.; Furth, H.P.; Valeo, E.J.; Budny, R.V.; Jassby, D.L.; Micklich, B.J.; Post, D.E.; Goldhaber, M.; Happer, W.

    1982-11-01

    New techniques of bulk polarization could be used to fuel a reactor with polarized hydrogenic atoms, so as to form a plasma of polarized nuclei. Theoretical calculations indicate that, once the nuclei of the plasma are polarized in some preferred state, they can maintain this state with a probability near 100% during their lifetime in the reactor, including possible recycling. There are a number of practical advantages to be gained from the use of polarized plasma in a fusion reactor. The nuclear reaction rates can be increased or decreased, and/or the direction of emission of the reaction products can be controlled. The D-T reaction rate can be enhanced by as much as 50%, with the reaction products emitted perpendicular to the magnetic field. Alternatively, it is possible to direct the reaction products primarily along the field, with no enhancement. In this case of the D-D reaction, the theoretical predictions are somewhat less certain. Enhancement of the reaction rate by a factor of 1.5-2.5 is to be expected. In a different polarization state, suppression of D-D reactions may be feasible - a possibility that would be of interest for a neutron-free D-He/sup 3/ reactor. A quantitative discussion of the relevant nuclear physics as well as of the various mechanisms producing depolarization is given.

  9. Utilizing FFTF: the keystone for breeder development

    SciTech Connect

    Ziff, J.J.; Arneson, S.O.

    1981-05-01

    This paper describes the role of the Fast Flux Test Facility (FFTF) in the US Department of Energy sponsored Liquid Metal Fast Breeder Reactor (LMFBR) Program. The programs that are in place to ensure that the FFTF fulfills its role as an essential key to the development of LMFBR technology are delineated. A detailed FFTF Operating Plan has been developed to present in integrated form the strategy for gaining maximum useful information from the planned FFTF operations. The three principal areas of FFTF Utilization: Plant Utilization, Irradiation Testing, and Safety, combine to form the overall FFTF Operating Plan. Primary areas where FFTF is already making major contributions to LMFBR development are described.

  10. Benchmark specifications for EBR-II shutdown heat removal tests

    SciTech Connect

    Sofu, T.; Briggs, L. L.

    2012-07-01

    Argonne National Laboratory (ANL) is hosting an IAEA-coordinated research project on benchmark analyses of sodium-cooled fast reactor passive safety tests performed at the Experimental Breeder Reactor-II (EBR-II). The benchmark project involves analysis of a protected and an unprotected loss of flow tests conducted during an extensive testing program within the framework of the U.S. Integral Fast Reactor program to demonstrate the inherently safety features of EBR-II as a pool-type, sodium-cooled fast reactor prototype. The project is intended to improve the participants' design and safety analysis capabilities for sodium-cooled fast reactors through validation and qualification of safety analysis codes and methods. This paper provides a description of the EBR-II tests included in the program, and outlines the benchmark specifications being prepared to support the IAEA-coordinated research project. (authors)

  11. Generating unstructured nuclear reactor core meshes in parallel

    SciTech Connect

    Jain, Rajeev; Tautges, Timothy J.

    2014-10-24

    Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor core examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.

  12. Generating unstructured nuclear reactor core meshes in parallel

    DOE PAGESBeta

    Jain, Rajeev; Tautges, Timothy J.

    2014-10-24

    Recent advances in supercomputers and parallel solver techniques have enabled users to run large simulations problems using millions of processors. Techniques for multiphysics nuclear reactor core simulations are under active development in several countries. Most of these techniques require large unstructured meshes that can be hard to generate in a standalone desktop computers because of high memory requirements, limited processing power, and other complexities. We have previously reported on a hierarchical lattice-based approach for generating reactor core meshes. Here, we describe efforts to exploit coarse-grained parallelism during reactor assembly and reactor core mesh generation processes. We highlight several reactor coremore » examples including a very high temperature reactor, a full-core model of the Korean MONJU reactor, a ¼ pressurized water reactor core, the fast reactor Experimental Breeder Reactor-II core with a XX09 assembly, and an advanced breeder test reactor core. The times required to generate large mesh models, along with speedups obtained from running these problems in parallel, are reported. A graphical user interface to the tools described here has also been developed.« less

  13. Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

    SciTech Connect

    Mckerley, Bill; Bustamante, Jacqueline M; Costa, David A; Drypolcher, Anthony F; Hickey, Joseph

    2010-01-01

    We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is in support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to consolidate the

  14. Recent developments in Topaz II reactor safety assessments

    SciTech Connect

    Marshall, A.C.

    1993-07-01

    In December 1991, the Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of a US launch of a Russian Topaz II space nuclear power system. The primary mission goal would be to demonstrate and evaluate Nuclear Electric Propulsion technology to establish a capability for future civilian and military missions. A preliminary nuclear safety assessment, involving selected safety analyses, was initiated to determine whether or not a space mission could be conducted safely and within budget constraints. This paper describes the preliminary safety assessment results and the nuclear safety program now being established for the Nuclear Electric Propulsion Space Test Program (NEPSTP).

  15. Recent developments in Topaz-II reactor safety assessments

    SciTech Connect

    Marshall, A.C. )

    1993-01-01

    In December 1991, the Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of a US launch of a Russian Topaz-II space nuclear power system. The primary mission goal would be to demonstrate and evaluate nuclear electric propulsion technology to establish a capability for future civilian and military missions. A preliminary nuclear safety analysis was initiated to determine whether or not a space mission could be conducted safely and within budget constraints. This paper describes preliminary safety analysis results and the nuclear safety program now being established for the NEP space test (NEPST).

  16. EBIS charge breeder for CARIBU.

    PubMed

    Kondrashev, S; Barcikowski, A; Dickerson, C; Fischer, R; Ostroumov, P N; Vondrasek, R; Pikin, A

    2014-02-01

    A high-efficiency charge breeder based on an Electron Beam Ion Source (EBIS) is being developed by the ANL Physics Division to increase the intensity and improve the purity of accelerated radioactive ion beams. A wide variety of low-energy neutron-rich ion beams are produced by the Californium Rare Isotope Breeder Upgrade (CARIBU) for the Argonne Tandem Linac Accelerator System (ATLAS). These beams will be charge-bred by an EBIS charge breeder to a charge-to-mass ratio (q/A) ≥ 1/7 and accelerated by ATLAS to energies of about 10 MeV/u. The assembly of the CARIBU EBIS charge breeder except the injection/extraction beam lines has been completed. This summer we started electron beam commissioning of the EBIS. The first results on electron beam extraction, transport from the electron gun to a high power electron collector are presented and discussed. PMID:24593606

  17. EBIS charge breeder for CARIBU

    NASA Astrophysics Data System (ADS)

    Kondrashev, S.; Barcikowski, A.; Dickerson, C.; Fischer, R.; Ostroumov, P. N.; Vondrasek, R.; Pikin, A.

    2014-02-01

    A high-efficiency charge breeder based on an Electron Beam Ion Source (EBIS) is being developed by the ANL Physics Division to increase the intensity and improve the purity of accelerated radioactive ion beams. A wide variety of low-energy neutron-rich ion beams are produced by the Californium Rare Isotope Breeder Upgrade (CARIBU) for the Argonne Tandem Linac Accelerator System (ATLAS). These beams will be charge-bred by an EBIS charge breeder to a charge-to-mass ratio (q/A) ≥ 1/7 and accelerated by ATLAS to energies of about 10 MeV/u. The assembly of the CARIBU EBIS charge breeder except the injection/extraction beam lines has been completed. This summer we started electron beam commissioning of the EBIS. The first results on electron beam extraction, transport from the electron gun to a high power electron collector are presented and discussed.

  18. Artificial intelligence program in a computer application supporting reactor operations

    SciTech Connect

    Stratton, R.C.; Town, G.G.

    1985-01-01

    Improving nuclear reactor power plant operability is an ever-present concern for the nuclear industry. The definition of plant operability involves a complex interaction of the ideas of reliability, safety, and efficiency. This paper presents observations concerning the issues involved and the benefits derived from the implementation of a computer application which combines traditional computer applications with artificial intelligence (AI) methodologies. A system, the Component Configuration Control System (CCCS), is being installed to support nuclear reactor operations at the Experimental Breeder Reactor II.

  19. HYLIFE-II inertial confinement fusion reactor design

    SciTech Connect

    Moir, R.W.

    1990-12-14

    The HYLIFE-2 inertial fusion power plant design study uses a liquid fall, in the form of jets to protect the first structural wall from neutron damage, x rays, and blast to provide a 30-y lifetime. HYLIFE-1 used liquid lithium. HYLIFE 2 avoids the fire hazard of lithium by using a molten salt composed of fluorine, lithium, and beryllium (Li{sub 2}BeF{sub 4}) called Flibe. Access for heavy-ion beams is provided. Calculations for assumed heavy-ion beam performance show a nominal gain of 70 at 5 MJ producing 350 MJ, about 5.2 times less yield than the 1.8 GJ from a driver energy of 4.5 MJ with gain of 400 for HYLIFE-1. The nominal 1 GWe of power can be maintained by increasing the repetition rate by a factor of about 5.2, from 1.5 to 8 Hz. A higher repetition rate requires faster re-establishment of the jets after a shot, which can be accomplished in part by decreasing the jet fall height and increasing the jet flow velocity. Multiple chambers may be required. In addition, although not considered for HYLIFE-1, there is undoubtedly liquid splash that must be forcibly cleared because gravity is too slow, especially at high repetition rates. Splash removal can be accomplished by either pulsed or oscillating jet flows. The cost of electricity is estimated to be 0.09 $/kW{center dot}h in constant 1988 dollars, about twice that of future coal and light water reactor nuclear power. The driver beam cost is about one-half the total cost. 15 refs., 9 figs., 3 tabs.

  20. Liquid Metal Fast Breeder Reactors: a bibliography

    SciTech Connect

    Raleigh, H.D.

    1980-11-01

    This bibliogralphy includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

  1. Liquid Metal Fast Breeder Reactors: a bibliography

    SciTech Connect

    Raleigh, H.D.

    1980-11-01

    This bibliography includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

  2. The search for advanced remote technology in fast reactor reprocessing

    SciTech Connect

    Burch, W.D.; Herndon, J.N.; Stradley, J.G.

    1990-01-01

    Research and development in fast reactor reprocessing has been under way about 20 years in several countries throughout the world. During the past decade in France and the United Kingdom, active development programs have been carried out in breeder reprocessing. Actual fuels from their demonstration reactors have been reprocessed in small-scale facilities. Early US work in breeder reprocessing was carried out at the EBR-II facilities with the early metal fuels, and interest has renewed recently in metal fuels. A major, comprehensive program, focused on oxide fuels, has been carried out in the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL) since 1974. Germany and Japan have also carried out development programs in breeder reprocessing, and Japan appears committed to major demonstration of breeder reactors and their fuel cycles. While much of the effort in all of these programs addressed process chemistry and process hardware, a significant element of many of these programs, particularly the CFRP, has been on advancements in facility concepts and remote maintenance features. This paper will focus principally on the search for improved facility concepts and better maintenance systems in the CFRP and, in turn, on how developments at ORNL have influenced the technology elsewhere.

  3. Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.

    PubMed

    Henry, R; Tiselj, I; Snoj, L

    2015-03-01

    New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. PMID:25576735

  4. IEC-^3He Breeder for D-^3He Satellite Systems.

    NASA Astrophysics Data System (ADS)

    Chacon, L.; Miley, G. H.

    1996-11-01

    D-^3He fusion minimizes neutrons and maximizes charged fusion products, enabling increased energy recovery efficiency by direct conversion. However, scarce ^3He terrestrial resources have deterred R&D on this alternative. Here, we explore ^3He production through Inertial Electrostatic Confinement^1 (IEC) D-breeders, which supply ^3He to FRC D-^3He satellite reactors.^2 Favorable features for the IEC breeder include simplicity, low cost, easy extraction of fusion products, and compatibility with direct conversion. The breeder-satellite system energy balance is analyzed taking the net energy gain of the overall system, Q_N, as the figure of merit. Breeding is applicable for systems where the satellite Q-value, Q_S, > the breeder Q-value, Q_B. For improved performance, i.e., for high Q_N, QS >= QB >> 1 is needed; however, lower QB values (typical of the IEC) are permissible and still offer sufficient Q_N. An economic study determined breeding produces ^3He at a cost comparable to lunar ^3He, already shown to lead to competitive power.^3 The cost of electricity (COE) for the breeder-satellite complex was compared with the ARTEMIS COE,^4 using lunar ^3He fuel: assuming one satellite (1000 MWe)/breeder (170 MWe), the ratio of the breeding system COE to the lunar mining base COE is ~ 1.2. However, economic breeding is driven by large IEC breeder powers, i.e., increased ^3He breeding rates. Thus, the COE ratio approaches unity with two or three satellites/breeder, requiring increased breeder size and power (340 MWe for 2 satellites, 510 MWe for 3 satellites). Such systems potentially provide a ``bridge'' to a future lunar ^3He economy. 1. G.H. Miley et al., Dense Z-pinches, AIP Conf. 299, AIP Press, 675-689 (1994). 2. G.H. Miley, Nucl. Instrum. Methods, A271, 197-202 (1988). 3. L.J. Wittenberg et al., Fusion Technol., 10, 167-178 (1986). 4. H. Momota et al., Fusion Technol., 21, 2307-2323 (1992).

  5. The regulatory quagmire underlying the TOPAZ II exhibition: The nuclear regulatory commission's jurisdiction over the TOPAZ II reactor system

    NASA Astrophysics Data System (ADS)

    Lawrence, John W.

    1992-01-01

    At the 8th Symposium on Space Nuclear Power Systems, 6-10 January 1990, the Union of Soviet Socialist Republics displayed a TOPAZ II thermionic space nuclear reactor. Underlying that exhibition was a regulatory quagmire created by a decision of the Nuclear Regulatory Commission that an import license was required to bring the device into the United States, and that an amendment to their regulations governing exports was required to return the device to the Soviet Union latter that summer. This paper briefly reviews the jurisdictional issue of how the Nuclear Regulatory Commission exerted its authority over the TOPAZ II reactor system, as well as the manner in which the import and export licensing actions were accomplished. In sum, the paper offers an independent interpretation of the applicable import and export regulations, and concludes that the Nuclear Regulatory Commission likely need not have exercised its import jurisdiction, and notwithstanding the initial assumption of jurisdiction, an export license likely could have been issued without an amendment to the then existing regulations.

  6. Startup experience at the University of Texas TRIGA Mark II Reactor

    SciTech Connect

    Bauer, Thomas L.; Wehring, Bernard W.

    1992-07-01

    After eight years of singular effort, the UT-TRIGA Mark II research reactor was licensed and is fully operational. This reactor is the focus of a new reactor laboratory facility which is located at the Balcones Research Center, a north Austin campus of The University of Texas at Austin. The UT-TRIGA reactor is licensed for 1.1 MW steady power operation and 3 dollar pulsing. A startup program was implemented upon receipt of the facility license on January 17, 1992. Several facility features are unique to this startup. Among these were the use of fuel with various burnup and a digital control system. The reactor laboratory staff with assistance from a General Atomics instrumentation engineer performed all phases of the startup program. Core loading began in February 1992 with final testing completed in May 1992. Several unusual problems were encountered during this time. Experiment authorizations have been written to resume Neutron Activation Analysis programs and isotope production. Several neutron beam tube experiments are in the design and test phase. (author)

  7. US solid breeder blanket design for ITER

    SciTech Connect

    Gohar, Y.; Attaya, H.; Billone, M.; Lin, C.; Johnson, C.; Majumdar, S.; Smith, D. ); Goranson, P.; Nelson, B.; Williamson, D.; Baker, C. ); Raffray, A.; Badawi, A.; Gorbis, Z.; Ying, A.; Abdou, M. ); Sviatoslavsky, I.; Blanchard, J.; Mogahed, E.; Sawan, M.; Kulcinski, G. )

    1990-09-01

    The US blanket design activity has focused on the developments and the analyses of a solid breeder blanket concept for ITER. The main function of this blanket is to produce the necessary tritium required for the ITER operation and the test program. Safety, power reactor relevance, low tritium inventory, and design flexibility are the main reasons for the blanket selection. The blanket is designed to operate satisfactorily in the physics and the technology phases of ITER without the need for hardware changes. Mechanical simplicity, predictability, performance, minimum cost, and minimum R D requirements are the other criteria used to guide the design process. The design aspects of the blanket are summarized in this paper. 2 refs., 7 figs., 3 tabs.

  8. Lithium reprocessing technology for ceramic breeders

    NASA Astrophysics Data System (ADS)

    Tsuchiya, Kunihiko; Kawamura, Hiroshi; Saito, Minoru; Tatenuma, Katuyashi; Kainose, Mitsuru

    1995-03-01

    Lithium ceramics have been receiving considerable attention as tritium breeding materials for fusion reactors. Reprocessing technology development for these materials is proposed to recover lithium, as an effective use of resources and to remove radioactive isotopes. Four potential ceramic breeders (Li 2O, LiAlO 2, Li 2ZrO 3 and Li 4SiO 4) were prepared in order to estimate their dissolution properties in water and various acids (HCl, HNO 3, H 2SO 4, HF and aqua regia). The dissolution rates were determined by comparing the weight of the residue with that of the starting powder (the weight method). Recovery properties of lithium were examined by the precipitation method.

  9. Solid breeder/structure mechanical interaction and thermal stability

    SciTech Connect

    Liu, Y.Y.; Billone, M.C.; Taghavi, K.

    1985-04-01

    Solid breeder/structure mechanical interaction (BSMI) during fusion reactor blanket operation is a potential failure mode which could limit the lifetime of the blanket. The severity of BSMI will generally depend on the materials, specific blanket designs, and blanket operating conditions. Thermomechanical analyses performed for a helium-cooled blanket employing Li/sub 2/O/HT-9 plates indicate that BSMI could be a serious concern for this blanket.

  10. Startup control of the TOPAZ-II space nuclear reactor. Master`s thesis

    SciTech Connect

    Astrin, C.D.

    1996-09-01

    The Russian designed and manufactured TOPAZ-II Thermionic Nuclear Space Reactor has been supplied to the Ballistic Missile Defense Organization for study as part of the TOPAZ International Program. A Preliminary Nuclear Safety Assessment investigated the readiness to use the TOPAZ-II in support of a Nuclear Electric Propulsion Space Test Mission (NEPSTP). Among the anticipated system modifications required for launching the TOPAZ-II system within safety goals is for a U.S. designed Automatic Control System. The requirements and desired features of such a control system are developed based upon U.S. safety standards. System theory and design are presented in order to establish the basis for development of a hybrid control model from available simulations. The model is verified and then used in exploration of various control schemes and casualty analysis providing groundwork for future Automatic Control System design.

  11. EBR-II: twenty years of operating experience

    SciTech Connect

    Lentz, G.L.; Buschman, H.W.; Smith, R.N.

    1985-01-01

    Experimental Breeder Reactor No. 2 (EBR-II) is an unmoderated, sodium-cooled reactor with a design power of 62.5 MWt. For the last 20 years EBR-II has operated safely, has demonstrated stable operating characteristics, has shown excellent performance of its sodium components, and has had an excellent plant factor. These years of operating experience provide a valuable resource to the nuclear community for the development and design of future liquid metal fast reactors. This report provides a brief description of the EBR-II plant and its early operating experience, describes some recent problems of interest to the nuclear community, and also mentions some of the significant operating achievements of EBR-II. Finally, a few words and speculations on EBR-II's future are offered. 4 figs., 1 tab.

  12. Spallator - accelerator breeder

    SciTech Connect

    Steinberg, M.

    1985-01-01

    The concept involves the use of spallation neutrons produced by interaction of a high energy proton (1 to 2 GeV) from a linear accelerator (LINAC) with a heavy metal target (uranium). The principal spallator concept is based on generating fissile fuel for use in LWR nuclear power plants. The spallator functions in conjunction with a reprocessing plant to regenerate and produce the Pu-239 or U-233 for fabrication into fresh LWR reactor fuel elements. Advances in proton accelerator technology has provided a solid base for predicting performance and optimizing the design of a reliable, continuous wave, high-current LINAC required by a fissile fuel production machine.

  13. Evaluation of tritium release properties of advanced tritium breeders

    SciTech Connect

    Hoshino, T.; Ochiai, K.; Edao, Y.; Kawamura, Y.

    2015-03-15

    Demonstration power plant (DEMO) fusion reactors require advanced tritium breeders with high thermal stability. Lithium titanate (Li{sub 2}TiO{sub 3}) advanced tritium breeders with excess Li (Li{sub 2+x}TiO{sub 3+y}) are stable in a reducing atmosphere at high temperatures. Although the tritium release properties of tritium breeders are documented in databases for DEMO blanket design, no in situ examination under fusion neutron (DT neutron) irradiation has been performed. In this study, a preliminary examination of the tritium release properties of advanced tritium breeders was performed, and DT neutron irradiation experiments were performed at the fusion neutronics source (FNS) facility in JAEA. Considering the tritium release characteristics, the optimum grain size after sintering is <5 μm. From the results of the optimization of granulation conditions, prototype Li{sub 2+x}TiO{sub 3+y} pebbles with optimum grain size (<5 μm) were successfully fabricated. The Li{sub 2+x}TiO{sub 3+y} pebbles exhibited good tritium release properties similar to the Li{sub 2}TiO{sub 3} pebbles. In particular, the released amount of HT gas for easier tritium handling was higher than that of HTO water. (authors)

  14. Experimental power density distribution benchmark in the TRIGA Mark II reactor

    SciTech Connect

    Snoj, L.; Stancar, Z.; Radulovic, V.; Podvratnik, M.; Zerovnik, G.; Trkov, A.; Barbot, L.; Domergue, C.; Destouches, C.

    2012-07-01

    In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the few available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)

  15. Modification of the radial beam port of ITU TRIGA Mark II research reactor for BNCT applications.

    PubMed

    Akan, Zafer; Türkmen, Mehmet; Çakir, Tahir; Reyhancan, İskender A; Çolak, Üner; Okka, Muhittin; Kiziltaş, Sahip

    2015-05-01

    This paper aims to describe the modification of the radial beam port of ITU (İstanbul Technical University) TRIGA Mark II research reactor for BNCT applications. Radial beam port is modified with Polyethylene and Cerrobend collimators. Neutron flux values are measured by neutron activation analysis (Au-Cd foils). Experimental results are verified with Monte Carlo results. The results of neutron/photon spectrum, thermal/epithermal neutron flux, fast group photon fluence and change of the neutron fluxes with the beam port length are presented. PMID:25746919

  16. Independent Safety Assessment of the TOPAZ-II space nuclear reactor power system (Revised)

    SciTech Connect

    1993-09-01

    The Independent Safety Assessment described in this study report was performed to assess the safety of the design and launch plans anticipated by the U.S. Department of Defense (DOD) in 1993 for a Russian-built, U.S.-modified, TOPAZ-II space nuclear reactor power system. Its conclusions, and the bases for them, were intended to provide guidance for the U.S. Department of Energy (DOE) management in the event that the DOD requested authorization under section 91b. of the Atomic Energy Act of 1954, as amended, for possession and use (including ground testing and launch) of a nuclear-fueled, modified TOPAZ-II. The scientists and engineers who were engaged to perform this assessment are nationally-known nuclear safety experts in various disciplines. They met with participants in the TOPAZ-II program during the spring and summer of 1993 and produced a report based on their analysis of the proposed TOPAZ-II mission. Their conclusions were confined to the potential impact on public safety and did not include budgetary, reliability, or risk-benefit analyses.

  17. Advanced Safeguards Approaches for New Fast Reactors

    SciTech Connect

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  18. Ceramic breeder materials : status and needs.

    SciTech Connect

    Johnson, C.E.

    1998-02-02

    The tritium breeding blanket is one of the most important components of a fusion reactor because it directly involves both energy extraction and tritium production, both of which are critical to fusion power. Because of their overall desirable properties, lithium-containing ceramic solids are recognized as attractive tritium breeding materials for fusion reactor blankets. Indeed, their inherent thermal stability and chemical inertness are significant safety advantages. In numerous in-pile experiments, these materials have performed well, showing good thermal stability and good tritium release characteristics. Tritium release is particularly facile when an argon or helium purge gas containing hydrogen, typically at levels of about 0.1%, is used. However, the addition of hydrogen to the purge gas imposes a penalty when it comes to recovery of the tritium produced in the blanket. In particular, a large amount of hydrogen in the purge gas will necessitate a large multiple-stage tritium purification unit, which could translate into higher costs. Optimizing tritium release while minimizing the amount of hydrogen necessary in the purge gas requires a deeper understanding of the tritium release process, especially the interactions of hydrogen with the surface of the lithium ceramic. This paper reviews the status of ceramic breeder research and highlights several issues and data needs.

  19. Fusion breeder: its potential role and prospects

    SciTech Connect

    Lee, J.D.

    1981-01-01

    The fusion breeder is a concept that utilizes 14 MeV neutrons from D + T ..-->.. n(14.1 MeV) + ..cap alpha..(3.5 MeV) fusion reactions to produce more fuel than the tritium (T) needed to sustain the fusion process. This excess fuel production capacity is used to produce fissile material (Pu-239 or U-233) for subsequent use in fission reactors. We are concentrating on a class of blankets we call fission suppressed. The blanket is the region surrounding the fusion plasma in which fusion neutrons interact to produce fuel and heat. The fission-suppressed blanket uses non-fission reactions (mainly (n,2n) or (n,n't)) to generate excess neutrons for the production of net fuel. This is in contrast to the fast fission class of blankets which use (n,fiss) reactions to generate excess neutrons. Fusion reactors with fast fission blankets are commony known as fusion-fission hybrids because they combine fusion and fission in the same device.

  20. Neutron flux characterisation of the Pavia TRIGA Mark II research reactor for radiobiological and microdosimetric applications.

    PubMed

    Alloni, D; Prata, M; Salvini, A; Ottolenghi, A

    2015-09-01

    Nowadays the Pavia TRIGA reactor is available for national and international collaboration in various research fields. The TRIGA Mark II nuclear research reactor of the Pavia University offers different in- and out-core neutron irradiation channels, each characterised by different neutron spectra. In the last two years a campaign of measurements and simulations has been performed in order to guarantee a better characterisation of these different fluxes and to meet the demands of irradiations that require precise information on these spectra in particular for radiobiological and microdosimetric studies. Experimental data on neutron fluxes have been collected analysing and measuring the gamma activity induced in thin target foils of different materials irradiated in different TRIGA experimental channels. The data on the induced gamma activities have been processed with the SAND II deconvolution code and finally compared with the spectra obtained with Monte Carlo simulations. The comparison between simulated and measured spectra showed a good agreement allowing a more precise characterisation of the neutron spectra and a validation of the adopted method. PMID:25958412

  1. Characteristics and control response of the TOPAZ II Reactor System Real-time Dynamic Simulator

    SciTech Connect

    Kwok, K.S.

    1993-11-12

    A dynamic simulator of the TOPAZ II reactor system has been developed for the Nuclear Electric Propulsion Space Test Program. The simulator combines first-principle modeling and empirical correlations in its algorithm to attain the modeling accuracy and computational through-put that are required for real-time execution. The overall execution time of the simulator for each time step is 15 ms when no data is written to the disk, and 18 ms when nine double precision data points are written to the disk once in every time step. The simulation program has been tested and it is able to handle a step decrease of $8 worth of reactivity. It also provides simulations of fuel, emitter, collector, stainless steel, and ZrH moderator failures. Presented in this paper are the models used in the calculations, a sample simulation session, and a discussion of the performance and limitations of the simulator. The simulator has been found to provide realistic real-time dynamic response of the TOPAZ II reactor system under both normal and casualty conditions.

  2. Real-time dynamic simulator for the Topaz II reactor power system

    SciTech Connect

    Kwok, K.S.

    1994-10-01

    A dynamic simulator of the TOPAZ II reactor system has been developed for the Nuclear Electric Propulsion Space Test Program. The simulator is a self-contained IBM-PC compatible based system that executes at a speed faster than real-time. The simulator combines first-principle modeling and empirical correlations in its algorithm to attain the modeling accuracy and computational through-put that are required for real-time execution. The overall execution time of the simulator for each time step is 15 ms when no data is written to the disk, and 18 ms when nine double precision data points are written to the disk once in every time step. The simulation program has been tested and it is able to handle a step decrease of $8 worth of reactivity. It also provides simulation of fuel, emitter, collector, stainless steel, and ZrH moderator failures. Presented in this paper are the models used in the calculations, a sample simulation session, and a discussion of the performance and limitations of the simulator. The simulator has been found to provide realistic real-time dynamic response of the TOPAZ II reactor system under both normal and causality conditions.

  3. Testing the applicability of the k0-NAA method at the MINT's TRIGA MARK II reactor

    NASA Astrophysics Data System (ADS)

    Siong, Wee Boon; Dung, Ho Manh; Wood, Ab. Khalik; Salim, Nazaratul Ashifa Abd.; Elias, Md. Suhaimi

    2006-08-01

    The Analytical Chemistry Laboratory at MINT is using the NAA technique since 1980s and is the only laboratory in Malaysia equipped with a research reactor, namely the TRIGA MARK II. Throughout the years the development of NAA technique has been very encouraging and was made applicable to a wide range of samples. At present, the k0 method has become the preferred standardization method of NAA ( k0-NAA) due to its multi-elemental analysis capability without using standards. Additionally, the k0 method describes NAA in physically and mathematically understandable definitions and is very suitable for computer evaluation. Eventually, the k0-NAA method has been adopted by MINT in 2003, in collaboration with the Nuclear Research Institute (NRI), Vietnam. The reactor neutron parameters ( α and f) for the pneumatic transfer system and for the rotary rack at various locations, as well as the detector efficiencies were determined. After calibration of the reactor and the detectors, the implemented k0 method was validated by analyzing some certified reference materials (including IAEA Soil 7, NIST 1633a, NIST 1632c, NIST 1646a and IAEA 140/TM). The analysis results of the CRMs showed an average u score well below the threshold value of 2 with a precision of better than ±10% for most of the elemental concentrations obtained, validating herewith the introduction of the k0-NAA method at the MINT.

  4. EBR-II Data Digitization

    SciTech Connect

    Yoon, Su-Jong; Rabiti, Cristian; Sackett, John

    2014-08-01

    1. Objectives To produce a validation database out of those recorded signals it will be necessary also to identify the documents need to reconstruct the status of reactor at the time of the beginning of the recordings. This should comprehends the core loading specification (assemblies type and location and burn-up) along with this data the assemblies drawings and the core drawings will be identified. The first task of the project will be identify the location of the sensors, with respect the reactor plant layout, and the physical quantities recorded by the Experimental Breeder Reactor-II (EBR-II) data acquisition system. This first task will allow guiding and prioritizing the selection of drawings needed to numerically reproduce those signals. 1.1 Scopes and Deliverables The deliverables of this project are the list of sensors in EBR-II system, the identification of storing location of those sensors, identification of a core isotopic composition at the moment of the start of system recording. Information of the sensors in EBR-II reactor system was summarized from the EBR-II system design descriptions listed in Section 1.2.

  5. Failed fuel identification techniques for liquid-metal cooled reactors

    SciTech Connect

    Lambert, J.D.B.; Gross, K.C.; Mikaili, R.; Frank, S.M.; Cutforth, D.C.; Angelo, P.L.

    1995-06-01

    The Experimental Breeder Reactor II (EBR-II), located in Idaho and operated for the US Department of Energy by Argonne National Laboratory, has been used as an irradiation testbed for LMR fuels and components for thirty years. During this time many endurance tests have been carried out with experimental LMR metal, oxide, carbide and nitride fuel elements, in which cladding failures were intentionally allowed to occur. This paper describes methods that have been developed for the detection, identification and verification of fuel failures.

  6. Launch Vehicle Fire Accident Preliminary Analysis of a Liquid-Metal Cooled Thermionic Nuclear Reactor: TOPAZ-II

    NASA Astrophysics Data System (ADS)

    Hu, G.; Zhao, S.; Ruan, K.

    2012-01-01

    In this paper, launch vehicle propellant fire accident analysis of TOPAZ-II reactor has been done by a thermionic reactor core analytic code-TATRHG(A) developed by author. When a rocket explodes on a launch pad, its payload-TOPAZ-II can be subjected to a severe thermal environment from the resulting fireball. The extreme temperatures associated with propellant fires can create a destructive environment in or near the fireball. Different kind of propellants - liquid propellant and solid propellant which will lead to different fire temperature are considered. Preliminary analysis shows that the solid propellant fires can melt the whole toxic beryllium radial reflector.

  7. Deactivation of the EBR-II complex

    SciTech Connect

    Michelbacher, J.A.; Earle, O.K.; Henslee, S.P.

    1997-12-31

    In January of 1994, the Department of Energy mandated the termination of the Integral Fast Reactor (IFR) Program, effective October 1, 1994. To comply with this decision, Argonne National Laboratory-West (ANL-W) prepared a plan providing detailed requirements to place the Experimental Breeder Reactor-II (EBR-II) in a radiologically and industrially safe condition, including removal of all irradiated fuel assemblies from the reactor plant, and removal and stabilization of the primary and secondary sodium, a liquid metal used to transfer heat within the reactor plant. The ultimate goal of the deactivation process is to place the EBR-II complex in a stable condition until a decontamination and decommissioning (D&D) plan can be prepared, thereby minimizing requirements for maintenance and surveillance and maximizing the amount of time for radioactive decay. The final closure state will be achieved in full compliance with federal, state and local environmental, safety, and health regulations and requirements. The decision to delay the development of a detailed D&D plan has necessitated this current action. The EBR-II is a pool-type reactor. The primary system contains approximately 87,000 gallons of sodium, while the secondary system has 13,000 gallons. In order to properly dispose of the sodium in compliance with the Resource Conservation and Recovery Act (RCRA), a facility has been built to react the sodium to a dry carbonate powder in a two stage process. Deactivation of a liquid metal fast breeder reactor (LMFBR) presents unique concerns. Residual amounts of sodium remaining in the primary and secondary systems must be either reacted or inerted to preclude future concerns with sodium-air reactions that generate explosive mixtures of hydrogen and leave corrosive compounds. Residual amounts of sodium on components will effectively {open_quotes}solder{close_quotes} components in place, making future operation or removal unfeasible.

  8. A US perspective on fast reactor fuel fabrication technology and experience. Part II: Ceramic fuels

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Fielding, Randall S.; Porter, Douglas L.; Meyer, Mitchell K.; Makenas, Bruce J.

    2009-08-01

    This paper is Part II of a review focusing on the United States experience with oxide, carbide, and nitride fast reactor fuel fabrication. Over 60 years of research in fuel fabrication by government, national laboratories, industry, and academia has culminated in a foundation of research and resulted in significant improvements to the technologies employed to fabricate these fuel types. This part of the review documents the current state of fuel fabrication technologies in the United States for each of these fuel types, some of the challenges faced by previous researchers, and how these were overcome. Knowledge gained from reviewing previous investigations will aid both researchers and policy makers in forming future decisions relating to nuclear fuel fabrication technologies.

  9. Analysis of safety limits of the Moroccan TRIGA MARK II research reactor

    NASA Astrophysics Data System (ADS)

    Erradi, L.; Essadki, H.

    2001-06-01

    The main objective of this study is to check the ability of the Moroccan TRIGA MARK II research reactor, designed to use natural convection cooling, to operate at its nominal power (2 MW) with sufficient safety margins. The neutronic analysis of the core has been performed using Leopard and Mcrac codes and the parameters of interest were the power distributions, the power peaking factors and the core excess reactivity. The thermal hydraulic analysis of the TRIGA core was performed using the French code FLICA designed for transient and study state situations. The main safety related parameters of the core have been evaluated with special emphasises on the following: maximum fuel temperature, minimum DNBR and maximum void fraction. The obtained results confirm the designer predictions except for the void fraction.

  10. Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia.

    PubMed

    Yavar, A R; Khalafi, H; Kasesaz, Y; Sarmani, S; Yahaya, R; Wood, A K; Khoo, K S

    2012-10-01

    A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. PMID:22885391

  11. Reentry safety for the Topaz II Space Reactor: Issues and analyses

    SciTech Connect

    Connell, L.W.; Trost, L.C.

    1994-03-01

    This report documents the reentry safety analyses conducted for the TOPAZ II Nuclear Electric Propulsion Space Test Program (NEPSTP). Scoping calculations were performed on the reentry aerothermal breakup and ground footprint of reactor core debris. The calculations were used to assess the risks associated with radiologically cold reentry accidents and to determine if constraints should be placed on the core configuration for such accidents. Three risk factors were considered: inadvertent criticality upon reentry impact, atmospheric dispersal of U-235 fuel, and the Special Nuclear Material Safeguards risks. Results indicate that the risks associated with cold reentry are very low regardless of the core configuration. Core configuration constraints were therefore not established for radiologically cold reentry accidents.

  12. 78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-14

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos. (As Shown In Attachment 1), License Nos. (As Shown In Attachment 1), EA-13-109; Order Modifying Licenses With Regard to Reliable Hardened...

  13. Helium-cooled molten-salt fusion breeder

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Fulton, F.J.; Huegel, F.; Neef, W.S. Jr.; Sherwood, A.E.; Berwald, D.H.; Whitley, R.H.; Wong, C.P.C.; Devan, J.H.

    1984-12-01

    We present a new conceptual design for a fusion reactor blanket that is intended to produce fissile material for fission power plants. Fast fission is suppressed by using beryllium instead of uranium to multiply neutrons. Thermal fission is suppressed by minimizing the fissile inventory. The molten-salt breeding medium (LiF + BeF/sub 2/ + ThF/sub 4/) is circulated through the blanket and to the on-line processing system where /sup 233/U and tritium are continuously removed. Helium cools the blanket and the austenitic steel tubes that contain the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion by molten salt. We estimate that a breeder having 3000 MW of fusion power will produce 6500 kg of /sup 233/U per year. This amount is enough to provide makeup for 20 GWe of light-water reactors per year or twice that many high-temperature gas-cooled reactors or Canadian heavy-water reactors. Safety is enhanced because the afterheat is low and blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times a light-water reactor of the same power. The estimated cost of the /sup 233/U produced is $40/g for fusion plants costing 2.35 times that of a light-water reactor if utility owned or $16/g if government owned.

  14. VISA-II: a computer code for predicting the probability of reactor pressure vessel failure

    SciTech Connect

    Simonen, F.A.; Johnson, K.I.; Liebetrau, A.M.; Engel, D.W.; Simonen, E.P.

    1986-03-01

    The VISA-II (Vessel Integrity Simulation Analysis code was originally developed as part of the NRC staff evaluation of pressurized thermal shock. VISA-II uses Monte Carlo simulation to evaluate the failure probability of a pressurized water reactor (PWR) pressure vessel subjected to a pressure and thermal transient specified by the user. Linear elastic fracture mechanics methods are used to model crack initiation and propagation. Parameters for initial crack size and location, copper content, initial reference temperature of the nil-ductility transition, fluence, crack-initiation fracture toughness, and arrest fracture toughness are treated as random variables. This report documents an upgraded version of the original VISA code as described in NUREG/CR-3384. Improvements include a treatment of cladding effects, a more general simulation of flaw size, shape and location, a simulation of inservice inspection, an updated simulation of the reference temperature of the nil-ductility transition, and treatment of vessels with multiple welds and initial flaws. The code has been extensively tested and verified and is written in FORTRAN for ease of installation on different computers. 38 refs., 25 figs.

  15. Comparison of codes and neutron IC data used in US and Russia for the Topaz-II nuclear reactor assessment

    SciTech Connect

    Glushkov, Y.S.; Ponomarev-Stepnoi, N.N.; Kompanietz, G.V.; Gomin, Y.A.; Maiorov, L.V.; Lobynstev, V.A.; Polyakov, D.N.; Sapir, J.; Streetman, J.R.

    1993-11-01

    Topaz-II is a heterogeneous, epithermal reactor, fueled with highly enriched uranium-dioxide, cooled with NaK, and moderated with zirconium-hydride. The reactor core contains 37 single-cell thermionic fuel elements, and is surrounded by a radial beryllium reflector that contains 12 rotatable control drums with poison segments. For the physics analysis of TOPAZ II it is necessary to use the Monte Carlo method. The United States (US) and Russia used two different Monte Carlo codes, namely MCNP and MCU-2, respectively. The work described in this paper was aimed at comparing the codes and neutronic data used in the US and Russia for verification of Topaz-II nuclear safety. For this purpose, the US and Russia developed a joint benchmark model of the Topaz-II reactor. The American and Russian teams performed independent computations for a series of variants representing potential water immersion accidents. Comparison of the MCNP and MCU-2 codes showed somewhat different results both for the absolute values of k{sub eff} and for reactivity effects. Future calculations will be performed to obtain a detailed understanding of the reasons for such discrepancies. For these analyses it will be necessary for the US and Russian teams to exchange neutronic data on Topaz-II physics calculations.

  16. Measurements of miniature ionization chamber currents in the JSI TRIGA Mark II reactor demonstrate the importance of the delayed contribution to the photon field in nuclear reactors

    NASA Astrophysics Data System (ADS)

    Radulović, Vladimir; Fourmentel, Damien; Barbot, Loïc; Villard, Jean-François; Kaiba, Tanja; Gašper, Žerovnik; Snoj, Luka

    2015-12-01

    The characterization of experimental locations of a research nuclear reactor implies the determination of neutron and photon flux levels within, with the best achievable accuracy. In nuclear reactors, photon fluxes are commonly calculated by Monte Carlo simulations but rarely measured on-line. In this context, experiments were conducted with a miniature gas ionization chamber (MIC) based on miniature fission chamber mechanical parts, recently developed by the CEA (French Atomic Energy and Alternative Energies Commission) irradiated in the core of the Jožef Stefan Institute TRIGA Mark II reactor in Ljubljana, Slovenia. The aim of the study was to compare the measured MIC currents with calculated currents based on simulations with the MCNP6 code. A discrepancy of around 50% was observed between the measured and the calculated currents; in the latter taking into consideration only the prompt photon field. Further experimental measurements of MIC currents following reactor SCRAMs (reactor shutdown with rapid insertions of control rods) provide evidence that over 30% of the total measured signal is due to the delayed photon field, originating from fission and activation products, which are untreated in the calculations. In the comparison between the measured and calculated values, these findings imply an overall discrepancy of less than 20% of the total signal which is still unexplained.

  17. Metallic fuels: The EBR-II legacy and recent advances

    SciTech Connect

    Douglas L. Porter; Steven L. Hayes; J. Rory Kennedy

    2012-09-01

    Experimental Breeder ReactorII (EBR-II) metallic fuel was qualified for high burnup to approximately 10 atomic per cent. Subsequently, the electrometallurgical treatment of this fuel was demonstrated. Advanced metallic fuels are now investigated for increased performance, including ultra-high burnup and actinide burning. Advances include additives to mitigate the fuel/cladding chemical interaction and uranium alloys that combine Mo, Ti and Zr to improve alloy performance. The impacts of the advances—on fabrication, waste streams, electrorefining, etc.—are found to be minimal and beneficial. Owing to extensive research literature and computational methods, only a modest effort is required to complete their development.

  18. Decontamination and Decommissioning of the SPERT-II and SPERT-III reactors at the Idaho National Engineering Laboratory

    SciTech Connect

    Hine, R.E.

    1981-02-01

    This report describes the Decontamination and Decommissioning (D and D) of the SPERT-II and SPERT-III reactor facilities performed during the period June through September 1980 at the Idaho National Engineering Laboratory. It includes a detailed description of the D and D accomplished and the post-D and D condition of the reactor facilities. The report also serves to document the radiological condition of the facilities after D and D, the waste volume generated and its disposition, and the project cost and schedule.

  19. Status and perspective of the R&D on ceramic breeder materials for testing in ITER

    NASA Astrophysics Data System (ADS)

    Ying, A.; Akiba, M.; Boccaccini, L. V.; Casadio, S.; Dell'Orco, G.; Enoeda, M.; Hayashi, K.; Hegeman, J. B.; Knitter, R.; van der Laan, J.; Lulewicz, J. D.; Wen, Z. Y.

    2007-08-01

    The main line of ceramic breeder materials research and development is based on the use of the breeder material in the form of pebble beds. At present, there are three candidate pebble materials (Li 4SiO 4, and two forms of Li 2TiO 3) for DEMO reactors that will be used for testing in ITER. This paper reviews the R&D of as-fabricated pebble materials against the blanket performance requirements and makes recommendations on necessary steps toward the qualification of these materials for testing in ITER.

  20. Breeder Spent Fuel Handling (BSFH) cask study for FY83. Final report

    SciTech Connect

    Diggs, J M

    1985-01-01

    This report documents a study conducted to investigate the applicability of existing LWR casks to shipment of long-cooled LMFBR fuel from the Clinch River Breeder Reactor Plant (CRBRP) to the Breeder Reprocessing Engineering Test (BRET) Facility. This study considered a base case of physical constraints of plants and casks, handling capabilities of plants, through-put requirements, shielding requirements due to transportation regulation, and heat transfer capabilities of the cask designs. Each cask design was measured relative to the base case. 15 references, 4 figures, 6 tables.

  1. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  2. High performance inboard shield design for the compact TIBER-II test reactor: Appendix A-2

    SciTech Connect

    El-Guebaly, L.A.; Sviatoslavsky, I.N.

    1987-01-01

    The compactness of the TIBER-II reactor has placed a premium on the design of a high performance inboard shield to protect the inner legs of the toroidal field (TF) coils. The available space for shield is constrained to 48 cm and the use of tungsten is mandatory to protect the magnet against the 1.53 MW/m/sup 2/ neutron wall loading. The primary requirement for the shield is to limit the fast neutron fluence to 10/sup 19/ n/cm/sup 2/. In an optimization study, the performance of various candidate materials for protecting the magnet was examined. The optimum shield consists of a 40 cm thick W layer, followed by an 8 cm thick H/sub 2/O/LiNO/sub 3/ layer. The mechanical design of the shield calls for tungsten blocks within SS stiffened panels. All the coolant channels are vertical with more of them in the front where there is a high heat load. The coolant pressure is 0.2 MPa and the maximum structural surface temperature is <95/sup 0/C. The effects of the detailed mechanical design of the shield and the assembly gaps between the shield sectors on the damage in the magnet were analyzed and peaking factors of approx.2 were found at the hot spots. 2 refs., 6 figs., 2 tabs.

  3. High-resolution neutron powder diffractometer SPODI at research reactor FRM II

    NASA Astrophysics Data System (ADS)

    Hoelzel, M.; Senyshyn, A.; Juenke, N.; Boysen, H.; Schmahl, W.; Fuess, H.

    2012-03-01

    SPODI is a high-resolution thermal neutron diffractometer at the research reactor Heinz Maier-Leibnitz (FRM II) especially dedicated to structural studies of complex systems. Unique features like a very large monochromator take-off angle of 155° and a 5 m monochromator-sample distance in its standard configuration achieve both high-resolution and a good profile shape for a broad scattering angle range. Two dimensional data are collected by an array of 80 vertical position sensitive 3He detectors. SPODI is well suited for studies of complex structural and magnetic order and disorder phenomena at non-ambient conditions. In addition to standard sample environment facilities (cryostats, furnaces, magnet) specific devices (rotatable load frame, cell for electric fields, multichannel potentiostat) were developed. Thus the characterisation of functional materials at in-operando conditions can be achieved. In this contribution the details of the design and present performance of the instrument are reported along with its specifications. A new concept for data reduction using a 2 θ dependent variable height for the intensity integration along the Debye-Scherrer lines is introduced.

  4. Analysis of Experimental Data for High Burnup PWR Spent Fuel Isotopic Validation - Vandellos II Reactor

    SciTech Connect

    Ilas, Germina; Gauld, Ian C

    2011-01-01

    This report is one of the several recent NUREG/CR reports documenting benchmark-quality radiochemical assay data and the use of the data to validate computer code predictions of isotopic composition for spent nuclear fuel, to establish the uncertainty and bias associated with code predictions. The experimental data analyzed in the current report were acquired from a high-burnup fuel program coordinated by Spanish organizations. The measurements included extensive actinide and fission product data of importance to spent fuel safety applications, including burnup credit, decay heat, and radiation source terms. Six unique spent fuel samples from three uranium oxide fuel rods were analyzed. The fuel rods had a 4.5 wt % {sup 235}U initial enrichment and were irradiated in the Vandellos II pressurized water reactor operated in Spain. The burnups of the fuel samples range from 42 to 78 GWd/MTU. The measurements were used to validate the two-dimensional depletion sequence TRITON in the SCALE computer code system.

  5. Simulation of Collimator for Neutron Imaging Facility of TRIGA MARK II PUSPATI Reactor

    NASA Astrophysics Data System (ADS)

    Zin, Muhammad Rawi Mohamed; Jamro, Rafhayudi; Yazid, Khairiah; Hussain, Hishamuddin; Yazid, Hafizal; Ahmad, Megat Harun Al Rashid Megat; Azman, Azraf; Mohamad, Glam Hadzir Patai; Hamzah, Nai'im Syaugi; Abu, Mohamad Puad

    Neutron Radiography facility in TRIGA MARK II PUSPATI reactor is being upgraded to obtain better image resolution as well as reducing exposure time. Collimator and exposure room are the main components have been designed for fabrication. This article focuses on the simulation part that was carried out to obtain the profile of collimated neutron beam by utilizing the neutron transport protocol code in the Monte Carlo N-Particle (MCNP) software. Particular interest is in the selection of materials for inlet section of the collimator. Results from the simulation indicates that a combination of Bismuth and Sapphire, each of which has 5.0 cm length that can significantly filter both the gamma radiation and the fast neutrons. An aperture made of Cadmium with 1.0 cm opening diameter provides thermal neutron flux about 1.8 x108 ncm-2s-1 at the inlet, but reduces to 2.7 x106 ncm-2s-1 at the sample plane. Still the flux obtained is expected to reduces exposure time as well as gaining better image resolution.

  6. Evaluation of Fe(II) oxidation at an acid mine drainage site using laboratory-scale reactors

    NASA Astrophysics Data System (ADS)

    Brown, Juliana; Burgos, William

    2010-05-01

    Acid mine drainage (AMD) is a severe environmental threat to the Appalachian region of the Eastern United States. The Susquehanna and Potomac River basins of Pennsylvania drain to the Chesapeake Bay, which is heavily polluted by acidity and metals from AMD. This study attempted to unravel the complex relationships between AMD geochemistry, microbial communities, hydrodynamic conditions, and the mineral precipitates for low-pH Fe mounds formed downstream of deep mine discharges, such as Lower Red Eyes in Somerset County, PA, USA. This site is contaminated with high concentrations of Fe (550 mg/L), Mn (115 mg/L), and other trace metals. At the site 95% of dissolved Fe(II) and 56% of total dissolved Fe is removed without treatment, across the mound, but there is no change in the concentration of trace metals. Fe(III) oxides were collected across the Red Eyes Fe mound and precipitates were analyzed by X-ray diffraction, electron microscopy and elemental analysis. Schwertmannite was the dominant mineral phase with traces of goethite. The precipitates also contained minor amounts of Al2O3, MgO,and P2O5. Laboratory flow-through reactors were constructed to quantify Fe(II) oxidation and Fe removal over time at terrace and pool depositional facies. Conditions such as residence time, number of reactors in sequence and water column height were varied to determine optimal conditions for Fe removal. Reactors with sediments collected from an upstream terrace oxidized more than 50% of dissolved Fe(II) at a ten hour residence time, while upstream pool sediments only oxidized 40% of dissolved Fe(II). Downstream terrace and pool sediments were only capable of oxidizing 25% and 20% of Fe(II), respectively. Fe(II) oxidation rates measured in the reactors were determined to be between 3.99 x 10-8and 1.94 x 10-7mol L-1s-1. The sediments were not as efficient for total dissolved Fe removal and only 25% was removed under optimal conditions. The removal efficiency for all sediments

  7. Effect of continuously dosing Cu(II) on pollutant removal and soluble microbial products in a sequencing batch reactor.

    PubMed

    Yan, YangWei; Wang, YuWen; Liu, Yan; Liu, Xiang; Yao, ChenChao; Ma, LuMing

    2015-01-01

    The effects of synthetic wastewater that contained 20 mg/L Cu(II) on the removal of organic pollutants in a sequencing batch reactor were investigated. Results of continuous 20 mg/L Cu(II) exposure for 120 days demonstrated that the chemical oxygen demand (COD) removal efficiency decreased to 42% initially, followed by a subsequent gradual recovery, which peaked at 78% by day 97. Effluent volatile fatty acid (VFA) concentration contributed 67 to 89% of the influent COD in the experimental reactor, which indicated that the degradation of the organic substances ceased at the VFA production step. Meanwhile, the varieties of soluble microbial products (SMP) content and main components (protein, polysaccharide, and DNA) were discussed to reveal the response of activated sludge to the toxicity of 20 mg/L Cu(II). The determination of Cu(II) concentrations in extracellular polymeric substances (EPS) and SMP throughout the experiment indicated an inverse relationship between extracellular Cu(II) concentration and COD removal efficiency. PMID:26524458

  8. Modelling of tritium transport in a pin-type solid breeder blanket

    SciTech Connect

    Martin, R.; Ghoniem, N.M.

    1986-02-01

    This study supplements a larger study of a solid breeder blanket design featuring lithium ceramic pins. This aspect of the study looks at tritium transport, release, and inventory within this blanket design. Li/sub 2/O and ..gamma..-LiAlO/sub 2/ are the two primary candidates for ceramic solid breeders. ..gamma..-LiAlO/sub 2/ was chosen for this blanket design due to its higher structural stability. Analysis of tritium behavior in solid breeder blankets is of great importance due to its impact on several critical issues: the generation of an adequate amount of fusion fuel, the safety-related issue of keeping radioactive blanket inventories as low as possible, and the release, purge, and economical processing of the bred tritium without undue contamination of the coolant and other reactor structures.

  9. Nuclear reactor composite fuel assembly

    DOEpatents

    Burgess, Donn M.; Marr, Duane R.; Cappiello, Michael W.; Omberg, Ronald P.

    1980-01-01

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  10. Preliminary Design of a Helium-Cooled Ceramic Breeder Blanket for CFETR Based on the BIT Concept

    NASA Astrophysics Data System (ADS)

    Ma, Xuebin; Liu, Songlin; Li, Jia; Pu, Yong; Chen, Xiangcun

    2014-04-01

    CFETR is the “ITER-like” China fusion engineering test reactor. The design of the breeding blanket is one of the key issues in achieving the required tritium breeding radio for the self-sufficiency of tritium as a fuel. As one option, a BIT (breeder insider tube) type helium cooled ceramic breeder blanket (HCCB) was designed. This paper presents the design of the BIT—HCCB blanket configuration inside a reactor and its structure, along with neutronics, thermo-hydraulics and thermal stress analyses. Such preliminary performance analyses indicate that the design satisfies the requirements and the material allowable limits.