Science.gov

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. 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.

  14. 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.

  15. 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.

  16. 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)

  17. 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.

  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. 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.

  3. 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…

  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. 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.

  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. 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.

  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. 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.

  6. 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.

  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. 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)

  15. 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.

  16. 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.

  17. 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.

  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. 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.

  2. 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.

  3. 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.

  4. 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.

  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

    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.

  5. 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.

  6. 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.

  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. 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.

  16. 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)

  17. 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.

  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

    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

  12. 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.

  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. 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.

  9. 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

  10. 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.

  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-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.

  3. 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.

  4. 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.

  5. 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.

  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.

  11. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

    A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

  12. Effects of adsorbents and copper(II) on activated sludge microorganisms and sequencing batch reactor treatment process.

    PubMed

    Ong, S A; Lim, P E; Seng, C E

    2003-10-31

    Wastewater treatment systems employing simultaneous adsorption and biodegradation processes have proven to be effective in treating toxic pollutants present in industrial wastewater. The objective of this study is to evaluate the effect of Cu(II) and the efficacy of the powdered activated carbon (PAC) and activated rice husk (ARH) in reducing the toxic effect of Cu(II) on the activated sludge microorganisms. The ARH was prepared by treatment with concentrated nitric acid for 15 h at 60-65 degrees C. The sequencing batch reactor (SBR) systems were operated with FILL, REACT, SETTLE, DRAW and IDLE modes in the ratio of 0.5:3.5:1:0.75:0.25 for a cycle time of 6 h. The Cu(II) and COD removal efficiency were 90 and 85%, respectively, in the SBR system containing 10 mg/l Cu(II) with the addition of 143 mg/l PAC or 1.0 g PAC per cycle. In the case of 715 mg/l ARH or 5.0 g ARH per cycle addition, the Cu(II) and COD removal efficiency were 85 and 92%, respectively. ARH can be used as an alternate adsorbent to PAC in the simultaneous adsorption and biodegradation wastewater treatment process for the removal of Cu(II). The specific oxygen uptake rate (SOUR) and kinetic studies show that the addition of PAC and ARH reduce the toxic effect of Cu(II) on the activated sludge microorganisms. PMID:14573344

  13. Effect of Cu(II) shock loads on shortcut biological nitrogen removal in a hybrid biofilm nitrogen removal reactor.

    PubMed

    Yin, Jun; Xu, Hengjuan; Shen, Dongsheng; Wang, Kun; Lin, Ying

    2015-06-01

    The effect of Cu(II) shock loads on shortcut biological nitrogen removal during a continuous-flow anoxic/aerobic process was investigated using a hybrid biofilm nitrogen removal reactor. The results demonstrated that [Formula: see text]-N removal was not affected by any Cu(II) shock loads, but TN removal was inhibited by Cu(II) of shock loads of 2 and 5 mg/L, and the performance could not be recovered at 5 mg/L. Furthermore, the TN removal pathway also changed in response to Cu(II) concentrations of 2 and 5 mg/L. Denitrification is more sensitive to Cu(II) shock in SBNR processes. Examination of amoA communities using quantitative PCR showed that the abundance of AOB in the aerobic tank decreased after Cu(II) shock with 5 mg/L, which supported the observed changes in [Formula: see text]-N removal efficiency. The abundance of denitrification genes declined obviously at Cu(II) concentrations of 2 and 5 mg/L, which explained the decreased TN removal efficiency at those concentrations. PMID:25833010

  14. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  15. Alternative energy source II; Proceedings of the Second Miami International Conference, Miami Beach, Fla., December 10-13, 1979. Volume 6 - Nuclear energy

    NASA Astrophysics Data System (ADS)

    Veziroglu, T. N.

    This volume examines conventional nuclear energy, breeder reactors, and thermonuclear energy. The particular papers presented consider current developments in nuclear breeder technology, fusion-driven fissile fuel breeder systems, and the fusion fission hybrid reactor. The implications of nuclear energy utilization in the Phillipines and the internationally safeguarded atomic fuel exchanger center for the Asian-Pacific basin are also discussed.

  16. Recent operating experiences and programs at EBR-II

    SciTech Connect

    Lentz, G.L.

    1984-01-01

    Experimental Breeder Reactor No. II (EBR-II) is a pool-type, unmoderated, sodium-cooled reactor with a design power of 62.5 MWt and an electrical generation capability of 20 MW. It has been operated by Argonne National Laboratory for the US government for almost 20 years. During that time, it has operated safely and has demonstrated stable operating characteristics, high availability, and excellent performance of its sodium components. The 20 years of operating experience of EBR-II is a valuable resource to the nuclear community for the development and design of future LMFBR's. Since past operating experience has been extensively reported, this report will focus on recent programs and events.

  17. Expert system applications in support of system diagnostics and prognostics at EBR-II

    SciTech Connect

    Lehto, W.K.; Gross, K.C.; Argonne National Lab., IL )

    1989-01-01

    Expert systems have been developed to aid in the monitoring and diagnostics of the Experimental Breeder Reactor-II (EBR-II) at the Idaho National Engineering Laboratory (INEL) in Idaho Falls, Idaho. Systems have been developed for failed fuel surveillance and diagnostics and reactor coolant pump monitoring and diagnostics. A third project is being done jointly by ANL-W and EG G Idaho to develop a transient analysis system to enhance overall plant diagnostic and prognostic capability. The failed fuel surveillance and diagnosis system monitors, processes, and interprets information from nine key plant sensors. It displays to the reactor operator diagnostic information needed to make proper decisions regarding technical specification conformance during reactor operation with failed fuel. 8 refs., 9 figs., 2 tabs.

  18. The UF6 Breeder - A solution to the problems of nuclear power

    NASA Technical Reports Server (NTRS)

    Williams, J. R.; Clement, J. D.; Rust, J. A.

    1975-01-01

    One of the major advantages of uranium hexafluoride reactors for power generation is the simplified fuel reprocessing scheme which the gaseous fuel makes possible. Critical experiments related to the development of the reactors for electric power generation are discussed along with UF6 breeder reactor studies. Previous energy conversion studies are reported, taking into account gas turbine power plants, thermionic conversion, and MHD conversion. Thermodynamic cycle analyses show that high efficiencies can be achieved using UF6 as the working fluid for Rankine or Brayton cycles without requiring excessive temperatures.

  19. Proceedings of the NEACRP/IAEA Specialists meeting on the international comparison calculation of a large sodium-cooled fast breeder reactor at Argonne National Laboratory on February 7-9, 1978

    SciTech Connect

    LeSage, L.G.; McKnight, R.D.; Wade, D.C.; Freese, K.E.; Collins, P.J.

    1980-08-01

    The results of an international comparison calculation of a large (1250 MWe) LMFBR benchmark model are presented and discussed. Eight reactor configurations were calculated. Parameters included with the comparison were: eigenvalue, k/sub infinity/, neutron balance data, breeding reaction rate ratios, reactivity worths, central control rod worth, regional sodium void reactivity, core Doppler and effective delayed neutron fraction. Ten countries participated in the comparison, and sixteen solutions were contributed. The discussion focuses on the variation in parameter values, the degree of consistency among the various parameters and solutions, and the identification of unexpected results. The results are displayed and discussed both by individual participants and by groupings of participants (e.g., results from adjusted data sets versus non-adjusted data sets).

  20. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

  1. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    PubMed

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor. PMID:27552124

  2. Design of a 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.

    1985-02-01

    A new conceptual blanket design for a fusion reactor produces fissile material for fission power plants. Fission is suppressed by using beryllium, rather than uranium, to multiply neutrons and also by minimizing the fissile inventory. The molten-salt breeding media (LiF + BeF/sub 2/ + TghF/sub 4/) is circulated through the blanket and on to the online processing system where /sup 233/U and tritium are continuously removed. Helium cools the blanket including the steel pipes containing the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion rate by molten salt. We estimate the breeder, having 3000 MW of fusion power, produces 6400 kg of /sup 233/U per year, which is enough to provide make up for 20 GWe of LWR per year (or 14 LWR plants of 4440 MWt) or twice that many HTGRs or CANDUs. Safety is enhanced because the afterheat is low and the 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 an LWR of the same power. The estimated present value cost of the /sup 2/anumber/sup 3/U produced is $40/g if utility financed or $16/g if government financed.

  3. Ceramic breeder research and development: progress and focus

    NASA Astrophysics Data System (ADS)

    van der Laan, J. G.; Kawamura, H.; Roux, N.; Yamaki, D.

    2000-12-01

    The world-wide efforts on ceramic breeder materials in the last two years concerned Li2O, Li4SiO4, Li2TiO3 and Li2ZrO3, with a clear emphasis on the development of Li2TiO3. Pebble-manufacturing processes have been developed up to a 10 kg scale. Characterisation of materials has advanced. A jump-wise progress is observed in the characterisation of pebble-beds, in particular of their thermo-mechanical behaviour. Thermal property data are still limited. A number of breeder materials have been or are being irradiated in material test reactors like HFR and JMTR. The EXOTIC-8 series of in-pile experiments is a major source of tritium release data. This paper discusses the technical advancements and proposes a focus for further research and development (R&D) : pebble-bed mechanical and thermal behaviour and its interactions with the blanket structure as a function of temperature, burn-up, irradiation dose and time; tritium release and retention properties; determination of the key factors limiting blanket life.

  4. Operating experience of the EBR-II intermediate heat exchanger and the steam generator system

    SciTech Connect

    Buschman, H.W.; Longua, K.J.; Penney, W.H.

    1983-01-01

    Experimental Breeder Reactor-II (EBR-II) is an experimental liquid metal fast breeder reactor located at the Idaho National Engineering Laboratory. It consists of an unmoderated, heterogeneous, sodium-cooled reactor with a nominal thermal power output of 62.5 MW; an intermediate closed loop of secondary sodium coolant; and a steam plant that produces 20 MW of electrical power through a conventional turbine generator. The EBR-II heat transport system continues to operate satisfactorily after 18 years. This represents about 89,000 hours of steaming, which results in a total integrated thermal power production of about 215,000 MWd. In this time, the steam generator has experienced over 580 plant startups and 349 reactor scrams. The plant capacity factor for the past five years has been in excess of 70%, and in fact has averaged almost 60% over the last thirteen years. This excellent record is partly attributable to the trouble-free operation of the steam generator which, aside from an initial construction tube-to-tubesheet weld defect, has had a plant availability of 100%.

  5. Uranium resources and their implications for fission breeder and fusion hybrid development

    SciTech Connect

    Max, C.E.

    1984-05-15

    Present estimates of uranium resources and reserves in the US and the non-Communist world are reviewed. The resulting implications are considered for two proposed breeder technologies: the liquid metal fast breeder reactor (LMFBR) and the fusion hybrid reactor. Using both simple arguments and detailed scenarios from the published literature, conditions are explored under which the LMFBR and fusion hybrid could respectively have the most impact, considering both fuel-supply and economic factors. The conclusions emphasize strong potential advantages of the fusion hybrid, due to its inherently large breeding rate. A discussion is presented of proposed US development strategies for the fusion hybrid, which at present is far behind the LMFBR in its practical application and maturity.

  6. Development of Underwater Laser Cladding and Underwater Laser Seal Welding Techniques for Reactor Components (II)

    SciTech Connect

    Masataka Tamura; Shohei Kawano; Wataru Kouno; Yasushi Kanazawa

    2006-07-01

    Stress corrosion cracking (SCC) is one of the major reasons to reduce the reliability of aged reactor components. Toshiba has been developing underwater laser welding onto surface of the aged components as maintenance and repair techniques. Because most of the reactor internal components to apply this underwater laser welding technique have 3-dimensional shape, effect of welding positions and welded shapes are examined and presented in this report. (authors)

  7. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    SciTech Connect

    Benedict, R.W.; Goff, K.M.

    1993-01-01

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

  8. Transuranic material recovery in the Integral Fast Reactor fuel cycle demonstration

    SciTech Connect

    Benedict, R.W.; Goff, K.M.

    1993-03-01

    The Integral Fast Reactor is an innovative liquid metal reactor concept that is being developed by Argonne National Laboratory. It takes advantage of the properties of metallic fuel and liquid metal cooling to offer significant improvements in reactor safety, operation, fuel cycle economics, environmental protection, and safeguards. The plans for demonstrating the IFR fuel cycle, including its waste processing options, by processing irradiated fuel from the Experimental Breeder Reactor-II fuel in its associated Fuel Cycle Facility have been developed for the first refining series. This series has been designed to provide the data needed for the further development of the IFR program. An important piece of the data needed is the recovery of TRU material during the reprocessing and waste operations.

  9. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    SciTech Connect

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  10. Design and technology development of solid breeder blanket cooled by supercritical water in Japan

    NASA Astrophysics Data System (ADS)

    Enoeda, M.; Kosaku, Y.; Hatano, T.; Kuroda, T.; Miki, N.; Honma, T.; Akiba, M.; Konishi, S.; Nakamura, H.; Kawamura, Y.; Sato, S.; Furuya, K.; Asaoka, Y.; Okano, K.

    2003-12-01

    This paper presents results of conceptual design activities and associated R&D of a solid breeder blanket system for demonstration of power generation fusion reactors (DEMO blanket) cooled by supercritical water. The Fusion Council of Japan developed the long-term research and development programme of the blanket in 1999. To make the fusion DEMO reactor more attractive, a higher thermal efficiency of more than 40% was strongly recommended. To meet this requirement, the design of the DEMO fusion reactor was carried out. In conjunction with the reactor design, a new concept of a solid breeder blanket cooled by supercritical water was proposed and design and technology development of a solid breeder blanket cooled by supercritical water was performed. By thermo-mechanical analyses of the first wall, the tresca stress was evaluated to be 428 MPa, which clears the 3Sm value of F82H. By thermal and nuclear analyses of the breeder layers, it was shown that a net TBR of more than 1.05 can be achieved. By thermal analysis of the supercritical water power plant, it was shown that a thermal efficiency of more than 41% is achievable. The design work included design of the coolant flow pattern for blanket modules, module structure design, thermo-mechanical analysis and neutronics analysis of the blanket module, and analyses of the tritium inventory and permeation. Preliminary integration of the design of a solid breeder blanket cooled by supercritical water was achieved in this study. In parallel with the design activities, engineering R&D was conducted covering all necessary issues, such as development of structural materials, tritium breeding materials, and neutron multiplier materials; neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was

  11. BEATRIX-II: In situ tritium test

    SciTech Connect

    Baker, D.E. ); Kuraswa, T. ); Miller, J.M. . Chalk River Nuclear Labs.); Slagle, O.D. )

    1990-01-01

    The BEATRIX-II irradiation experiment is an in-situ tritium release experiment being carried out in the Fast Flux Test Facility (FFTF) reactor to evaluate the tritium release characteristics of fusion solid breeder materials. A sophisticated tritium gas handling system has been developed to continuously monitor the tritium recovery from the specimens and facilitate tritium removal from the experiment's sweep gas flow stream. The in-situ recovery experiment accommodates two different in-reactor specimen canisters with individual gas streams and temperature monitoring/control. Ionization chambers have been specifically designed to respond to the rapid changes in the tritium release rate at the anticipated tritium concentrations. Two ceramic electrolysis cells have proved effective in reducing the moisture in the gas streams to hydrogen/tritium. A tritium getter system, capable of reducing the tritium level by a factor greater than 4000, is used to reduce the tritium in the sweep gas to a level acceptable for release.

  12. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

  13. Nuclear reactor control

    SciTech Connect

    Ingham, R.V.

    1980-01-01

    A liquid metal cooled fast breeder nuclear reactor has power setback means for use in an emergency. On initiation of a trip-signal a control rod is injected into the core in two stages, firstly, by free fall to effect an immediate power-set back to a safe level and, secondly, by controlled insertion. Total shut-down of the reactor under all emergencies is avoided. 4 claims.

  14. Long-lived activation products in TRIGA Mark II research reactor concrete shield: calculation and experiment

    NASA Astrophysics Data System (ADS)

    Žagar, Tomaž; Božič, Matjaž; Ravnik, Matjaž

    2004-12-01

    In this paper, a process of long-lived activity determination in research reactor concrete shielding is presented. The described process is a combination of experiment and calculations. Samples of original heavy reactor concrete containing mineral barite were irradiated inside the reactor shielding to measure its long-lived induced radioactivity. The most active long-lived (γ emitting) radioactive nuclides in the concrete were found to be 133Ba, 60Co and 152Eu. Neutron flux, activation rates and concrete activity were calculated for actual shield geometry for different irradiation and cooling times using TORT and ORIGEN codes. Experimental results of flux and activity measurements showed good agreement with the results of calculations. Volume of activated concrete waste after reactor decommissioning was estimated for particular case of Jožef Stefan Institute TRIGA reactor. It was observed that the clearance levels of some important long-lived isotopes typical for barite concrete (e.g. 133Ba, 41Ca) are not included in the IAEA and EU basic safety standards.

  15. Implementation of k0-INAA standardisation at ITU TRIGA Mark II research reactor, Turkey based on k0-IAEA software

    NASA Astrophysics Data System (ADS)

    Esen, Ayse Nur; Haciyakupoglu, Sevilay

    2016-02-01

    The purpose of this study is to test the applicability of k0-INAA method at the Istanbul Technical University TRIGA Mark II research reactor. The neutron spectrum parameters such as epithermal neutron flux distribution parameter (α), thermal to epithermal neutron flux ratio (f) and thermal neutron flux (φth) were determined at the central irradiation channel of the ITU TRIGA Mark II research reactor using bare triple-monitor method. HPGe detector calibrations and calculations were carried out by k0-IAEA software. The α, f and φth values were calculated to be -0.009, 15.4 and 7.92·1012 cm-2 s-1, respectively. NIST SRM 1633b coal fly ash and intercomparison samples consisting of clay and sandy soil samples were used to evaluate the validity of the method. For selected elements, the statistical evaluation of the analysis results was carried out by z-score test. A good agreement between certified/reported and experimental values was obtained.

  16. Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

    SciTech Connect

    Wang, C.Y.

    1993-06-01

    This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts` ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

  17. Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

    SciTech Connect

    Wang, C.Y.

    1993-01-01

    This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts' ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

  18. Continuous production of tritium in an isotope-production reactor with a separate circulation system

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium is allowed to flow through the reactor in separate loops in order to facilitate the production and removal of tritium.

  19. Modeling of the HiPco process for carbon nanotube production. II. Reactor-scale analysis

    NASA Technical Reports Server (NTRS)

    Gokcen, Tahir; Dateo, Christopher E.; Meyyappan, M.

    2002-01-01

    The high-pressure carbon monoxide (HiPco) process, developed at Rice University, has been reported to produce single-walled carbon nanotubes from gas-phase reactions of iron carbonyl in carbon monoxide at high pressures (10-100 atm). Computational modeling is used here to develop an understanding of the HiPco process. A detailed kinetic model of the HiPco process that includes of the precursor, decomposition metal cluster formation and growth, and carbon nanotube growth was developed in the previous article (Part I). Decomposition of precursor molecules is necessary to initiate metal cluster formation. The metal clusters serve as catalysts for carbon nanotube growth. The diameter of metal clusters and number of atoms in these clusters are some of the essential information for predicting carbon nanotube formation and growth, which is then modeled by the Boudouard reaction with metal catalysts. Based on the detailed model simulations, a reduced kinetic model was also developed in Part I for use in reactor-scale flowfield calculations. Here this reduced kinetic model is integrated with a two-dimensional axisymmetric reactor flow model to predict reactor performance. Carbon nanotube growth is examined with respect to several process variables (peripheral jet temperature, reactor pressure, and Fe(CO)5 concentration) with the use of the axisymmetric model, and the computed results are compared with existing experimental data. The model yields most of the qualitative trends observed in the experiments and helps to understanding the fundamental processes in HiPco carbon nanotube production.

  20. A US perspective on fast reactor fuel fabrication technology and experience part I: metal fuels and assembly design

    NASA Astrophysics Data System (ADS)

    Burkes, Douglas E.; Fielding, Randall S.; Porter, Douglas L.; Crawford, Douglas C.; Meyer, Mitchell K.

    2009-06-01

    This paper is part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II (EBR-II) and the Fast Flux Test Facility (FFTF). Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated in a considerable amount of research that resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, 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.

  1. Analyses of various options for the breeder fuel cycle in the United States

    SciTech Connect

    Burch, W.D.; Stradley, J.G.; Yook, H.R.; Lerch, R.E.

    1985-01-01

    The US Department of Energy (DOE) has established a program to develop innovative liquid metal reactor (LMR) designs to assist in developing US future reactor strategy. This paper describes studies in progress to examine various fuel cycle strategies that relate to the reactor strategy. Three potential fuel cycle options that focus on supporting an initial 1300-MWe reactor station have been defined: completion and utilization of the Breeder Reprocessing Engineering Test/Secure Automated Fabrication (BRET/SAF) in the Fuels and Materials Examination Facility (FMEF); a co-located fuel cycle facility; and delayed closure of the fuel cycle for five to ten years. It appears feasible to increase the capacity of the original BRET design and SAF in the FMEF to accommodate the projected output (up to 35 MTHM/year) from the 1300-MWe liquid-metal concepts under study. Working with the reactor manufacturers, criteria were developed for a small fuel cycle facility co-located at a utility reactor site. The requirements considered the need to be able to support as little as approximately 400 MWe to as much as 35 MTHM/year reprocessing capability. A rough order-of-magnitude capital cost estimate ($300 million) was developed. Plans developed within the US Consolidated Management Office for an initial reactor project have envisioned that a cost savings could be realized by delaying closure of the fuel cycle as long as supplies of plutonium could be obtained relatively inexpensively. This might prove to be only five to ten years, but even that period might be long enough for the fuel cycle costs to be spread over more than one reactor rather than loaded on the initial project. This concept is being explored as is the question of the future coupling of a light water reactor (LWR) reprocessing industry for plutonium supply to breeder recycle. 12 figs.

  2. A transient overpower experiment in EBR-II

    SciTech Connect

    Herzog, J.P.; Tsai, H.; Dean, E.M.; Aoyama, T.; Yamamoto, K.

    1994-03-01

    The TOPI-IE test was a transient overpower test on irradiate mixed-oxide fuel pins in the Experimental Breeder Reactor-II (EBR-II). The test, the fifth in a series, was part of a cooperative program between the US Department of Energy and the Power Reactor and Nuclear Fuel Development Corporation of Japan to conduct operational transient testing on mixed-oxide fuel pins in the metal-fueled EBR-II. The principle objective of the TOPI-1E test was to assess breaching margins for irradiated mixed-oxide fuel pins over the Plant Protection System (PPS) thresholds during a slow, extended overpower transient. This paper describes the effect of the TOPI-1E experiment on reactor components and the impact of the experiment on the long-term operability of the reactor. The paper discusses the role that SASSYS played in the pre-test safety analysis of the experiment. The ability of SASSYS to model transient overpower events is detailed by comparisons of data from the experiment with computed reactor variables from a SASSYS post-test simulation of the experiment.

  3. Alternative reproductive tactics in female striped mice: Solitary breeders have lower corticosterone levels than communal breeders.

    PubMed

    Hill, Davina L; Pillay, Neville; Schradin, Carsten

    2015-05-01

    Alternative reproductive tactics (ARTs), where members of the same sex and population show distinct reproductive phenotypes governed by decision-rules, have been well-documented in males of many species, but are less well understood in females. The relative plasticity hypothesis (RPH) predicts that switches between plastic ARTs are mediated by changes in steroid hormones. This has received much support in males, but little is known about the endocrine control of female ARTs. Here, using a free-living population of African striped mice (Rhabdomys pumilio) over five breeding seasons, we tested whether females following different tactics differed in corticosterone and testosterone levels, as reported for male striped mice using ARTs, and in progesterone and oestrogen, which are important in female reproduction. Female striped mice employ three ARTs: communal breeders give birth in a shared nest and provide alloparental care, returners leave the group temporarily to give birth, and solitary breeders leave to give birth and do not return. We expected communal breeders and returners to have higher corticosterone, owing to the social stress of group-living, and lower testosterone than solitary breeders, which must defend territories alone. Solitary breeders had lower corticosterone than returners and communal breeders, as predicted, but testosterone and progesterone did not differ between ARTs. Oestrogen levels were higher in returners (measured before leaving the group) than in communal and solitary breeders, consistent with a modulatory role. Our study demonstrates hormonal differences between females following (or about to follow) different tactics, and provides the first support for the RPH in females. PMID:25828632

  4. An analysis of thermionic space nuclear reactor power system: II. Merits of using safety drums for backup control

    SciTech Connect

    El-Genk, M.S.; Huimin Xue )

    1993-01-10

    An analysis is performed to investigate the merits of using the TOPAZ-II safety drums for a backup control to prevent a reactivity excursion, stabilize the reactor, and achieve steady-state power operation, following a severe hypothetical reactivity initiated accident (RIA). Such an RIA is assumed to occur during the system start-up in orbit due to a malfunction of the drive mechanism of the control drums, causing the nine drums to accidentally rotate the full 180[degree] outward. Results show that an immediate, inward rotation of the three safety drums to an angle of 80[degree] will shutdown the reactor, however, a delay time of 10 s will not only prevents a reactivity excursion, but also enables operating the reactor at a steady-state thermal power of about 33.3 kW (0.9 kW per TFE). Conversely, when the immediate rotation of the safety drums is to a larger angle of 100[degree], a steady-state operation at about 37 kW can be achieved, but a delay of 10 s causes a reactivity excursion and overheating of the TFEs. It is therefor concluded that, should the drive mechanism be modified to enable rotating the safety drums for TOPAZ-II reactor at variable speeds of and below 22.5[degree]/s, the three safety drums could be used successfully for a backup control, following an RIA. However, since the reactivity worth of the three safety drums is only $2.0, the maximum steady-state electric power achievable for the system is limited to approximately 0.25 kW, at which the fission power is about 37 kW and the emitter temperature is approximaely 1500 K. To alleviate such a limitation and enable operation at nominal design conditions (fission power of about 107 kW or a system's total electric power of 5.6 kW), the reactivity worth of the safety drums would have to be increased by at least $0.24.

  5. The cost of tritium production in a fusion reactor

    SciTech Connect

    Wittenberg, L.J. . Fusion Technology Inst.)

    1992-03-01

    In this paper, a computational model is presented in order to assess the cost of tritium breeding in a fusion power reactor. This model compares the differential cost of the Li-bearing breeder blanket with that of a steel shield and adds the loss of revenue due to the lower energy multiplication of the breeder blanket compared to the steel shield. The cost of tritium production ranges from $215-$300/g for a simple breeder up to $1420/g for a high temperature breeder.

  6. Neutron dosimetry and damage calculations for the TRIGA MARK-II reactor in Vienna

    NASA Astrophysics Data System (ADS)

    Weber, H. W.; Böck, H.; Unfried, E.; Greenwood, L. R.

    1986-02-01

    In order to improve the source characterization of the reactor, especially for recent irradiation experiments in the central irradiation thimble, neutron activation experiments were made on 16 nuclides and the neutron flux spectrum was adjusted using the computer code STAY'SL. The results for the total, thermal and fast neutron flux density at a reactor power of 250 kW are as follows: 2.1 × 10 17, 6.1 × 10 16 ( E < 0.55 eV), 7.6 × 10 16 ( E > 0.1 MeV) and 4.0 × 10 16 ( E > 1 MeV) m -2 s -1. respectively. Calculated damage energy cross sections and gas production rates are presented for selected elements.

  7. Computational and experimental prediction of dust production in pebble bed reactors, Part II

    SciTech Connect

    Mie Hiruta; Gannon Johnson; Maziar Rostamian; Gabriel P. Potirniche; Abderrafi M. Ougouag; Massimo Bertino; Louis Franzel; Akira Tokuhiro

    2013-10-01

    This paper is the continuation of Part I, which describes the high temperature and high pressure helium environment wear tests of graphite–graphite in frictional contact. In the present work, it has been attempted to simulate a Pebble Bed Reactor core environment as compared to Part I. The experimental apparatus, which is a custom-designed tribometer, is capable of performing wear tests at PBR relevant higher temperatures and pressures under a helium environment. This environment facilitates prediction of wear mass loss of graphite as dust particulates from the pebble bed. The experimental results of high temperature helium environment are used to anticipate the amount of wear mass produced in a pebble bed nuclear reactor.

  8. NO removal in continuous BioDeNOx reactors: Fe(II)EDTA2- regeneration, biomass growth, and EDTA degradation.

    PubMed

    van der Maas, Peter; van den Brink, Paula; Utomo, Sudarno; Klapwijk, Bram; Lens, Piet

    2006-06-20

    BioDeNOx is a novel technique for NOx removal from industrial flue gases. In principle, BioDeNOx is based on NO absorption into an aqueous Fe(II)EDTA2- solution combined with biological regeneration of that scrubber liquor in a bioreactor. The technical and economical feasibility of the BioDeNOx concept is strongly determined by high rate biological regeneration of the aqueous Fe(II)EDTA2- scrubber liquor and by EDTA degradation. This investigation deals with the Fe(II)EDTA2- regeneration capacity and EDTA degradation in a lab-scale BioDeNOx reactor (10-20 mM Fe(II)EDTA2-, pH 7.2 +/- 0.2, 55 degrees C), treating an artificial flue gas (1.5 m3/h) containing 60-155 ppm NO and 3.5-3.9% O2. The results obtained show a contradiction between the optimal redox state of the aqueous FeEDTA solution for NO absorption and the biological regeneration. A low redox potential (below -150 mV vs. Ag/AgCl) is needed to obtain a maximal NO removal efficiency from the gas phase via Fe(II)EDTA2- absorption. Fe(III)EDTA- reduction was found to be too slow to keep all FeEDTA in the reduced state. Stimulation of Fe(III)EDTA- reduction via periodical sulfide additions (2 mM spikes twice a week for the conditions applied in this study) was found to be necessary to regenerate the Fe(II)EDTA2- scrubber liquor and to achieve stable operation at redox potentials below -150 mV (pH 7.2 +/- 0.2). However, redox potentials of below -200 mV should be avoided since sulfide accumulation is unwanted because it is toxic for NO reduction. Very low values for biomass growth rate and yield, respectively, 0.043/d and 0.009 mg protein per mg ethanol, were observed. This might be due to substrate limitations, that is the electron acceptors NO and presumably polysulfide, or to physiological stress conditions induced by the EDTA rich medium or by radicals formed in the scrubber upon the oxidation of Fe(II)EDTA2- by oxygen present in the flue gas. Radicals possibly also induce EDTA degradation, which occurs at a

  9. Modeling of the HiPco process for carbon nanotube production. II. Reactor-scale analysis.

    PubMed

    Gökçen, Tahir; Dateo, Christopher E; Meyyappan, M

    2002-10-01

    The high-pressure carbon monoxide (HiPco) process, developed at Rice University, has been reported to produce single-walled carbon nanotubes from gas-phase reactions of iron carbonyl in carbon monoxide at high pressures (10-100 atm). Computational modeling is used here to develop an understanding of the HiPco process. A detailed kinetic model of the HiPco process that includes of the precursor, decomposition metal cluster formation and growth, and carbon nanotube growth was developed in the previous article (Part I). Decomposition of precursor molecules is necessary to initiate metal cluster formation. The metal clusters serve as catalysts for carbon nanotube growth. The diameter of metal clusters and number of atoms in these clusters are some of the essential information for predicting carbon nanotube formation and growth, which is then modeled by the Boudouard reaction with metal catalysts. Based on the detailed model simulations, a reduced kinetic model was also developed in Part I for use in reactor-scale flowfield calculations. Here this reduced kinetic model is integrated with a two-dimensional axisymmetric reactor flow model to predict reactor performance. Carbon nanotube growth is examined with respect to several process variables (peripheral jet temperature, reactor pressure, and Fe(CO)5 concentration) with the use of the axisymmetric model, and the computed results are compared with existing experimental data. The model yields most of the qualitative trends observed in the experiments and helps to understanding the fundamental processes in HiPco carbon nanotube production. PMID:12908292

  10. Random vibration analysis of the Topaz-II nuclear reactor power system. Master`s thesis

    SciTech Connect

    Campbell, S.E.

    1995-06-01

    The TOPAZ-II Ya-21U is one of six Russian made space nuclear power systems which is based on theomionic power conversion. The U.S. is presently analyzing TOPAZ-II to determine the reliability and feasibility of using this system. A structural analysis test was conducted on the TOPAZ unit in May 1993 to provide data from which modal parameters could be identified. This test showed the fundamental frequency to be 10.5 Hz, yet the test results that the Russians conducted identified a fundamental frequency of 5 Hz. Another finite element model was created incorporating new developments in TOPAZ-II and modifications to the finite element model to better simulate the mass properties of the TOPAZ-II2. A second structural analysis test was conducted on the TOPAZ unit 06-09 September 1994. This thesis focuses on the random vibration analysis of the TOPAZ-II Ya-2lU utilizing the most recent test results and the Master Series (updated version) I-DEAS software. The modal respose of the model and simulated random vibration tests were within 8.33%. This model is a feasible tool which can be used to analyze the TOPAZ unit without testing the unit to fatigue.

  11. Current experimental activities for solid breeder development

    SciTech Connect

    Johnson, C.E.; Hollenberg, G.W.; Roux, N.; Watanabe, H.

    1988-01-01

    The current data base for ceramic breeder materials does not exhibit any negative features as regards to thermophysical, mechanical, and irradiation behavior. All candidate materials show excellent stability for irradiation testing to 3% burnup. In-situ tritium recovery tests show very low tritium inventories for all candidates. Theoretical models are being developed to accurately predict real time release rates. Fabrication of kilogram quantities of materials has been achieved and technology is available for further scale-up.

  12. Current status of the Run-Beyond-Cladding Breach (RBCB) tests for the Integral Fast Reactor (IFR). Metallic Fuels Program

    SciTech Connect

    Batte, G.L.; Pahl, R.G.; Hofman, G.L.

    1993-09-01

    This paper describes the results from the Integral Fast Reactor (IFR) metallic fuel Run-Beyond-Cladding-Breach (RBCB) experiments conducted in the Experimental Breeder Reactor II (EBR-II). Included in the report are scoping test results and the data collected from the prototypical tests as well as the exam results and discussion from a naturally occurring breach of one of the lead IFR fuel tests. All results showed a characteristic delayed neutron and fission gas release pattern that readily allows for identification and evaluation of cladding breach events. Also, cladding breaches are very small and do not propagate during extensive post breach operation. Loss of fuel from breached cladding was found to be insignificant. The paper will conclude with a brief description of future RBCB experiments planned for irradiation in EBR-II.

  13. Do avian cooperative breeders live longer?

    PubMed Central

    Beauchamp, Guy

    2014-01-01

    Cooperative breeding is not common in birds but intriguingly over-represented in several families, suggesting that predisposing factors, similar ecological constraints or a combination of the two facilitate the evolution of this breeding strategy. The life-history hypothesis proposes that cooperative breeding is facilitated by high annual survival, which increases the local population and leads to a shortage of breeding opportunities. Clutch size in cooperative breeders is also expected to be smaller. An earlier comparative analysis in a small sample of birds supported the hypothesis but this conclusion has been controversial. Here, I extend the analysis to a larger, worldwide sample and take into account potential confounding factors that may affect estimates of a slow pace of life and clutch size. In a sample of 81 species pairs consisting of closely related cooperative and non-cooperative breeders, I did not find an association between maximum longevity and cooperative breeding, controlling for diet, body mass and sampling effort. However, in a smaller sample of 37 pairs, adult annual survival was indeed higher in the cooperative breeders, controlling for body mass. There was no association between clutch size and cooperative breeding in a sample of 93 pairs. The results support the facilitating effect of high annual survival on the evolution of cooperative breeding in birds but the effect on clutch size remains elusive. PMID:24898375

  14. Accumulation of radioactive corrosion products on steel surfaces of VVER-type nuclear reactors. II. 60Co

    NASA Astrophysics Data System (ADS)

    Varga, Kálmán; Hirschberg, Gábor; Németh, Zoltán; Myburg, Gerrit; Schunk, János; Tilky, Péter

    2001-10-01

    In the case of intact fuel claddings, the predominant source of radioactivity in the primary circuits of water-cooled nuclear reactors is the activation of corrosion products in the core. The most important corrosion product radionuclides in the primary coolant of pressurized water reactors (PWRs) are 60Co, 58Co, 51Cr, 54Mn, 59Fe (as well as 110mAg in some Soviet-made VVER-type reactor). The second part of this series is focused on the complex studies of the formation and build-up of 60Co-containing species on an austenitic stainless steel type 08X18H10T (GOST 5632-61) and magnetite-covered carbon steel often to be used in Soviet-planned VVERs. The kinetics and mechanism of the cobalt accumulation were studied by a combination (coupling) of an in situ radiotracer method and voltammetry in a model solution of the primary circuit coolant. In addition, independent techniques such as X-ray photoelectron spectroscopic (XPS) and ICP-OES are also used to analyze the chemical state of Co species in the passive layer formed on stainless steel as well as the chemical composition of model solution. The experimental results have revealed that: (i) The passive behavior of the austenitic stainless steel at open-circuit conditions, the slightly alkaline pH and the reducing water chemistry can be considered to be optimal to minimize the 60Co contamination. (ii) The highly potential dependent deposition of various Co-oxides at E>1.10 V (vs. RHE) offers a unique possibility to elaborate a novel electrochemical method for the decrease or removal of cobalt traces from borate-containing coolants contaminated with 60Co and/or 58Co radionuclides.

  15. The Inversion Point of the Isothermal Reactivity Coefficient of the IPEN/MB-01 Reactor - II: Theoretical Analysis

    SciTech Connect

    Santos, A. dos; Fuga, R.; Abe, A.Y.

    2005-10-15

    TORT, an S{sub N} three-dimensional transport code, is employed for the analysis of the inversion point of the isothermal reactivity coefficient of the IPEN/MB-01 reactor. The analyses are performed in companion NJOY, AMPX-II, and TORT systems considering the data libraries ENDF/B-VI.8, JENDL3.3, and JEF3.0. The analyses reveal that for this peculiar problem, there is a need to convert all the computer codes to DOUBLE-PRECISION as well as to increase to seven the number of digits of the ANISN library generated by XSDRNPM. Contrary to the traditional diffusion theory codes, TORT k{sub eff} results are very sensitive to the number of both fine and broad groups. For instance, the traditional and very well known two- and four-group structure, largely utilized in several diffusion codes, produced simply unacceptable k{sub eff} results. The highest deviation between calculated and experimental values found for the inversion point was -4.48 deg. C. At first glance, there appears to be a significant discrepancy. However, in terms of reactivity coefficient, this discrepancy means a deviation of -0.90 {+-} 0.05 pcm/deg. C, which indicates that the calculational methodology and related nuclear data libraries meet the desired accuracy (-1.0 pcm/deg. C) for the determination of this parameter for thermal reactors.

  16. II. Electrodeposition/removal of nickel in a spouted electrochemical reactor

    PubMed Central

    Grimshaw, Pengpeng; Calo, Joseph M.; Shirvanian, Pezhman A.; Hradil, George

    2011-01-01

    An investigation is presented of nickel electrodeposition from acidic solutions in a cylindrical spouted electrochemical reactor. The effects of solution pH, temperature, and applied current on nickel removal/recovery rate, current efficiency, and corrosion rate of deposited nickel on the cathodic particles were explored under galvanostatic operation. Nitrogen sparging was used to decrease the dissolved oxygen concentration in the electrolyte in order to reduce the nickel corrosion rate, thereby increasing the nickel electrowinning rate and current efficiency. A numerical model of electrodeposition, including corrosion and mass transfer in the particulate cathode moving bed, is presented that describes the behavior of the experimental net nickel electrodeposition data quite well. PMID:22039317

  17. II. Electrodeposition/removal of nickel in a spouted electrochemical reactor.

    PubMed

    Grimshaw, Pengpeng; Calo, Joseph M; Shirvanian, Pezhman A; Hradil, George

    2011-08-17

    An investigation is presented of nickel electrodeposition from acidic solutions in a cylindrical spouted electrochemical reactor. The effects of solution pH, temperature, and applied current on nickel removal/recovery rate, current efficiency, and corrosion rate of deposited nickel on the cathodic particles were explored under galvanostatic operation. Nitrogen sparging was used to decrease the dissolved oxygen concentration in the electrolyte in order to reduce the nickel corrosion rate, thereby increasing the nickel electrowinning rate and current efficiency. A numerical model of electrodeposition, including corrosion and mass transfer in the particulate cathode moving bed, is presented that describes the behavior of the experimental net nickel electrodeposition data quite well. PMID:22039317

  18. Deep-Earth reactor: Nuclear fission, helium, and the geomagnetic field

    PubMed Central

    Hollenbach, D. F.; Herndon, J. M.

    2001-01-01

    Geomagnetic field reversals and changes in intensity are understandable from an energy standpoint as natural consequences of intermittent and/or variable nuclear fission chain reactions deep within the Earth. Moreover, deep-Earth production of helium, having 3He/4He ratios within the range observed from deep-mantle sources, is demonstrated to be a consequence of nuclear fission. Numerical simulations of a planetary-scale geo-reactor were made by using the SCALE sequence of codes. The results clearly demonstrate that such a geo-reactor (i) would function as a fast-neutron fuel breeder reactor; (ii) could, under appropriate conditions, operate over the entire period of geologic time; and (iii) would function in such a manner as to yield variable and/or intermittent output power. PMID:11562483

  19. Evaluation of Fluid Conduction and Mixing within a Subassembly of the Actinide Burner Test Reactor

    SciTech Connect

    Cliff B. Davis

    2007-09-01

    The RELAP5-3D code is being considered as a thermal-hydraulic system code to support the development of the sodium-cooled Actinide Burner Test Reactor as part of the Global Nuclear Energy Partnership. An evaluation was performed to determine whether the control system could be used to simulate the effects of non-convective mechanisms of heat transport in the fluid, including axial and radial heat conduction and subchannel mixing, that are not currently represented with internal code models. The evaluation also determined the relative importance of axial and radial heat conduction and fluid mixing on peak cladding temperature for a wide range of steady conditions and during a representative loss-of-flow transient. The evaluation was performed using a RELAP5-3D model of a subassembly in the Experimental Breeder Reactor-II, which was used as a surrogate for the Actinide Burner Test Reactor.

  20. Homogeneous fast-flux isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a liquid metal fast breeder reactor. Lithium target material is dissolved in the liquid metal coolant in order to facilitate the production and removal of tritium.

  1. Conversion Analyses for the VR-1 Reactor, part I and II.

    SciTech Connect

    Hannan, N. A.; Matos, J. E.; Stillman, J. A.; Olson, A. P.; Garner, P.L.

    2005-11-14

    At the request of the Czech Technical University (CTU) in Prague, ANL has performed independent verification calculations using the MCNP Monte Carlo code for three core configurations of the VR-1 reactor: a current core configuration B1 with HEU (36%) IRT-3M fuel assemblies and planned core configurations C1 and C2 with LEU (19.7%) IRT-4M fuel assemblies. Details of these configurations were provided to ANL by CTU. For core configuration B1, criticality calculations were performed for two sets of control rod positions provided to ANL by CTU. Fore core configurations C1 and C2, criticality calculations were done for cases with all control rods at the top positions, all control rods at the bottom positions, and two critical states of the reactor for different control rod positions. In addition, sensitivity studies for variation of the {sup 235}U mass in each fuel assembly and variation of the fuel meat and cladding thicknesses in each of the fuel tubes were doe for the C1 core configuration. The reactivity worth of the individual control rods was calculated for the B1, C1, and C2 core configurations. Finally, the reactivity feedback coefficients, the prompt neutron lifetime, and the total effective delay neutron fraction were calculated for each of the three cores.

  2. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    NASA Astrophysics Data System (ADS)

    Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.

    2013-06-01

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  3. Nodal equivalence theory for hexagonal geometry, thermal reactor analysis

    SciTech Connect

    Zika, M.; Downar, T. )

    1992-01-01

    An important aspect of advanced nodal methods is the determination of equivalent few-group parameters for the relatively large homogenized regions used in the nodal flux solution. The theoretical foundation for light water reactor (LWR) assembly homogenization methods has been clearly established, and during the last several years, its successes have secured its position in the stable of dependable LWR analysis methods. Groupwise discontinuity factors that correct for assembly homogenization errors are routinely generated along with the group constants during lattice physics analysis. During the last several years, there has been interest in applying equivalence theory to other reactor types and other geometries. A notable effort has been the work at Argonne National Laboratory to incorporate nodal equivalence theory (NET) for hexagonal lattices into the nodal diffusion option of the DIF3D code. This work was originally intended to improve the neutronics methods used for the analysis of the Experimental Breeder Reactor II (EBR-II), and Ref. 4 discusses the success of that application. More recently, however, attempts were made to apply NET to advanced, thermal reactor designs such as the modular high-temperature gas reactor (MHTGR) and the new production heavy water reactor (NPR/HWR). The same methods that were successful for EBR-II have encountered problems for these reactors. Our preliminary analysis indicates that the sharp global flux gradients in these cores requires large discontinuity factors (greater than 4 or 5) to reproduce the reference solution. This disrupts the convergence of the iterative methods used to solve for the node-wise flux moments and partial currents. Several attempts to remedy the problem have been made over the last few years, including bounding the discontinuity factors and providing improved initial guesses for the flux solution, but nothing has been satisfactory.

  4. An assessment of BWR (boiling water reactor) Mark-II containment challenges, failure modes, and potential improvements in performance

    SciTech Connect

    Kelly, D.L.; Jones, K.R.; Dallman, R.J. ); Wagner, K.C. )

    1990-07-01

    This report assesses challenges to BWR Mark II containment integrity that could potentially arise from severe accidents. Also assessed are some potential improvements that could prevent core damage or containment failure, or could mitigate the consequences of such failure by reducing the release of fission products to the environment. These challenges and improvements are analyzed via a limited quantitative risk/benefit analysis of a generic BWR/4 reactor with Mark II containment. Point estimate frequencies of the dominant core damage sequences are obtained and simple containment event trees are constructed to evaluate the response of the containment to these severe accident sequences. The resulting containment release modes are then binned into source term release categories, which provide inputs to the consequence analysis. The output of the consequences analysis is used to construct an overall base case risk profile. Potential improvements and sensitivities are evaluated by modifying the event tree spilt fractions, thus generating a revised risk profile. Several important sensitivity cases are examined to evaluate the impact of phenomenological uncertainties on the final results. 75 refs., 25 figs., 65 tabs.

  5. Preliminary nuclear safety assessment of the NEPST (Topaz II) space reactor program

    SciTech Connect

    Marshall, A.C.

    1993-01-01

    The United States (US) Strategic Defense Initiative Organization (SDIO) decided to investigate the possibility of launching a Russian Topaz II space nuclear power system. A preliminary nuclear safety assessment was conducted to determine whether or not a space mission could be conducted safely and within budget constraints. As part of this assessment, a safety policy and safety functional requirements were developed to guide both the safety assessment and future Topaz II activities. A review of the Russian flight safety program was conducted and documented. Our preliminary nuclear safety assessment included a number of deterministic analyses, such as; neutronic analysis of normal and accident configurations, an evaluation of temperature coefficients of reactivity, a reentry and disposal analysis, an analysis of postulated launch abort impact accidents, and an analysis of postulated propellant fire and explosion accidents. Based on the assessment to date, it appears that it will be possible to safely launch the Topaz II system in the US with a modification to preclude water flooded criticality. A full scale safety program is now underway.

  6. Instrumentation, Monitoring and NDE for New Fast Reactors

    SciTech Connect

    Bond, Leonard J.; Doctor, Steven R.; Bunch, Kyle J.; Good, Morris S.; Waltar, Alan E.

    2007-07-28

    The Global Nuclear Energy Partnership (GNEP) has been proposed as a viable system in which to close the fuel cycle in a manner consistent with markedly expanding the global role of nuclear power while significantly reducing proliferation risks. A key part of this system relies on the development of actinide transmutation, which can only be effectively accomplished in a fast-spectrum reactor. The fundamental physics for fast reactors is well established. However, to achieve higher standards of safety and reliability, operate with longer intervals between outages, and achieve high operating capacity factors, new instrumentation and on-line monitoring capabilities will be required--during both fabrication and operation. Since the Fast Flux Test Facility (FFTF) and Experimental Breeder ReactorII (EBR-II) reactors were operational in the USA, there have been major advances in instrumentation, not the least being the move to digital systems. Some specific capabilities have been developed outside the USA, but new or at least re-established capabilities will be required. In many cases the only available information is in reports and papers. New and improved sensors and instrumentation will be required. Advanced instrumentation has been developed for high-temperature/high-flux conditions in some cases, but most of the original researchers and manufacturers are retired or no longer in business.

  7. Fusion breeder studies program: Final report

    SciTech Connect

    Berwald, D.H.

    1986-10-17

    This report is an assessment of technology related to hybrid reactors, especially the Fission-suppressed hybrid. A description of a typical fission-suppressed reactor is given. The economic advantages of the use of a hybrid reactor as part of a fuel cycle center are discussed at length. The inherent safety advantages of the hybrid reactor are analyzed. The report concludes with a proposed timetable for research and development. (JDH)

  8. US-DOE Fusion-Breeder Program: blanket design and system performance

    SciTech Connect

    Lee, J.D.

    1983-01-01

    Conceptual design studies are being used to assess the technical and economic feasibility of fusion's potential to produce fissile fuel. A reference design of a fission-suppressed blanket using conventional materials is under development. Theoretically, a fusion breeder that incorporates this fusion-suppressed blanket surrounding a 3000-MW tandem mirror fusion core produces its own tritium plus 5600 kg of /sup 233/U per year. The /sup 233/U could then provide fissile makeup for 21 GWe of light-water reactor (LWR) power using a denatured thorium fuel cycle with full recycle. This is 16 times the net electric power produced by the fusion breeder (1.3 GWe). The cost of electricity from this fusion-fission system is estimated to be only 23% higher than the cost from LWRs that have makeup from U/sub 3/O/sub 8/ at present costs (55 $/kg). Nuclear performance, magnetohydrodynamics (MHD), radiation effects, and other issues concerning the fission-suppressed blanket are summarized, as are some of the present and future objectives of the fusion breeder program.

  9. Radiation Damage Calculations for the FUBR and BEATRIX Irradiations of Lithium Compounds in EBR-II and FFTF

    SciTech Connect

    LR Greenwood

    1999-06-17

    The Fusion Breeder Reactor (FUBR) and Breeder Exchange Matrix (BEATRIX) experiments were cooperative efforts by members of the International Energy Agency to investigate the irradiation behavior of solid breeder materials for tritium production to support future fusion reactors. Lithium ceramic materials including Li{sub 2}O, LiAlO{sub 2}, Li{sub 4}SiO{sub 4}, and Li{sub 2}ZrO{sub 3} with varying {sup 6}Li enrichments from 0 to 95% were irradiated in a series of experiments in the Experimental Breeder Reactor (EBR II) and in the Fast Flux Test Facility (FFTF) over a period of about 10 years from 1982 to 1992. These experiments were characterized in terms of the nominal fast neutron fluences and measured {sup 6}Li burnup factors, as determined by either mass spectrometry or helium measurements. Radiation damage in these compounds is caused by both the {sup 6}Li-burnup reaction and by all other possible neutron reactions with the atoms in the compound materials. In this report, displacements per atom (dpa) values have been calculated for each type of material in each of the various irradiations that were conducted. Values up to 11% {sup 6}Li-burnup and 130 dpa are predicted for the longest irradiations. The dpa cross sections were calculated for each compound using the SPECOMP computer code. Details of the dpa calculations are presented in the report. Total dpa factors were determined with the SPECTER computer code by averaging the dpa cross sections over the measured or calculated neutron flux spectra for each series of irradiations. Using these new calculations, previously measured radiation damage effects in these lithium compounds can be compared or correlated with other irradiation data on the basis of the dpa factor as well as {sup 6}Li-burnup.

  10. Feasibility study of a fission-suppressed tokamak fusion breeder

    SciTech Connect

    Moir, R.W.; Lee, J.D.; Neef, W.S.; Berwald, D.H.; Garner, J.K.; Whitley, R.H.; Ghoniem, N.; Wong, C.P.C.; Maya, I.; Schultz, K.R.

    1984-12-01

    The preliminary conceptual design of a tokamak fissile fuel producer is described. The blanket technology is based on the fission suppressed breeding concept where neutron multiplication occurs in a bed of 2 cm diameter beryllium pebbles which are cooled by helium at 50 atmospheres pressure. Uranium-233 is bred in thorium metal fuel elements which are in the form of snap rings attached to each beryllium pebble. Tritium is bred in lithium bearing material contained in tubes immersed in the pebble bed and is recovered by a purge flow of helium. The neutron wall load is 3 MW/m/sup 2/ and the blanket material is ferritic steel. The net fissile breeding ratio is 0.54 +- 30% per fusion reaction. This results in the production of 4900 kg of /sup 233/U per year from 3000 MW of fusion power. This quantity of fuel will provide makeup fuel for about 12 LWRs of equal thermal power or about 18 1 GW/sub e/ LWRs. The calculated cost of the produced uranium-233 is between $23/g and $53/g or equivalent to $10/kg to $90/kg of U/sub 3/O/sub 8/ depending on government financing or utility financing assumptions. Additional topics discussed in the report include the tokamak operating mode (both steady state and long pulse considered), the design and breeding implications of using a poloidal divertor for impurity control, reactor safety, the choice of a tritium breeder, and fuel management.

  11. Disposition of weapon-grade plutonium with pebble bed type HTGRs using Pu burner balls and Th breeder balls

    SciTech Connect

    Yamashita, Kiyonobu; Tokuhara, Kazumi; Fujimoto, Nozomu; Kunitomi, Kazuhiko

    1996-08-01

    A concept of reactor system was developed with which weapons-grade plutonium could be made perfectly worthless in use for weapons. It is a pebble bed type HTGR using Pu burner ball fuels and Th breeder ball fuels. The residual amounts of {sup 239}Pu in spent Pu balls become less than 1% of the initial loading. Furthermore, a method was found that the power coefficient could be made negative by heavy Pu loading in the Pu burner ball fuels.

  12. Affinity chromatography purification of angiotensin II reactor using photoactivable biotinylated probes

    SciTech Connect

    Marie, J.; Seyer, R.; Lombard, C.; Desarnaud, F.; Aumelas, A.; Jard, A.; Bonnafous, J.C. )

    1990-09-25

    The authors have developed biotinylated photoactivable probes that are suitable for covalent labeling of angiotensin II (AII) receptors and the subsequent purification of covalent complexes through immobilized avidin or streptavidin. One of these probes, biotin-NH(CH{sub 2}){sub 2}SS(CH{sub 2}){sub 2}CO-(Ala{sup 1}, Phe(4N{sub 3}){sup 8})AII, which contains a cleavage disulfide bridge in its spacer arm and which displays, in its radioiodinated form, very high affinity for AII receptors (K{sub d}{approximately}1 nM), proved to be suitable for indirect affinity chromatography of rate liver receptor with facilitated recovery from avidin gels by use of reducing agents. This constituted the central step of an efficient partial purification scheme involving hydroxylapatite chromatography, streptavidin chromatography, and thiopropyl-Sepharose chromatography. SDS-PAGE analysis and autoradiography established the identity of the purified entity (molecular weight 65K) as the AII receptor. Possible ways of completing purification to homogeneity and extrapolation of the protocols to a preparative scale are discussed, as well as the potential contribution of our new probes to the study of the structural properties of angiotensin receptors.

  13. Performance and microbial communities of Mn(II)-based autotrophic denitrification in a Moving Bed Biofilm Reactor (MBBR).

    PubMed

    Su, Jun Feng; Luo, Xian Xin; Wei, Li; Ma, Fang; Zheng, Sheng Chen; Shao, Si Cheng

    2016-07-01

    In this study, Mn(II) as electron donor was tested for the effects on denitrification in the MBBR under the conditions of initial nitrate concentration (10mgL(-1), 30mgL(-1), 50mgL(-1)), pH (5, 6, 7) and hydraulic retention time (HRT) (4h, 8h, 12h) which conducted by response surface methodology (RSM), the results demonstrated that the highest nitrate removal efficiency was occurred under the conditions of initial nitrate concentration of 47.64mgL(-1), HRT of 11.96h and pH 5.21. Analysis of SEM and flow cytometry suggested that microorganisms were immobilized on the Yu Long plastic carrier media successfully before the reactor began to operate. Furthermore, high-throughput sequencing was employed to characterize and compare the community compositions and structures of MBBR under the optimum conditions, the results showed that Pseudomonas sp. SZF15 was the dominant contributor for effective removal of nitrate in the MBBR. PMID:27061262

  14. A U. S. Perspective on Fast Reactor Fuel Fabrication Technology and Experience Part I: Metal Fuels and Assembly Design

    SciTech Connect

    Douglas E. Burkes; Randall S. Fielding; Douglas L. Porter; Douglas C. Crawford; Mitchell K. Meyer

    2009-06-01

    This paper is Part I of a review focusing on the United States experience with metallic fast reactor fuel fabrication and assembly design for the Experimental Breeder Reactor-II and the Fast Flux Test Facility, and it also refers to the impact of development in other nations. Experience with metal fuel fabrication in the United States is extensive, including over 60 years of research conducted by the government, national laboratories, industry, and academia. This experience has culminated into a foundation of research and resulted in significant improvements to the technologies employed to fabricate metallic fast reactor fuel. This part of the review documents the current state of fuel fabrication technologies for metallic fuels, 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.

  15. Chromium-molybdenum steels for fusion-reactor applications

    SciTech Connect

    Klueh, R.L.

    1981-08-01

    Because ferritic steels have been found to have excellent resistance to swelling when irradiated in a fast-breeder reactor, Cr-Mo steels have recently become of interest for nuclear applications, both as cladding and duct material for fast-breeder reactors and as a first-wall and blanket structural material for fusion reactors. In this paper we will assess the Cr-Mo steels for fusion reactor applications. Possible approaches on how Cr-Mo steels may be further developed for this application will be proposed.

  16. Development of advanced tritium breeders and neutron multipliers for DEMO solid breeder blankets

    NASA Astrophysics Data System (ADS)

    Tsuchiya, K.; Hoshino, T.; Kawamura, H.; Mishima, Y.; Yoshida, N.; Terai, T.; Tanaka, S.; Munakata, K.; Kato, S.; Uchida, M.; Nakamichi, M.; Yamada, H.; Yamaki, D.; Hayashi, K.

    2007-09-01

    In efforts to develop advanced tritium breeders, the effects of additives to lithium titanate (Li2TiO3) have been investigated, and good prospects have been obtained by using oxide additives such as TiO2, CaO and Li2O. As for the neutron multiplier, the development of a real-size electrode fabrication technique and the characterization of beryllium-based intermetallic compounds such as Be-Ti and Be-V have been performed. Properties of Be-Ti alloys have been found to be better than those of beryllium metal. In particular, steam interaction of a Be-Ti alloy was about 1/1000 as small as that of beryllium metal. These activities have led to bright prospects for the realization of the water-cooled DEMO breeder blanket by application of these advanced materials.

  17. System modelling to support accelerated fuel transfer rate at EBR-II

    SciTech Connect

    Imel, G.R.; Houshyar, A.; Planchon, H.P.; Cutforth, D.C.

    1995-06-01

    The Experimental Breeder Reactor-II (EBR-II) ia a 62.5 MW(th) liquid metal reactor operated by Argonne National Laboratory for The United States Department of Energy. The reactor is located near Idaho Falls, Idaho at the Argonne-West site (ANL-W). Full power operation was achieved in 1964,- the reactor operated continuously since that time until October 1994 in a variety of configurations depending on the programmatic mission. A three year program was initiated in October, 1993 to replace the 330 depleted uranium blanket subassemblies (S/As) with stainless steel reflectors. It was intended to operate the reactor during the three year blanket unloading program, followed by about a half year of driver fuel unloading. However, in the summer of 1994, Congress dictacted that EBR-II be shut down October 1, and complete defueling without operation. To assist in the planning for resources needed for this defueling campaign, a mathematical model of the fuel handling sequence was developed utilizing the appropriate reliability factors and inherent mm constraints of each stage of the process. The model allows predictions of transfer rates under different scenarios. Additionally, it has facilitated planning of maintenance activities, as well as optimization of resources regarding manpower and modification effort. The model and its application is described in this paper.

  18. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    SciTech Connect

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high

  19. The CANDU Reactor System: An Appropriate Technology.

    PubMed

    Robertson, J A

    1978-02-10

    CANDU power reactors are characterized by the combination of heavy water as moderator and pressure tubes to contain the fuel and coolant. Their excellent neutron economy provides the simplicity and low costs of once-through natural-uranium fueling. Future benefits include the prospect of a near-breeder thorium fuel cycle to provide security of fuel supply without the need to develop a new reactor such as the fast breeder. These and other features make the CANDU system an appropriate technology for countries, like Canada, of intermediate economic and industrial capacity. PMID:17788102

  20. Characterization of the effects of continuous salt processing on the performance of molten salt fusion breeder blankets

    SciTech Connect

    Patterson-Hine, F.A.

    1984-05-01

    Several continuous salt processing options are available for use in molten salt fusion breeder blanket designs. The effects of processing on blanket performance have been assessed for three levels of processing and various equilibrium uranium concentrations in the salt. A one-dimensional model of the blanket was used in the neutronics analysis which incorporated transport calculations with time-dependent isotope generation and depletion calculations. The level of salt processing was found to have little effect on the behavior of the blanket during reactor operation; however, significant effects were observed during the decay period after reactor shutdown.

  1. Performance of low smeared density sodium-cooled fast reactor metal fuel

    SciTech Connect

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-06-17

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  2. Performance of low smeared density sodium-cooled fast reactor metal fuel

    NASA Astrophysics Data System (ADS)

    Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  3. Performance of Low Smeared Density Sodium-cooled Fast Reactor Metal Fuel

    SciTech Connect

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  4. A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

    SciTech Connect

    Jacques V Hugo; David I Gertman; Jeffrey C Joe

    2014-08-01

    This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

  5. Thermal-hydraulic-structural behavior of the EBR-II IHX for overpower transients

    SciTech Connect

    Mohr, D.; Chang, L.K.; Lee, M.J.; Feldman, E.E.

    1982-01-01

    A detailed study has been made of the effects of the Operational Reliability Testing (ORT) program on major plant components of the Experimental Breeder Reactor No. II (EBR-II). This paper describes the integrated thermal-hydraulic-structural analyses conducted for the intermediate heat exchanger (IHX) with the aid of the NATDEMO, THTB, and ANSYS codes. An extensive analysis revealed the stress limiting area to be the junction between the upper head and upper tube sheet. The analyses indicate, however, that the EBR-II IHX, the major plant component most affected by the ORT program, will be able to withstand the thermal stress and accumulated fatigue damage during the lifetime of the plant including the ORT program.

  6. A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

    NASA Astrophysics Data System (ADS)

    Kamal, I.; Yunus, S. M.; Datta, T. K.; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S.; Berliner, R.; Yelon, W. B.

    2016-07-01

    A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6mm thickness each. The monochromator design was optimized to provide maximum flux on 3mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30˚ (2θ) at each step and covers 120˚ in 4 steps. When the detector is placed at 1.6 m it subtends 20˚ at each step and covers 120˚ in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology

  7. Enhanced reduction of Fe(II)EDTA-NO/Fe(III)EDTA in NO(x) scrubber solution using a three-dimensional biofilm-electrode reactor.

    PubMed

    Zhou, Ya; Gao, Lin; Xia, Yin-Feng; Li, Wei

    2012-11-20

    A promising technique called chemical absorption-biological reduction (CABR) integrated approach has been developed recently for the nitrogen oxides (NO(x)) removal from flue gases. The major challenge for this approach is how to enhance the rate of the biological reduction step. To tackle the challenge, a three-dimensional biofilm-electrode reactor (3D-BER) was utilized. This reactor provides not only considerable amount of sites for biofilm, but also many electron donors for bioreduction. Factors affecting the performance of 3D-BER were optimized, including material of the third electrode (graphite), glucose concentration (1000 mg·L(-1)), and volume current density (30.53 A·m(-3) NCC). Experimental results clearly demonstrated that this method significantly promotes the bioreduction rate of Fe(II)EDTA-NO (0.313 mmol·L(-1)·h(-1)) and Fe(III)EDTA (0.564 mmol·L(-1)·h(-1)) simultaneously. Experiments on the mechanism showed that Fe(II)EDTA serves as the primary electron donor in the reduction of Fe(II)EDTA-NO, whereas the reduction of Fe(III)EDTA took advantage of both glucose and electrolysis-generated H(2) as electron donors. High concentration of Fe(II)EDTA-NO or Fe(III)EDTA interferes the bioreduction of the other one. The proposed methodology shows a promising prospect for NO(x) removal from flue gas. PMID:23113866

  8. Beam injection improvement for electron cyclotron resonance charge breeders

    SciTech Connect

    Lamy, T.; Angot, J.; Sortais, P.; Thuillier, T.

    2012-02-15

    The injection of a 1+ beam into an electron cyclotron resonance (ECR) charge breeder is classically performed through a grounded tube placed on its axis at the injection side. This tube presents various disadvantages for the operation of an ECR charge breeder. First experiments without a grounded tube show a better use of the microwave power and a better charge breeding efficiency. The optical acceptance of the charge breeder without decelerating tube allows the injection of high intensity 1+ ion beams at high energy, allowing metals sputtering inside the ion source. The use of this method for refractory metallic ion beams production is evaluated.

  9. Thermal-Hydraulic Analysis of the 3-MW TRIGA MARK-II Research Reactor Under Steady-State and Transient Conditions

    SciTech Connect

    Huda, M.Q.; Bhuiyan, S.I.; Chakrobortty, T.K.; Sarker, M.M.; Mondal, M.A.W

    2001-07-15

    Important thermal-hydraulic parameters of the 3-MW TRIGA MARK-II research reactor operating under both steady-state and transient conditions are reported. Neutronic analyses were performed by using the CITATION diffusion code and the MCNP4B2 Monte Carlo code. The output of CITATION and MCNP4B2 were input to the PARET thermal-hydraulic code to study the steady-state and transient thermal-hydraulic behavior of the reactor. To benchmark the PARET model, data were obtained from different measurements performed by thermocouples in the instrumented fuel (IF) rod during the steady-state operation both under forced- and natural-convection mode and compared with the calculation. The mass flow rates needed for input to PARET were taken from the Final Safety Analysis Report for a downward forced coolant flow equivalent to 3500 gal/min. For natural convection cooling of the reactor, the mass flow rate was generated using the NCTRIGA code. Peak fuel temperatures measured by the thermocouples in the IF rods at different power levels of the TRIGA core were compared with the values calculated by PARET. The axial distribution of the temperatures of the fuel centerline, fuel surface, and the cladding surface in the hot channel were calculated for the reactor operating at the full-power level. Fuel surface heat flux and heat transfer coefficients for the hot channel were also calculated for the reactor operating at the full-power level. The investigated results were found to be in good agreement with the experimental and operational values. The testing of the PARET model calculations through benchmarking the available TRIGA experimental and operational data for pulse-mode operations showed that PARET can successfully be used to analyze the transient behavior of the reactor. Major transient parameters, such as peak power and prompt energy released after pulse, full-width at half-maximum of pulse peak, and maximum fuel centerline temperatures for different fuel elements at different

  10. LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor: Appendix A-4

    SciTech Connect

    Sviatoslavsky, I.N.; Attaya, H.M.; Corradini, M.L.; Lomperski, S.

    1987-01-01

    This paper describes the preliminary analysis of LOFA (loss of flow accident) and LOCA (loss of coolant accident) in the TIBER-II engineering test reactor breeding shield. TIBER-II is a compact reactor with a major radius of 3 m and thus requires a thin, high efficiency shield on the inboard side. The use of tungsten in the inboard shield implies a rather high rate of afterheat upon plasma shutdown, which must be dissipated in a controlled manner to avoid the possibility of radioactivity release or threatening the investment. Because the shield is cooled with an aqueous solution, LOFA does not pose a problem as long as natural convection can be established. LOCA, however, has more serious consequences, particularly on the inboard side. Circulation of air by natural convection is proposed as a means for dissipating the inboard shield decay heat. The safety and environmental implications of such a scheme are evaluated. It is shown that the inboard shield temperature never exceeds 510/sup 0/C following LOCA posing no hazard to reactor personnel and not threatening the investment. 7 refs., 6 figs.

  11. Estimation of (41)Ar activity concentration and release rate from the TRIGA Mark-II research reactor.

    PubMed

    Hoq, M Ajijul; Soner, M A Malek; Rahman, A; Salam, M A; Islam, S M A

    2016-03-01

    The BAEC TRIGA research reactor (BTRR) is the only nuclear reactor in Bangladesh. Bangladesh Atomic Energy Regulatory Authority (BAERA) regulations require that nuclear reactor licensees undertake all reasonable precautions to protect the environment and the health and safety of persons, including identifying, controlling and monitoring the release of nuclear substances to the environment. The primary activation product of interest in terms of airborne release from the reactor is (41)Ar. (41)Ar is a noble gas readily released from the reactor stacks and most has not decayed by the time it moves offsite with normal wind speed. Initially (41)Ar is produced from irradiation of dissolved air in the primary water which eventually transfers into the air in the reactor bay. In this study, the airborne radioisotope (41)Ar generation concentration, ground level concentration and release rate from the BTRR bay region are evaluated theoretically during the normal reactor operation condition by several governing equations. This theoretical calculation eventually minimizes the doubt about radiological safety to determine the radiation level for (41)Ar activity whether it is below the permissible limit or not. Results show that the estimated activity for (41)Ar is well below the maximum permissible concentration limit set by the regulatory body, which is an assurance for the reactor operating personnel and general public. Thus the analysis performed within this paper is so much effective in the sense of ensuring radiological safety for working personnel and the environment. PMID:26736180

  12. Tritium percolation, convection, and permeation in fusion solid breeder blankets

    SciTech Connect

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

    1985-01-01

    Models are developed to describe the percolation of released tritium through the breeder interconnected porosity to the purge stream, convection of tritium by the helium purge stream, and leakage or permeation of tritium through the structural material to the primary coolant system. Important parameters in the models are tritium generation rate, breeder microstructure, tritium species in the gas phase, temperatures, tritium diffusivities and permeabilities, and effectiveness of oxide barriers.

  13. Control rod drive for reactor shutdown

    DOEpatents

    McKeehan, Ernest R.; Shawver, Bruce M.; Schiro, Donald J.; Taft, William E.

    1976-01-20

    A means for rapidly shutting down or scramming a nuclear reactor, such as a liquid metal-cooled fast breeder reactor, and serves as a backup to the primary shutdown system. The control rod drive consists basically of an in-core assembly, a drive shaft and seal assembly, and a control drive mechanism. The control rod is driven into the core region of the reactor by gravity and hydraulic pressure forces supplied by the reactor coolant, thus assuring that common mode failures will not interfere with or prohibit scramming the reactor when necessary.

  14. Fuel pins with both target and fuel pellets in an isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

  15. Assemblies with both target and fuel pins in an isotope-production reactor

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

  16. Vented target elements for use in an isotope-production reactor. [LMFBR

    DOEpatents

    Cawley, W.E.; Omberg, R.P.

    1982-08-19

    A method is described for producing tritium gas in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins equipped with vents, and tritium gas is recovered from the coolant.

  17. Lithium mass transport in ceramic breeder materials

    SciTech Connect

    Blackburn, P.E.; Johnson, C.E.

    1990-01-01

    The objective of this activity is to measure the lithium vaporization from lithium oxide breeder material under differing temperature and moisture partial pressure conditions. Lithium ceramics are being investigated for use as tritium breeding materials. The lithium is readily converted to tritium after reacting with a neutron. With the addition of 1000 ppM H{sub 2} to the He purge gas, the bred tritium is readily recovered from the blanket as HT and HTO above 400{degree}C. Within the solid, tritium may also be found as LiOT which may transport lithium to cooler parts of the blanket. The pressure of LiOT(g), HTO(g), or T{sub 2}O(g) above Li{sub 2}O(s) is the same as that for reactions involving hydrogen. In our experiments we were limited to the use of hydrogen. The purpose of this work is to investigate the transport of LiOH(g) from the blanket material. 8 refs., 1 fig., 3 tabs.

  18. Guideline for Performing Systematic Approach to Evaluate and Qualify Legacy Documents that Support Advanced Reactor Technology Activity

    SciTech Connect

    Honma, George

    2015-10-01

    The establishment of a systematic process for the evaluation of historic technology information for use in advanced reactor licensing is described. Efforts are underway to recover and preserve Experimental Breeder Reactor II and Fast Flux Test Facility historical data. These efforts have generally emphasized preserving information from data-acquisition systems and hard-copy reports and entering it into modern electronic formats suitable for data retrieval and examination. The guidance contained in this document has been developed to facilitate consistent and systematic evaluation processes relating to quality attributes of historic technical information (with focus on sodium-cooled fast reactor (SFR) technology) that will be used to eventually support licensing of advanced reactor designs. The historical information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The evaluation process is prescribed in terms of SFR technology, but the process can be used to evaluate historical information for any type of advanced reactor technology. An appendix provides a discussion of typical issues that should be considered when evaluating and qualifying historical information for advanced reactor technology fuel and source terms, based on current light water reactor (LWR) requirements and recent experience gained from Next Generation Nuclear Plant (NGNP).

  19. Part I. Fuel-motion diagnostics in support of fast-reactor safety experiments. Part II. Fission product detection system in support of fast reactor safety experiments

    SciTech Connect

    Devolpi, A.; Doerner, R.C.; Fink, C.L.; Regis, J.P.; Rhodes, E.A.; Stanford, G.S.; Braid, T.H.; Boyar, R.E.

    1986-05-01

    In all destructive fast-reactor safety experiments at TREAT, fuel motion and cladding failure have been monitored by the fast-neutron/gamma-ray hodoscope, providing experimental results that are directly applicable to design, modeling, and validation in fast-reactor safety. Hodoscope contributions to the safety program can be considered to fall into several groupings: pre-failure fuel motion, cladding failure, post-failure fuel motion, steel blockages, pretest and posttest radiography, axial-power-profile variations, and power-coupling monitoring. High-quality results in fuel motion have been achieved, and motion sequences have been reconstructed in qualitative and quantitative visual forms. A collimated detection system has been used to observe fission products in the upper regions of a test loop in the TREAT reactor. Particular regions of the loop are targeted through any of five channels in a rotatable assembly in a horizontal hole through the biological shield. A well-type neutron detector, optimized for delayed neutrons, and two GeLi gamma ray spectrometers have been used in several experiments. Data are presented showing a time history of the transport of Dn emitters, of gamma spectra identifying volatile fission products deposited as aerosols, and of fission gas isotopes released from the coolant.

  20. Development of electron beam ion source charge breeder for rare isotopes at Californium Rare Isotope Breeder Upgrade

    SciTech Connect

    Kondrashev, S.; Dickerson, C.; Levand, A.; Ostroumov, P. N.; Pardo, R. C.; Savard, G.; Vondrasek, R.; Alessi, J.; Beebe, E.; Pikin, A.; Kuznetsov, G. I.; Batazova, M. A.

    2012-02-15

    Recently, the Californium Rare Isotope Breeder Upgrade (CARIBU) to the Argonne Tandem Linac Accelerator System (ATLAS) was commissioned and became available for production of rare isotopes. Currently, an electron cyclotron resonance ion source is used as a charge breeder for CARIBU beams. To further increase the intensity and improve the purity of neutron-rich ion beams accelerated by ATLAS, we are developing a high-efficiency charge breeder for CARIBU based on an electron beam ion source (EBIS). The CARIBU EBIS charge breeder will utilize the state-of-the-art EBIS technology recently developed at Brookhaven National Laboratory (BNL). The electron beam current density in the CARIBU EBIS trap will be significantly higher than that in existing operational charge-state breeders based on the EBIS concept. The design of the CARIBU EBIS charge breeder is nearly complete. Long-lead components of the EBIS such as a 6-T superconducting solenoid and an electron gun have been ordered with the delivery schedule in the fall of 2011. Measurements of expected breeding efficiency using the BNL Test EBIS have been performed using a Cs{sup +} surface ionization ion source for external injection in pulsed mode. In these experiments we have achieved {approx}70% injection/extraction efficiency and breeding efficiency into the most abundant charge state of {approx}17%.

  1. Development of electron beam ion source charge breeder for rare isotopes at Californium Rare Isotope Breeder Upgrade

    SciTech Connect

    Kondrashev S.; Alessi J.; Dickerson, C.; Levand, A.; Ostroumov, P.N.; Pardo, R.C.; Savard, G.; Vondrasek, R.; Beebe, E.; Pikin, A.; Kuznetsov, G.I.; Batazova, M.A.

    2012-02-03

    Recently, the Californium Rare Isotope Breeder Upgrade (CARIBU) to the Argonne Tandem Linac Accelerator System (ATLAS) was commissioned and became available for production of rare isotopes. Currently, an electron cyclotron resonance ion source is used as a charge breeder for CARIBU beams. To further increase the intensity and improve the purity of neutron-rich ion beams accelerated by ATLAS, we are developing a high-efficiency charge breeder for CARIBU based on an electron beam ion source (EBIS). The CARIBU EBIS charge breeder will utilize the state-of-the-art EBIS technology recently developed at Brookhaven National Laboratory (BNL). The electron beam current density in the CARIBU EBIS trap will be significantly higher than that in existing operational charge-state breeders based on the EBIS concept. The design of the CARIBU EBIS charge breeder is nearly complete. Long-lead components of the EBIS such as a 6-T superconducting solenoid and an electron gun have been ordered with the delivery schedule in the fall of 2011. Measurements of expected breeding efficiency using the BNL Test EBIS have been performed using a Cs{sup +} surface ionization ion source for external injection in pulsed mode. In these experiments we have achieved {approx}70% injection/extraction efficiency and breeding efficiency into the most abundant charge state of {approx}17%.

  2. Multiple breeders, breeder shifts and inclusive fitness returns in an ant

    PubMed Central

    Bargum, Katja; Sundström, Liselotte

    2007-01-01

    In social insects, colonies may contain multiple reproductively active queens. This leads to potential conflicts over the apportionment of brood maternity, especially with respect to the production of reproductive offspring. We investigated reproductive partitioning in offspring females (gynes) and workers in the ant Formica fusca, and combined this information with data on the genetic returns gained by workers. Our results provide the first evidence that differential reproductive partitioning among breeders can enhance the inclusive fitness returns for sterile individuals that tend non-descendant offspring. Two aspects of reproductive partitioning contribute to this outcome. First, significantly fewer mother queens contribute to gyne (new reproductive females) than to worker brood, such that relatedness increases from worker to gyne brood. Second, and more importantly, adult workers were significantly more related to the reproductive brood raised by the colony, than to the contemporary worker brood. Thus, the observed breeder shift leads to genetic benefits for the adult workers that tend the brood. Our results also have repercussions for genetic population analyses. Given the observed pattern of reproductive partitioning, estimates of effective population size based on worker and gyne samples are not interchangeable. PMID:17439857

  3. Performance of low smeared density sodium-cooled fast reactor metal fuel

    DOE PAGESBeta

    Porter, D. L.; H. J. M. Chichester; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-06-17

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at. % burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactormore » designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low metaling points and gaseous precursors (Cs and Rb). Lastly, a model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.« less

  4. Tritium permeation and recovery for the helium-cooled molten salt fusion breeder

    SciTech Connect

    Sherwood, A.E.

    1984-09-01

    Design concepts are presented to control tritium permeation from a molten salt/helium fusion breeder reactor. This study assumes tritium to be a gas dissolved in molten salt, with TF formation suppressed. Tritium permeates readily through the hot steel tubes of the reactor and steam generator and will leak into the steam system at the rate of about one gram per day in the absence of special permeation barriers, assuming that 1% of the helium coolant flow rate is processed for tritium recovery at 90% efficiency per pass. The proposed permeation barrier for the reactor tubes is a 10 ..mu..m layer of tungsten which, in principle, will reduce tritium blanket permeation by a factor of about 300 below the bare-steel rate. A research and development effort is needed to prove feasibility or to develop alternative barriers. A 1 mm aluminum sleeve is proposed to suppress permeation through the steam generator tubes. This gives a calculated reduction factor of more than 500 relative to bare steel, including a factor of 30 due to an assumed oxide layer. The permeation equations are developed in detail for a multi-layer tube wall including a frozen salt layer and with two fluid boundary-layer resistances. Conditions are discussed for which Sievert's or Henry's Law materials become flux limiters. An analytical model is developed to establish the tritium split between wall permeation and reactor-tube flow.

  5. Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

    NASA Astrophysics Data System (ADS)

    Gao, Fangfang; Zhang, Xiaokang; Pu, Yong; Zhu, Qingjun; Liu, Songlin

    2016-08-01

    Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor (CFETR) operating on a Deuterium-Tritium (D-T) fuel cycle. It is necessary to study the tritium breeding ratio (TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder (WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket, the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code (MCNP) and the fusion activation code FISPACT-2007. The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation. In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015GB108002, and 2014GB119000), and by National Natural Science Foundation of China (No. 11175207)

  6. Planning and supervision of reactor defueling using discrete event techniques

    SciTech Connect

    Garcia, H.E.; Imel, G.R.; Houshyar, A.

    1995-12-31

    New fuel handling and conditioning activities for the defueling of the Experimental Breeder Reactor II are being performed at Argonne National Laboratory. Research is being conducted to investigate the use of discrete event simulation, analysis, and optimization techniques to plan, supervise, and perform these activities in such a way that productivity can be improved. The central idea is to characterize this defueling operation as a collection of interconnected serving cells, and then apply operational research techniques to identify appropriate planning schedules for given scenarios. In addition, a supervisory system is being developed to provide personnel with on-line information on the progress of fueling tasks and to suggest courses of action to accommodate changing operational conditions. This paper provides an introduction to the research in progress at ANL. In particular, it briefly describes the fuel handling configuration for reactor defueling at ANL, presenting the flow of material from the reactor grid to the interim storage location, and the expected contributions of this work. As an example of the studies being conducted for planning and supervision of fuel handling activities at ANL, an application of discrete event simulation techniques to evaluate different fuel cask transfer strategies is given at the end of the paper.

  7. Hot Fuel Examination Facility's neutron radiography reactor

    SciTech Connect

    Pruett, D.P.; Richards, W.J.; Heidel, C.C.

    1983-01-01

    Argonne National Laboratory-West is located near Idaho Falls, Idaho, and is operated by the University of Chicago for the United States Department of Energy in support of the Liquid Metal Fast Breeder Reactor Program, LMFBR. The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both nondestructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the nondestructive examination techniques utilized at HFEF is neutron radiography, which is provided by the NRAD reactor facility (a TRIGA type reactor) below the HFEF hot cell.

  8. Seismic design technology for Breeder Reactor structures. Volume 3: special topics in reactor structures

    SciTech Connect

    Reddy, D.P.

    1983-04-01

    This volume is divided into six chapters: analysis techniques, equivalent damping values, probabilistic design factors, design verifications, equivalent response cycles for fatigue analysis, and seismic isolation. (JDB)

  9. An assessment of liquid-metal centrifugal pumps at three fast reactors

    SciTech Connect

    Smith, M.S.; Wood, D.H.; Drischler, J.D. )

    1993-10-01

    The results of an analysis using data reports submitted to the Centralized Reliability Data Organization (CREDO) to predict the onset of the wearout life period for large sodium centrifugal pumps is described. For CREDO data collection and analysis purposes, a mechanical pump'' includes the pumping unit, its driver, and the coupling between the two. Statistical data were compiled from event reports received from three fast reactors: the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF) in the US and the JOYO Experimental Fast Reactor operated by the Power Reactor and Nuclear Fuel Development Corporation of Japan. Cumulative event rates were calculated for the investigated pumps at each facility and for the entire population. For all pumps, the event rate was computed as 34.4 event/million operating hours with 5 and 95% one-sided confidence limits of 26.3 and 44.4 event/million operating hours, respectively. The cumulative event rates for EBR-II, FFTF, and JOYO were computed as 30.0, 32.4, and 40.6 event/million pump operating hours, respectively. Results from EBR-II indicate that there is a definite time-dependent relationship between event rates and pump age; the common event mode at EBR-II is pump binding or seizing due to the buildup of sodium deposits in the vicinity of the lower labyrinth seal. There is no indication from FFTF that the six centrifugal pumps have reached the end of their useful life; these pumps have been event free for their last 40,000 operating hours. Following a 50,000-h even-free operating period at JOYO, bearings in the secondary pumps required additional unscheduled maintenance. However, there is no indication that these pumps have entered into the wearout life period; more data are required to draw any such conclusion.

  10. Validation of absolute axial neutron flux distribution calculations with MCNP with 197Au(n,γ)198Au reaction rate distribution measurements at the JSI TRIGA Mark II reactor.

    PubMed

    Radulović, Vladimir; Štancar, Žiga; Snoj, Luka; Trkov, Andrej

    2014-02-01

    The calculation of axial neutron flux distributions with the MCNP code at the JSI TRIGA Mark II reactor has been validated with experimental measurements of the (197)Au(n,γ)(198)Au reaction rate. The calculated absolute reaction rate values, scaled according to the reactor power and corrected for the flux redistribution effect, are in good agreement with the experimental results. The effect of different cross-section libraries on the calculations has been investigated and shown to be minor. PMID:24316530

  11. Recent advances in the development of solid breeder-blanket materials

    SciTech Connect

    Johnson, C.E.; Hollenburg, G.W.

    1983-01-01

    Increasing attention in breeder-blanket development has been given to the lithium-containing ceramic materials. The most promising of these materials include Li/sub 2/O, Li/sub 8/ZrO/sub 6/, Li/sub 4/SiO/sub 4/, and ..gamma..-LiAlO/sub 2/. Recent studies have focused on Li/sub 2/O because of its high tritium breeding potential and good thermal characteristics. Tritium solubility in Li/sub 2/O is within acceptable ranges and this oxide displays excellent behavior under neutron irradiation. A broad scope of laboratory and in-reactor irradiation experiments are underway to further investigate these materials.

  12. Development of welding technologies for the manufacturing of European Tritium Breeder blanket modules

    NASA Astrophysics Data System (ADS)

    Poitevin, Y.; Aubert, Ph.; Diegele, E.; de Dinechin, G.; Rey, J.; Rieth, M.; Rigal, E.; von der Weth, A.; Boutard, J.-L.; Tavassoli, F.

    2011-10-01

    Europe has developed two reference Tritium Breeder Blankets concepts for a DEMO fusion reactor: the Helium-Cooled Lithium-Lead and the Helium-Cooled Pebble-Bed. Both are using the reduced-activation ferritic-martensitic EUROFER-97 steel as structural material and will be tested in ITER under the form of test blanket modules. The fabrication of their EUROFER structures requires developing welding processes like laser, TIG, EB and diffusion welding often beyond the state-of-the-art. The status of European achievements in this area is reviewed, illustrating the variety of processes and key issues behind retained options, in particular with respect to metallurgical aspects and mechanical properties. Fabrication of mock-ups is highlighted and their characterization and performances with respect to design requirements are reviewed.

  13. Neutronics R&D efforts in support of the European breeder blanket development programme

    NASA Astrophysics Data System (ADS)

    Fischer, U.; Batistoni, P.; Klix, A.; Kodeli, I.; Leichtle, D.; Perel, R. L.

    2009-06-01

    The recent progress in the R&D neutronics efforts spent in the EU to support the development of the HCLL and HCPB breeder blankets is presented. These efforts include neutronic design activities performed in the framework of the European DEMO reactor study, validation efforts by means of neutronics mock-up experiments using 14 MeV neutron generators and the development of dedicated computational tools such as the conversion software McCad for the automatic generation of a Monte Carlo geometry model from available CAD data, and the MCSEN code for Monte Carlo based calculations of sensitivities and uncertainties by using the track length estimator. The supporting validation effort is devoted to the capability of the neutronics tools and data to predict the tritium production and other nuclear responses of interest in neutronics mock-up experiments. Such an experiment has been conducted on a HCPB mock-up while another on a HCLL mock-up is in progress.

  14. Application OF LIBS To Estimate The Age Of Broiler Breeders

    NASA Astrophysics Data System (ADS)

    Salam, Z. Abdel; Harith, M. A.

    2011-09-01

    Laser Induced Breakdown Spectroscopy (LIBS) is a well-known spectrochemical elemental analysis technique. In our investigations of the LIBS spectra it has been found that there is a remarkable correlation between the ionic to atomic spectral lines emission ratio and the surface hardness of eggshell for two Different Broiler Breeder at different age. The proposed technique has been applied successfully in poultry science to estimate the age of broiler breeders by measuring the surface hardness of their eggshell. The experiments have been performed on two different strains, Arbor Acres plus (AAP) and Hubard Classic (HC), and the results were satisfactory.

  15. Immobilization of catalase on electrospun PVA/PA6-Cu(II) nanofibrous membrane for the development of efficient and reusable enzyme membrane reactor.

    PubMed

    Feng, Quan; Zhao, Yong; Wei, Anfang; Li, Changlong; Wei, Qufu; Fong, Hao

    2014-09-01

    In this study, a mat/membrane consisting of overlaid PVA/PA6-Cu(II) composite nanofibers was prepared via the electrospinning technique followed by coordination/chelation with Cu(II) ions; an enzyme of catalase (CAT) was then immobilized onto the PVA/PA6-Cu(II) nanofibrous membrane. The amount of immobilized catalase reached a high value of 64 ± 4.6 mg/g, while the kinetic parameters (Vmax and Km) of enzyme were 3774 μmol/mg·min and 41.13 mM, respectively. Furthermore, the thermal stability and storage stability of immobilized catalase were improved significantly. Thereafter, a plug-flow type of immobilized enzyme membrane reactor (IEMR) was assembled from the PVA/PA6-Cu(II)-CAT membrane. With the increase of operational pressure from 0.02 to 0.2 MPa, the flux value of IEMR increased from 0.20 ± 0.02 to 0.76 ± 0.04 L/m(2)·min, whereas the conversion ratio of H2O2 decreased slightly from 92 ± 2.5% to 87 ± 2.1%. After 5 repeating cycles, the production capacity of IEMR was merely decreased from 0.144 ± 0.006 to 0.102 ± 0.004 mol/m(2)·min. These results indicated that the assembled IEMR possessed high productivity and excellent reusability, suggesting that the IEMR based on electrospun PVA/PA6-Cu(II) nanofibrous membrane might have great potential for various applications, particularly those related to environmental protection. PMID:25093534

  16. Preliminary Evaluation of the Adequacy of Lithium Resources of the World and China for D-T Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Wang, Yongliang; Ni, Muyi; Jiang, Jieqiong; Wu, Yican; FDS-Team

    2012-07-01

    This paper studied the adequacy of the World and China lithium resources, considering the most promising uses in the future, involving nuclear fusion and electric-vehicles. The lithium recycle model for D-T fusion power plant and electric-vehicles, and the logistic growth prediction model of the primary energy for the World and China were constructed. Based on these models, preliminary evaluation of lithium resources adequacy of the World and China for D-T fusion reactors was presented under certain assumptions. Results show that: a. The world terrestrial reserves of lithium seems too limited to support a significant D-T power program, but the lithium reserves of China are relatively abundant, compared with the world case. b. The lithium resources contained in the oceans can be called the “permanent" energy. c. The change in 6Li enrichment has no obvious effect on the availability period of the lithium resources using FDS-II (Liquid Pb-17Li breeder blanket) type of reactors, but it has a stronger effect when PPCS-B (Solid Li4 SiO4 ceramics breeder blanket) is used.

  17. Detection and characterization of chicken anemia virus from commercial broiler breeder chickens

    PubMed Central

    Hailemariam, Zerihun; Omar, Abdul Rahman; Hair-Bejo, Mohd; Giap, Tan Ching

    2008-01-01

    Background Chicken anemia virus (CAV) is the causative agent of chicken infectious anemia (CIA). Study on the type of CAV isolates present and their genetic diversity, transmission to their progeny and level of protection afforded in the breeder farms is lacking in Malaysia. Hence, the present study was aimed to detect CAV from commercial broiler breeder farms and characterize CAV positive samples based on sequence and phylogenetic analysis of partial VP1 gene. Results A total of 12 CAV isolates from different commercial broiler breeder farms were isolated and characterized. Detection of CAV positive embryos by the PCR assay in the range of 40 to 100% for different farms indicated high level of occurrence of vertical transmission of viral DNA to the progeny. CAV antigen was detected in the thymus and in the bone marrow but not in spleen, liver, duodenum, ovary and oviduct by indirect immunoperoxidase staining. The 12 CAV isolates were characterized based on partial sequences of VP1 gene. Six isolates (MF1A, MF3C, M3B5, NF4A, P12B and P24A) were found to have maximum homology with previously characterized Malaysian isolate SMSC-1, four isolates (M1B1, NF3A, PYT4 and PPW4) with isolate BL-5 and the remaining two (NF1D and NF2C) have maximum homology both with isolates 3-1 and BL-5. Meanwhile, seven of the isolates with amino acid profile of 75-I, 97-L, 139-Q and 144-Q were clustered together in cluster I together with other isolates from different geographical places. The remaining five isolates with amino acid profile of 75-V, 97-M, 139-K and 144-E were grouped under cluster II. All the CAV isolates demonstrated omega values (Ka/Ks) of less than one (the values ranging from 0.07 to 0.5) suggesting the occurrence of purifying (negative) selection in all the studied isolates. Conclusion The present study showed that CAV is widespread in the studied commercial broiler breeder farms. The result also indicated the occurrence of genetic variability in local CAV isolates

  18. Bimodal space nuclear power system with fast reactor and Topaz II-type single-cell TFE

    NASA Astrophysics Data System (ADS)

    Ponomarev-Stepnoi, N. N.; Usov, V. A.; Ogloblin, B. G.; Shalaev, A. I.; Klimov, A. V.; Kirillov, E. Ya.; Shumov, D. P.; Radchenko, I. S.; Nicolaev, Y. V.

    1996-03-01

    The paper deals with characteristics and conceptual studies of a bimodal space thermionic system with a fast reactor and single-cell TFEs which is designed to operate in two modes: rated power mode providing power supply to space vehicle-mounted systems with energy consumption level of 10-80 kW(e) and forced thermal propulsion mode with thrust of 2200 N.

  19. Bimodal space nuclear power system with fast reactor and Topaz II-type single-cell TFE

    SciTech Connect

    Ponomarev-Stepnoi, N.N.; Usov, V.A.; Ogloblin, B.G.; Shalaev, A.I.; Klimov, A.V.; Kirillov, E.Y.; Shumov, D.P.; Radchenko, I.S.; Nicolaev, Y.V.

    1996-03-01

    The paper deals with characteristics and conceptual studies of a bimodal space thermionic system with a fast reactor and single-cell TFEs which is designed to operate in two modes: rated power mode providing power supply to space vehicle-mounted systems with energy consumption level of 10{endash}80 kW(e) and forced thermal propulsion mode with thrust of 2200 N. {copyright} {ital 1996 American Institute of Physics.}

  20. The thermal decomposition of the benzyl radical in a heated micro-reactor. II. Pyrolysis of the tropyl radical.

    PubMed

    Buckingham, Grant T; Porterfield, Jessica P; Kostko, Oleg; Troy, Tyler P; Ahmed, Musahid; Robichaud, David J; Nimlos, Mark R; Daily, John W; Ellison, G Barney

    2016-07-01

    Cycloheptatrienyl (tropyl) radical, C7H7, was cleanly produced in the gas-phase, entrained in He or Ne carrier gas, and subjected to a set of flash-pyrolysis micro-reactors. The pyrolysis products resulting from C7H7 were detected and identified by vacuum ultraviolet photoionization mass spectrometry. Complementary product identification was provided by infrared absorption spectroscopy. Pyrolysis pressures in the micro-reactor were roughly 200 Torr and residence times were approximately 100 μs. Thermal cracking of tropyl radical begins at 1100 K and the products from pyrolysis of C7H7 are only acetylene and cyclopentadienyl radicals. Tropyl radicals do not isomerize to benzyl radicals at reactor temperatures up to 1600 K. Heating samples of either cycloheptatriene or norbornadiene never produced tropyl (C7H7) radicals but rather only benzyl (C6H5CH2). The thermal decomposition of benzyl radicals has been reconsidered without participation of tropyl radicals. There are at least three distinct pathways for pyrolysis of benzyl radical: the Benson fragmentation, the methyl-phenyl radical, and the bridgehead norbornadienyl radical. These three pathways account for the majority of the products detected following pyrolysis of all of the isotopomers: C6H5CH2, C6H5CD2, C6D5CH2, and C6H5 (13)CH2. Analysis of the temperature dependence for the pyrolysis of the isotopic species (C6H5CD2, C6D5CH2, and C6H5 (13)CH2) suggests the Benson fragmentation and the norbornadienyl pathways open at reactor temperatures of 1300 K while the methyl-phenyl radical channel becomes active at slightly higher temperatures (1500 K). PMID:27394106

  1. [Safety systems in nuclear power plants with pressurized water reactors and demonstration of their function in technical scale experiments. II].

    PubMed

    Hennies, H H

    1987-11-01

    Analyses and experiments carried out during the last decade on the sequence and consequences of accidents in German pressurized water reactors have shown that the functioning capability of the safety systems is guaranteed for the case of the MCA, the maximum credible accident. For the case of core meltdown, simulation experiments have also made it evident that the consequences remain largely restricted to the plant proper. PMID:3431585

  2. The thermal decomposition of the benzyl radical in a heated micro-reactor. II. Pyrolysis of the tropyl radical

    NASA Astrophysics Data System (ADS)

    Buckingham, Grant T.; Porterfield, Jessica P.; Kostko, Oleg; Troy, Tyler P.; Ahmed, Musahid; Robichaud, David J.; Nimlos, Mark R.; Daily, John W.; Ellison, G. Barney

    2016-07-01

    Cycloheptatrienyl (tropyl) radical, C7H7, was cleanly produced in the gas-phase, entrained in He or Ne carrier gas, and subjected to a set of flash-pyrolysis micro-reactors. The pyrolysis products resulting from C7H7 were detected and identified by vacuum ultraviolet photoionization mass spectrometry. Complementary product identification was provided by infrared absorption spectroscopy. Pyrolysis pressures in the micro-reactor were roughly 200 Torr and residence times were approximately 100 μs. Thermal cracking of tropyl radical begins at 1100 K and the products from pyrolysis of C7H7 are only acetylene and cyclopentadienyl radicals. Tropyl radicals do not isomerize to benzyl radicals at reactor temperatures up to 1600 K. Heating samples of either cycloheptatriene or norbornadiene never produced tropyl (C7H7) radicals but rather only benzyl (C6H5CH2). The thermal decomposition of benzyl radicals has been reconsidered without participation of tropyl radicals. There are at least three distinct pathways for pyrolysis of benzyl radical: the Benson fragmentation, the methyl-phenyl radical, and the bridgehead norbornadienyl radical. These three pathways account for the majority of the products detected following pyrolysis of all of the isotopomers: C6H5CH2, C6H5CD2, C6D5CH2, and C6H513CH2. Analysis of the temperature dependence for the pyrolysis of the isotopic species (C6H5CD2, C6D5CH2, and C6H513CH2) suggests the Benson fragmentation and the norbornadienyl pathways open at reactor temperatures of 1300 K while the methyl-phenyl radical channel becomes active at slightly higher temperatures (1500 K).

  3. Moon base reactor system

    NASA Technical Reports Server (NTRS)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  4. Run - Beyond - Cladding - Breach (RBCB) test results for the Integral Fast Reactor (IFR) metallic fuels program

    SciTech Connect

    Batte, G.L. ); Hoffman, G.L. )

    1990-01-01

    In 1984 Argonne National Laboratory (ANL) began an aggressive program of research and development based on the concept of a closed system for fast-reactor power generation and on-site fuel reprocessing, exclusively designed around the use of metallic fuel. This is the Integral Fast Reactor (IFR). Although the Experimental Breeder Reactor-II (EBR-II) has used metallic fuel since its creation 25 yeas ago, in 1985 ANL began a study of the characteristics and behavior of an advanced-design metallic fuel based on uranium-zirconium (U-Zr) and uranium-plutonium-zirconium (U-Pu-Zr) alloys. During the past five years several areas were addressed concerning the performance of this fuel system. In all instances of testing the metallic fuel has demonstrated its ability to perform reliably to high burnups under varying design conditions. This paper will present one area of testing which concerns the fuel system's performance under breach conditions. It is the purpose of this paper to document the observed post-breach behavior of this advanced-design metallic fuel. 2 figs., 1 tab.

  5. Measurement of DNA damage induced by irradiation with gamma-rays from a TRIGA Mark II research reactor in human cells using Fast Micromethod.

    PubMed

    Hassanein, Hamdy; Müller, Claudia I; Schlösser, Dietmar; Kratz, Karl-Ludwig; Senyuk, Olga F; Schröder, Heinz C

    2002-06-01

    The Fast Micromethod is a novel quick and convenient microplate assay for determination of DNA single-strand breaks. This method measures the rate of unwinding of cellular DNA upon exposure to alkaline conditions using a fluorescent dye which preferentially binds to double-stranded DNA. Here we applied this method to determine the levels of DNA single-strand breaks in HeLa cells induced by y-irradiation deriving from fission isotopes and activation products at the TRIGA Mark II research reactor in Mainz. An increased strand scission factor (SSF) value, which is indicative for DNA damage, was found at doses of 1 Gy and higher. A similar increase in SSF value, which further increased in a dose-dependent manner, was found in human peripheral blood mononuclear cells after irradiation with 6 MV X-rays from a linear accelerator to give a total exposure of 0.5 to 10 Gy. PMID:12064446

  6. Chromosome aberrations induced in human lymphocytes by U-235 fission neutrons: I. Irradiation of human blood samples in the "dry cell" of the TRIGA Mark II nuclear reactor.

    PubMed

    Fajgelj, A; Lakoski, A; Horvat, D; Remec, I; Skrk, J; Stegnar, P

    1991-11-01

    A set-up for irradiation of biological samples in the TRIGA Mark II research reactor in Ljubljana is described. Threshold activation detectors were used for characterisation of the neutron flux, and the accompanying gamma dose was measured by TLDs. Human peripheral blood samples were irradiated "in vitro" and biological effects evaluated according to the unstable chromosomal aberrations induced. Biological effects of two types of cultivation of irradiated blood samples, the first immediately after irradiation and the second after 96 h storage, were studied. A significant difference in the incidence of chromosomal aberrations between these two types of samples was obtained, while our dose-response curve fitting coefficients alpha 1 = (7.71 +/- 0.09) x 10(-2) Gy-1 (immediate cultivation) and alpha 2 = (11.03 +/- 0.08) x 10(-2) Gy-1 (96 h delayed cultivation) are in both cases lower than could be found in the literature. PMID:1962281

  7. Optimization of a partially non-magnetic primary radiation. shielding for the triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II

    NASA Astrophysics Data System (ADS)

    Pyka, N. M.; Noack, K.; Rogov, A.

    Monte Carlo simulations have been used to optimize the monochromator shielding of the polarized cold-neutron triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II. By using the Monte Carlo program MCNP-4B, the density of the total spectrum of incoming neutrons and γ radiation from the beam tube SR-2 has been determined during the three-dimensional diffusion process in different types of heavy concrete and other absorbing material. Special attention has been paid to build a compact and highly efficient shielding, partially non-magnetic, with a total biological radiation dose of less than 10 μSv/h at its outsides. Especially considered was the construction of an albedo reducer, which serves to reduce the background in the experiment outside the shielding.

  8. Nuclear reactor

    DOEpatents

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  9. Concept for Dismantling the Reactor Vessel and the Biological Shield of the Compact Sodium-Cooled Nuclear Reactor Facility (KNK)

    SciTech Connect

    Hillebrand, I.; Benkert, J.

    2002-02-27

    The Compact Sodium-cooled Nuclear Reactor Facility (KNK) was an experimental nuclear power plant of 20 MW electric power erected on the premises of the Karlsruhe Research Center. The plant was initially run as KNK I with a thermal core between 1971 and 1974 and then, between 1977 and 1991, with a fast core as the KNK II fast breeder plant. Under the decommissioning concept, the plant is to be decommissioned completely to green field conditions at the end of 2005 in ten steps, i.e. under the corresponding ten decommissioning permits. To this day, nine decommissioning permits have been issued, the first one in 1993 and the most recent one, number nine, in 2001. The decommissioning and demolition activities covered by decommissioning permits 1 to 7 have been completed. Under the 8th Decommissioning Permit, the components of the primary system and the rotating reactor top shield are to be removed by late 2001. Under the 9th Decommissioning Permit, the reactor vessel with its internals, the primary shield, and the biological shield are to be dismantled. The residual sodium volume in the reactor vessel was estimated to amount to approx. 30 l. The maximum Co-60 activation is on the order of 107-108 Bq/g; the maximum dose rate in the middle of the vessel was measured in April 1997 to be 55 Sv/h. The difficulty involved especially in dismantling KNK, on the one hand, is posed by the residual sodium in the plant, which determines the choice of neither wet nor thermical techniques to be used in disassembly. Another difficulty is caused by the depth of activation by fast neutrons, as a result of which not only the reactor vessel proper, but also the entire primary shield (60 cm of grey cast iron) and large parts of the biological shield must be disassembled and disposed of under remote control.

  10. Automated start-up of EBR-II: A preview

    SciTech Connect

    Kisner, R.A.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) and Argonne National Laboratory (ANL) are undertaking a joint project to develop control philosophies, strategies, and algorithms for computer control of the start-up mode of the Experimental Breeder Reactor II (EBR-II). The major objective of this project is to show that advanced liquid-metal reactor (LMR) plants can be operated from low power to full power using computer control. Development of an automated control system with this objective in view will help resolve specific issues and provide proof through demonstration that automatic control for plant start-up is feasible. This paper describes the approach that will be used to develop such a system and some of the features it is expected to have. Structured, rule-based methods, which will provide start-up capability from a variety of initial plant conditions and degrees of equipment operability, will be used for accomplishing mode changes during plant start-up. Several innovative features will be incorporated such as signal, command, and strategy validation to maximize reliability, flexibility to accommodate a wide range of plant conditions, and overall utility. Continuous control design will utilize figures of merit to evaluate how well the controller meets the mission requirements. The operator interface will have unique ''look ahead'' features to let the operator see what will happen next. 15 refs., 7 figs., 1 tab.

  11. Airlift column photobioreactors for Porphyridium sp. culturing: Part II. verification of dynamic growth rate model for reactor performance evaluation.

    PubMed

    Luo, Hu-Ping; Al-Dahhan, Muthanna H

    2012-04-01

    Dynamic growth rate model has been developed to quantify the impact of hydrodynamics on the growth of photosynthetic microorganisms and to predict the photobioreactor performance. Rigorous verification of such reactor models, however, is rare in the literature. In this part of work, verification of a dynamic growth rate model developed in Luo and Al-Dahhan (2004) [Biotech Bioeng 85(4): 382-393] was attempted using the experimental results reported in Part I of this work and results from literature. The irradiance distribution inside the studied reactor was also measured at different optical densities and successfully correlated by the Lambert-Beer Law. When reliable hydrodynamic data were used, the dynamic growth rate model successfully predicted the algae's growth rate obtained in the experiments in both low and high irradiance regime indicating the robustness of this model. The simulation results also indicate the hydrodynamics is significantly different between the real algae culturing system and an air-water system that signifies the importance in using reliable data input for the growth rate model. PMID:22068388

  12. Swelling, microstructural development and helium effects in type 316 stainless steel irradiated in HFIR and EBR-II

    SciTech Connect

    Maziasz, P.J.; Grossbeck, M.L.

    1981-01-01

    This work examines the swelling and microstructural development of a single heat of 20%-cold-worked type 316 stainless steel irradiated to produce displacement damage and a high, continuous helium generation rate, in the High Flux Isotope Reactor (HFIR). Similar irradiation of the same heat of steel in the Experimental Breeder Reactor (EBR)-II is used as a base line for comparing displacement damage accompanying a very low continuous helium generation rate. At temperatures above and below the void swelling regime (approx. 350 to 625/sup 0/C) swelling is greater in HFIR than in EBR-II. In the temprature range of 350 to 625/sup 0/C, cavity formation, precipitation and dislocation recovery are both enhanced and accelerated in HFIR, often causing swelling at lower dose than in EBR-II. In HFIR, however, cavities appear to be bubbles rather than voids. They are about 10 times smaller and 20 to 50 times more numerous than voids in EBR-II. Thus, the swelling becomes greater in EBR-II than in HFIR for 20%-CW 316 in the void swelling temperature ranges as fluence increases. Such differences in swelling and microstructural behavior must be understood in order to anticipate the behavior of materials during fusion irradiation.

  13. Final Report on Utilization of TRU TRISO Fuel as Applied to HTR Systems Part II: Prismatic Reactor Cross Section Generation

    SciTech Connect

    Vincent Descotes

    2011-03-01

    The deep-burn prismatic high temperature reactor is made up of an annular core loaded with transuranic isotopes and surrounded in the center and in the periphery by reflector blocks in graphite. This disposition creates challenges for the neutronics compared to usual light water reactor calculation schemes. The longer mean free path of neutrons in graphite affects the neutron spectrum deep inside the blocks located next to the reflector. The neutron thermalisation in the graphite leads to two characteristic fission peaks at the inner and outer interfaces as a result of the increased thermal flux seen in those assemblies. Spectral changes are seen at least on half of the fuel blocks adjacent to the reflector. This spectral effect of the reflector may prevent us from successfully using the two step scheme -lattice then core calculation- typically used for light water reactors. We have been studying the core without control mechanisms to provide input for the development of a complete calculation scheme. To correct the spectrum at the lattice level, we have tried to generate cross-sections from supercell calculations at the lattice level, thus taking into account part of the graphite surrounding the blocks of interest for generating the homogenised cross-sections for the full-core calculation. This one has been done with 2 to 295 groups to assess if increasing the number of groups leads to more accurate results. A comparison with a classical single block model has been done. Both paths were compared to a reference calculation done with MCNP. It is concluded that the agreement with MCNP is better with supercells, but that the single block model remains quite close if enough groups are kept for the core calculation. 26 groups seems to be a good compromise between time and accu- racy. However, some trials with depletion have shown huge variations of the isotopic composition across a block next to the reflector. It may imply that at least an in- core depletion for the

  14. The dependence of helium generation rate on nickel content of Fe Cr Ni alloys irradiated to high dpa levels in EBR-II

    NASA Astrophysics Data System (ADS)

    Garner, F. A.; Oliver, B. M.; Greenwood, L. R.

    1998-10-01

    Fusion-relevant helium-effects experiments conducted on austenitic steels in the Materials Open Test Assembly (MOTA) of the Fast Flux Test Facility (FFTF) fast reactor had to recognize the contributions of both the high neutron energy (n,α) reactions and that of the 58Ni(n,γ) 59Ni(n,α) reaction sequence with low energy neutrons. An experiment conducted in the harder neutron spectra found within the core of Experimental Breeder Reactor-II (EBR-II) has shown that the helium in this reactor was generated almost exclusively from the interaction of high energy neutrons with the natural isotopes of nickel. There was very little contribution from 59Ni. The helium production was found to scale directly with the nickel content over the range 25-75% Ni. Even at very high neutron exposures, the helium production in such reactors can be predicted within 5% accuracy on the basis of high energy reactions, as demonstrated by an experiment conducted on three Fe-15Cr-Ni ternary alloys irradiated to doses of 75-131 dpa in EBR-II.

  15. Expert system driven fuzzy control application to power reactors

    SciTech Connect

    Tsoukalas, L.H.; Berkan, R.C.; Upadhyaya, B.R.; Uhrig, R.E.

    1990-12-31

    For the purpose of nonlinear control and uncertainty/imprecision handling, fuzzy controllers have recently reached acclaim and increasing commercial application. The fuzzy control algorithms often require a ``supervisory`` routine that provides necessary heuristics for interface, adaptation, mode selection and other implementation issues. Performance characteristics of an on-line fuzzy controller depend strictly on the ability of such supervisory routines to manipulate the fuzzy control algorithm and enhance its control capabilities. This paper describes an expert system driven fuzzy control design application to nuclear reactor control, for the automated start-up control of the Experimental Breeder Reactor-II. The methodology is verified through computer simulations using a valid nonlinear model. The necessary heuristic decisions are identified that are vitally important for the implemention of fuzzy control in the actual plant. An expert system structure incorporating the necessary supervisory routines is discussed. The discussion also includes the possibility of synthesizing the fuzzy, exact and combined reasoning to include both inexact concepts, uncertainty and fuzziness, within the same environment.

  16. Expert system driven fuzzy control application to power reactors

    SciTech Connect

    Tsoukalas, L.H.; Berkan, R.C.; Upadhyaya, B.R.; Uhrig, R.E.

    1990-01-01

    For the purpose of nonlinear control and uncertainty/imprecision handling, fuzzy controllers have recently reached acclaim and increasing commercial application. The fuzzy control algorithms often require a supervisory'' routine that provides necessary heuristics for interface, adaptation, mode selection and other implementation issues. Performance characteristics of an on-line fuzzy controller depend strictly on the ability of such supervisory routines to manipulate the fuzzy control algorithm and enhance its control capabilities. This paper describes an expert system driven fuzzy control design application to nuclear reactor control, for the automated start-up control of the Experimental Breeder Reactor-II. The methodology is verified through computer simulations using a valid nonlinear model. The necessary heuristic decisions are identified that are vitally important for the implemention of fuzzy control in the actual plant. An expert system structure incorporating the necessary supervisory routines is discussed. The discussion also includes the possibility of synthesizing the fuzzy, exact and combined reasoning to include both inexact concepts, uncertainty and fuzziness, within the same environment.

  17. Role of the breeder in long-term energy economics

    SciTech Connect

    Kosobud, R.F.; Daly, T.A.; Chang, Y.I.

    1982-01-01

    Private and public decisions affecting the use of nuclear and other energy technologies over a long-run time horizon were studied using the ETA-MACRO model which provides for economic- and energy-sector interactions. The impact on the use of competing energy technologies of a public decision to apply benefit-cost analysis to the production of carbon dioxide that enters the atmosphere is considered. Assuming the public choice is to impose an appropriate penalty tax on those technologies which generate CO/sub 2/ and to allow decentralized private decisions to choose the optimal mix of energy technologies that maximize a nonlinear objective function subject to constraints, the study showed that breeder technology provides a much-larger share of domestically consumed energy. Having the breeder technology available as a substitute permits control of CO/sub 2/ without significant reductions in consumption or gross national product growth paths.

  18. Charge breeder for the SPIRAL1 upgrade: Preliminary results

    NASA Astrophysics Data System (ADS)

    Maunoury, L.; Delahaye, P.; Dubois, M.; Angot, J.; Sole, P.; Bajeat, O.; Barton, C.; Frigot, R.; Jeanne, A.; Jardin, P.; Kamalou, O.; Lecomte, P.; Osmond, B.; Peschard, G.; Lamy, T.; Savalle, A.

    2016-02-01

    In the framework of the SPIRAL1 upgrade under progress at the GANIL lab, the charge breeder based on a LPSC Phoenix ECRIS, first tested at ISOLDE has been modified to benefit of the last enhancements of this device from the 1+/n+ community. The modifications mainly concern the 1 + optics, vacuum techniques, and the RF—buffer gas injection into the charge breeder. Prior to its installation in the midst of the low energy beam line of the SPIRAL1 facility, it has been decided to qualify its performances and several operation modes at the test bench of LPSC lab. This contribution shall present preliminary results of experiments conducted at LPSC concerning the 1 + to n+ conversion efficiencies for noble gases as well as for alkali elements and the corresponding transformation times.

  19. Advanced burner test reactor preconceptual design report.

    SciTech Connect

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16

    advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

  20. 56. ARAII. View inside reactor building looking at SL1 reactor ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    56. ARA-II. View inside reactor building looking at SL-1 reactor vessel. November 19, 1957. Ineel photo no. 57-5864. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  1. Reproduction responses of broiler-breeders to anticoccidial agents.

    PubMed

    Jones, J E; Solis, J; Hughes, B L; Castaldo, D J; Toler, J E

    1990-01-01

    Two experiments were conducted to determine the effects of anticoccidial agents on production and reproduction of broiler breeders. In Experiment 1, nicarbazin (NCZ) was fed at 20, 50, and 100 ppm. There was no depression in egg production, egg weight, or fertility from feeding these levels. As level of NCZ increased, there was a linear decrease in hatchability. The amount of 4,4'-dinitrocarbanilide (DNC) in the egg yolks increased linearly as the levels of NCZ went up; the degree of egg-shell depigmentation was directly related to the level of NCZ fed starting at 50 ppm. Experiment 2 utilized a different strain of broiler breeders. Halofuginone (3 ppm), maduramicin (5 ppm), monensin (100 ppm), narasin (70 ppm), NCZ (125 ppm), robenidine (33 ppm), and salinomycin (60 ppm) were fed to broiler breeders at the levels listed. Only NCZ reduced egg production. Narasin induced a reduction in egg weight. Both narasin and salinomycin caused a significant drop in hatchability. Feeding NCZ also induced a rapid and more severe decrease in hatchability. Monensin was the only anticoccidial agent that reduced fertility. Halofuginone, maduramicin, and robenidine had no biologically significant effect on henday production, egg weight, hatch of fertile eggs, or shell depigmentation. Feeding NCZ at 125 ppm caused a complete bleaching of brown-shell eggs by the 3rd consecutive day of treatment; but 7 days after NCZ was withdrawn from the feed, pigmentation returned to the pretreatment level. PMID:2320530

  2. Overview of EU activities on DEMO liquid metal breeder blanket

    SciTech Connect

    Giancarli, L.; Proust, E.

    1994-12-31

    The European test-blanket development programme, started in 1988, is aiming at the selection by 1995 of two DEMO-relevant blanket lines to be tested in ITER. At present, four lines of blanket are under development, two of them using solid and the other two liquid breeder materials. As far as liquid breeders are concerned, two lines of blankets have been selected within the European Union, the water-cooled lithium-lead (the eutectic Pb-17Li) blankets and the dual-coolant Pb-17Li blankets. Designs have been developed considering an agreed set of DEMO specifications, such as, for instance, a fusion power of 2,200 MW, a neutron wall-loading of 2MW/m{sup 2}, a life-time of 20,000 hours, and the use of martensitic steel as a structural material. Moreover, an experimental program has been set up in order to address the main critical issues for each line. The present paper gives an overview of both design and experimental activities within the European Union concerning these two lines of liquid breeder blankets.

  3. Can breeder reproductive status, performance and egg quality be enhanced by supplementation and transition of n-3 fatty acids?

    PubMed

    Delezie, E; Koppenol, A; Buyse, J; Everaert, N

    2016-08-01

    The aim of this experiment was to investigate the effect of n-3 fatty acid (FA) supplemented diets on breeder performance, productivity and egg quality. Breeders (n = 480) were fed the supplemented diet from 18 weeks onwards; the inclusion level of n-3 FA was increased from 1.5% to 3.0% from 34 weeks of age onwards until 48 weeks of age. Ross-308 broiler breeders (n = 480) were fed one of four different diets: a basal diet rich in n-6 FA (control diet) or one of three diets rich in n-3 FA. For the n-3 FA diets, eicosapentaenoic acid (EPA, 20:5 n-3) and docosahexaenoic acid (DHA, 22:6 n-3) were fed to the broiler breeders at different ratios formulated to obtain EPA/DHA ratios of 1/1, 1/2 or 2/1. Differences in performance, reproduction and egg quality parameters due to n-3 supplementation were noted more for the 1.5% followed by the 3.0% fed broilers than their 1.5% supplemented counterparts. Egg weight (p < 0.001) and egg mass (p = 0.003) were significantly lower and feed conversion (p = 0.008) significantly higher for the n-3 FA (at 3.0% inclusion level) fed broilers compared to the control group. For the EPA- and DHA-fed breeders, a higher proportional abdominal fat percentage (p = 0.025) and proportional albumen weight (%) (p = 0.041) were found respectively. Dietary treatments did not affect reproduction. It can be concluded that the results of the present experiment indicate no significant differences between treatments at 1.5% inclusion levels. However, increasing this level to 3.0% is not recommended due to the rather negative effects on the measured parameters. It should be further investigated whether these adverse effects were obtained due to (i) the higher supplementation level, (ii) combining a supplementation level of 1.5% with 3% or (iii) the duration of supplementation. PMID:26854179

  4. A New Search for the Atomic EDM of 129Xe at FRM-II (Munich Research Reactor)

    NASA Astrophysics Data System (ADS)

    Singh, Jaideep; Fierlinger, Peter; Kraegeloh, Eva; Kuchler, Florian; Lins, Tobias; Marino, Mike; Meinel, Jonas; Neissen, Benjamin; Stuiber, Stefan; Burghoff, Martin; Fan, Isaac; Kilian, Wolfgang; Knappe-Grueneberg, Silvia; Schnabel, Allard; Seifert, Frank; Trahms, Lutz; Voigt, Jens; Chupp, Tim; Degenkolb, Skyler; Gong, Fei; Sachdeva, Natasha; Babcock, Earl

    2014-09-01

    Electric dipole moments (EDMs) arise due to the breaking of time-reversal or, equivalently, CP -symmetry. Although all searches have so far only set upper limits on EDMs, the motivation for more sensitive searches is stronger than ever. The present limit of 6 ×10-27 e * cm (95% CL) for the 129Xe EDM helps constrain CP -violating parameters within nuclei. A new effort at FRM-II incorporating a 3He comagnetometer can potentially improve this limit by over three orders of magnitude. The noble gas mixture is polarized by spin-exchange optical pumping and then transferred into a high-performance magnetically shielded room. A SQUID magnetometer array measures the precession frequencies in the presence of applied electric- & magnetic-fields. Recent test runs indicate that the experiment is capable of an EDM sensitivity of 10-28 e * cm in one day.

  5. A New Search for the Atomic EDM of 129 Xe at FRM-II (Munich Research Reactor)

    NASA Astrophysics Data System (ADS)

    Kuchler, Florian; Fierlinger, Peter; Kraegeloh, Eva; Lins, Tobias; Marino, Mike; Meinel, Jonas; Niessen, Benjamin; Stuiber, Stefan; Burghoff, Martin; Fan, Isaac; Kilian, Wolfgang; Knappe-Grueneberg, Silvia; Schnabel, Allard; Seifert, Frank; Trahms, Lutz; Voigt, Jens; Chupp, Tim; Degenkolb, Skyler; Gong, Fei; Sachdeva, Natasha; Babcock, Earl; Singh, Jaideep

    2015-04-01

    Electric dipole moments (EDMs) arise due to the breaking of time-reversal or, equivalently, CP-symmetry. Although all searches have so far only set upper limits on EDMs, the motivation for more sensitive searches is stronger than ever. The present limit of 6 × 10-27 e*cm (95 % CL) for the 129 Xe EDM helps constrain CP-violating parameters within nuclei. A new effort at FRM-II incorporating a 3 He comagnetometer can potentially improve this limit by over three orders of magnitude. The noble gas mixture is polarized by spin-exchange optical pumping and then transferred into a high-performance magnetically shielded room. A SQUID magnetometer array measures the precession frequencies in the presence of applied electric- and magnetic-fields. Recent test runs indicate that the experiment is capable of an EDM sensitivity of 10-28 e*cm in one day.

  6. Operational experience with the Argonne National Laboratory Californium Rare Ion Breeder Upgrade facility and electron cyclotron resonance charge breeder

    NASA Astrophysics Data System (ADS)

    Vondrasek, R.; Clark, J.; Levand, A.; Palchan, T.; Pardo, R.; Savard, G.; Scott, R.

    2014-02-01

    The Californium Rare Ion Breeder Upgrade (CARIBU) of the Argonne National Laboratory Argonne Tandem Linac Accelerator System (ATLAS) facility provides low-energy and accelerated neutron-rich radioactive beams to address key nuclear physics and astrophysics questions. A 350 mCi 252Cf source produces fission fragments which are thermalized and collected by a helium gas catcher into a low-energy particle beam with a charge of 1+ or 2+. An electron cyclotron resonance (ECR) ion source functions as a charge breeder in order to raise the ion charge sufficiently for acceleration in the ATLAS linac. The ECR charge breeder has achieved stable beam charge breeding efficiencies of 10.1% for 23Na7+, 17.9% for 39K10+, 15.6% for 84Kr17+, and 12.4% for 133Cs27+. For the radioactive beams, a charge breeding efficiency of 11.7% has been achieved for 143Cs27+ and 14.7% for 143Ba27+. The typical breeding times are 10 ms/charge state, but the source can be tuned such that this value increases to 100 ms/charge state with the best breeding efficiency corresponding to the longest breeding times—the variation of efficiencies with breeding time will be discussed. Efforts have been made to characterize and reduce the background contaminants present in the ion beam through judicious choice of q/m combinations. Methods of background reduction are being investigated based upon plasma chamber cleaning and vacuum practices.

  7. Characterization of degraded EBR-II fuel from the ICPP-603 basin: National spent nuclear fuel program, FY 1999 final report

    SciTech Connect

    Pahl, R. G.

    2000-04-17

    Characterization data is reported for sodium bonded Experimental Breeder Reactor II (EBR-II) fuel which had been stored underwater in containers since the late 1970's. Sixteen stainless steel storage containers were retrieved from the ICPP-603 storage pool at the Idaho National Engineering and Environmental Laboratory (INEEL) in Idaho. Ten of the containers had leaked water due to improper sealing. In the container chosen for detailed destructive analysis, the stainless steel cladding on the uranium alloy fuel had ruptured and fuel oxide particulate formed and filled the bottom of the container. Headspace gas analysis determined that greater than 99% hydrogen was present. Cesium-137, which had leached out of the fuel during the aqueous corrosion process, dominated the radionuclide source term of the water. The metallic sodium from the fuel element bond had reacted with the water, forming a caustic solution of NaOH.

  8. The role of feeding regimens in regulating metabolism of sexually mature broiler breeders.

    PubMed

    Ekmay, R D; de Beer, M; Rosebrough, R W; Richards, M P; McMurtry, J P; Coon, C N

    2010-06-01

    A trial was conducted to determine the effects of different rearing feed regimens on plasma hormone and metabolite levels and hepatic lipid metabolism and gene expression on sexually mature broiler breeders. Cobb 500 birds were divided into 2 groups at 4 wk and fed either an everyday (ED) or skip-a-day (SKP) regimen. At 24 wk of age, all birds were switched over to an ED regimen. At 26.4 wk, breeder hens were randomly selected and killed at intervals after feeding. Livers were sampled from 4 hens at 4-h intervals for 24 h for a total of 28 samples per treatment. Blood was sampled from 4 hens per sampling time; sampling times were 0, 30, and 60 min and 2 and 4 h after feeding and then every 4 h up to 24 h for a total of 36 samples per treatment. Main feeding regimen, time, and interaction effects were analyzed. Significant interaction effects were found between time and feeding regimen for acetyl-coenzyme A carboxylase and malic enzyme mRNA expression. The peak for acetyl-coenzyme A carboxylase expression was higher in ED-reared birds, whereas the peak for malic enzyme expression was higher in SKP-reared birds. Overall, plasma levels of insulin-like growth factor-II were higher in SKP-reared birds. Overall, plasma corticosterone levels were also higher in SKP-reared birds and significant interaction effects between time and feeding regimen were seen. The expression of apolipoprotein A1 was significantly higher in ED-reared birds: significant interaction effects were also noted. Other researchers also found some of the differences observed in the present study in 16-wk-old pullets. In summary, different feeding regimens alter metabolic responses, some of which carry over into sexual maturity. PMID:20460664

  9. UCLA program in reactor studies: The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

  10. Reactor control rod timing system

    DOEpatents

    Wu, Peter T. K.

    1982-01-01

    A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  11. Inertial-fusion-reactor studies at Lawrence Livermore National Laboratory

    SciTech Connect

    Monsler, M.J.; Meier, W.R.

    1982-08-01

    We present results of our reactor studies for inertial-fusion energy production. Design studies of liquid-metal wall chambers have led to reactors that are remarkably simple in design, and that promise long life and low cost. Variants of the same basic design, called HYLIFE, can be used for electricity production, as a fissile-fuel factory, a dedicated tritium breeder, or hybrids of each.

  12. Conceptual design of a laser-fusion power plant. Part II. Two technical options: 1. JADE reactor; 2. Heat transfer by heat pipes

    SciTech Connect

    Not Available

    1981-07-01

    A laser fusion reactor concept is described that employs liquid metal walls. The concept envisions a porous medium, called the JADE, of specific geometry lining the reactor cavity. Some advantages and disadvantages of the concept are pointed out. The possibility of using heat pipes for passive cooling in ICF reactors is discussed. Some of the problems are outlined. (MOW)

  13. Campylobacter epidemiology from breeders to their progeny in Eastern Spain.

    PubMed

    Ingresa-Capaccioni, S; Jiménez-Trigos, E; Marco-Jiménez, F; Catalá, P; Vega, S; Marin, C

    2016-03-01

    While horizontal transmission is a route clearly linked to the spread of Campylobacter at the farm level, few studies support the transmission of Campylobacter spp. from breeder flocks to their offspring. Thus, the present study was carried out to investigate the possibility of vertical transmission. Breeders were monitored from the time of housing day-old chicks, then throughout the laying period (0 to 60 wk) and throughout their progeny (broiler fattening, 1 to 42 d) until slaughter. All samples were analyzed according with official method ISO 10272:2006. Results revealed that on breeder farms, Campylobacter isolation started from wk 16 and reached its peak at wk 26, with 57.0% and 93.2% of positive birds, respectively. After this point, the rate of positive birds decreased slightly to 86.0% at 60 wk. However, in broiler production all day-old chicks were found negative for Campylobacter spp, and the bacteria was first isolated at d 14 of age (5.0%), with a significant increase in detection during the fattening period with 62% of Campylobacter positive animals at the end of the production cycle. Moreover, non-positive sample was determined from environmental sources. These results could be explained because Campylobacter may be in a low concentration or in a non-culturable form, as there were several studies that successfully detected Campylobacter DNA, but failed to culture. This form can survive in the environment and infect successive flocks; consequently, further studies are needed to develop more modern, practical, cost-effective and suitable techniques for routine diagnosis. PMID:26628341

  14. Reactivity control assembly for nuclear reactor. [LMFBR

    DOEpatents

    Bollinger, L.R.

    1982-03-17

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  15. Group size adjustment to ecological demand in a cooperative breeder

    PubMed Central

    Zöttl, Markus; Frommen, Joachim G.; Taborsky, Michael

    2013-01-01

    Environmental factors can determine which group size will maximize the fitness of group members. This is particularly important in cooperative breeders, where group members often serve different purposes. Experimental studies are yet lacking to check whether ecologically mediated need for help will change the propensity of dominant group members to accept immigrants. Here, we manipulated the perceived risk of predation for dominant breeders of the cooperatively breeding cichlid fish Neolamprologus pulcher to test their response to unrelated and previously unknown immigrants. Potential immigrants were more readily accepted if groups were exposed to fish predators or egg predators than to herbivorous fish or control situations lacking predation risk. Our data are consistent with both risk dilution and helping effects. Egg predators were presented before spawning, which might suggest that the fish adjust acceptance rates also to a potential future threat. Dominant group members of N. pulcher apparently consider both present and future need of help based on ecological demand. This suggests that acceptance of immigrants and, more generally, tolerance of group members on demand could be a widespread response to ecological conditions in cooperatively breeding animals. PMID:23390105

  16. Experimental reproduction of enterococcal spondylitis in male broiler breeder chickens.

    PubMed

    Martin, Leslie T; Martin, Michael P; Barnes, H John

    2011-06-01

    There has been a recent emergence of epidemic spinal infections with necrosis causing lameness and mortality in male broilers and broiler breeders. Mortality in affected flocks may be as high as 15%. The disease has been called enterococcal spondylitis (ES), based on the frequent isolation of Enterococcus cecorum from the lesions and necrosis and inflammation observed in the free thoracic vertebrae (FTV) of affected birds. Male broiler breeders in an experimental setting were challenged with pure E. cecorum isolates obtained from ES-affected commercial flocks. Challenge routes included oral gavage (10(8)), intravenous (i.v.; 10(3)), and air sac (AS; 10(3)). Half the study birds in each group were chemically immunosuppressed with dexamethasone. Spinal lesions were observed grossly in birds challenged intravenously (2.9%) and birds challenged orally (6.1%). Microscopic spinal lesions consistent with ES were more frequently identified compared with gross lesions in the orally challenged group (30.3%). Chemical immunosuppression with dexamethasone was not associated with a greater incidence of ES in this study. By recreating the disease experimentally, the study design reported here may help in the further development of an experimental challenge model for future studies on risk factors, prevention, and therapeutic intervention of ES. PMID:21793445

  17. Palladium-catalyzed oxidative diffusion for tritium extraction from breeder-blanket fluids at low concentrations

    NASA Astrophysics Data System (ADS)

    Hsu, Cheazone; Buxbaum, Robert E.

    1986-11-01

    Oxidative diffusion can extract hydrogen from metal solutions at extremely low partial pressures. The hydrogen diffuses through a metal membrane and is oxidized to water. The oxidation reaction produces the very low downstream pressures that drive the flux. This method is attractive because the flux can be proportional to the square-root of upstream pressure. For fusion reactors with liquid lithium or lithium-lead alloy breeder blankets, permeation windows provide a simple, cheap tritium extraction method. Interdiffusion rates, separation flux, window size, helium contents, tritium holdup costs, and overall costs are calculated for membranes of palladium-coated zirconium, niobium, vanadium, nickel and stainless-steel. For extracting tritium from liquid lithium using the cheapest windows, Zr-Pd, the material and labor cost is 8.0 M at 1 wppm, and is inversely proportional to tritium concentration in the lithium. The tritium holdup cost for the windows is 4.8 M, and for the blanket it is proportional to the blanket volume and concentration. An overall economic optimization suggests that 1 to 1.5 wppm in lithium is optimal. For extracting tritium from 17Li83Pb at 0.26 wppb, the cheapest window is V-Pd; the cost is 2.6 M$, and the tritium holdup is negligible.

  18. Helium-cooled, FLiBe-breeder, beryllium-multiplier blanket for MINIMARS

    SciTech Connect

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

    1986-06-01

    We adapted the helium-cooled, FLiBe-breeder blanket to the commercial tandem-mirror fusion-reactor design, MINIMARS. Vanadium was used to achieve high performance from the high-energy-release neutron-capture reactions and from the high-temperature operation permitted by the refractory property of the material, which increases the conversion efficiency and decreases the helium-pumping power. Although this blanket had the highest performance among the MINIMARS blankets designs, measured by Mn/sub th/ (blanket energy multiplication times thermal conversion efficiency), it had a cost of electricity (COE) 18% higher than the University of Wisconsin (UW) blanket design (42.5 vs 35.9 mills/kW.h). This increased cost was due to using higher-cost blanket materials (beryllium and vanadium) and a thicker blanket, which resulted in higher-cost central-cell magnets and the need for more blanket materials. Apparently, the high efficiency does not substantially affect the COE. Therefore, in the future, we recommend lowering the helium temperature so that ferritic steel can be used. This will result in a lower-cost blanket, which may compensate for the lower performance resulting from lower efficiency.

  19. Deployment of a three-dimensional array of Micro-Pocket Fission Detector triads (MPFD3) for real-time, in-core neutron flux measurements in the Kansas State University TRIGA Mark-II Nuclear Reactor

    NASA Astrophysics Data System (ADS)

    Ohmes, Martin Francis

    A Micro-Pocket Fission Detector (MPFD) is a miniaturized type of fission chamber developed for use inside a nuclear reactor. Their unique design allows them to be located between or even inside fuel pins while being built from materials which give them an operational lifetime comparable to or exceeding the life of the fuel. While other types of neutron detectors have been made for use inside a nuclear reactor, the MPFD is the first neutron detector which can survive sustained use inside a nuclear reactor while providing a real-time measurement of the neutron flux. This dissertation covers the deployment of MPFDs as a large three-dimensional array inside the Kansas State University TRIGA Mark-II Nuclear Reactor for real-time neutron flux measurements. This entails advancements in the design, construction, and packaging of the Micro-Pocket Fission Detector Triads with incorporated Thermocouple, or MPFD3-T. Specialized electronics and software also had to be designed and built in order to make a functional system capable of collecting real-time data from up to 60 MPFD3-Ts, or 180 individual MPFDs and 60 thermocouples. Design of the electronics required the development of detailed simulations and analysis for determining the theoretical response of the detectors and determination of their size. The results of this research shows that MPFDs can operate for extended times inside a nuclear reactor and can be utilized toward the use as distributed neutron detector arrays for advanced reactor control systems and power mapping. These functions are critical for continued gains in efficiency of nuclear power reactors while also improving safety through relatively inexpensive redundancy.

  20. Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

    NASA Astrophysics Data System (ADS)

    Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

    2015-09-01

    The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  1. Educating the Next Generation of Plant Breeders: The Need and the Challenge

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Plant breeding is critical to the future of productive agriculture, food security, and economic prosperity. Increasingly, many plant breeders are working in industry or governmental agencies that do not include education of the next generation of plant breeders as part of their mission. At the sam...

  2. Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

    NASA Technical Reports Server (NTRS)

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

    1977-01-01

    Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

  3. Impacts of breeder loss on social structure, reproduction and population growth in a social canid.

    PubMed

    Borg, Bridget L; Brainerd, Scott M; Meier, Thomas J; Prugh, Laura R

    2015-01-01

    The importance of individuals to the dynamics of populations may depend on reproductive status, especially for species with complex social structure. Loss of reproductive individuals in socially complex species could disproportionately affect population dynamics by destabilizing social structure and reducing population growth. Alternatively, compensatory mechanisms such as rapid replacement of breeders may result in little disruption. The impact of breeder loss on the population dynamics of social species remains poorly understood. We evaluated the effect of breeder loss on social stability, recruitment and population growth of grey wolves (Canis lupus) in Denali National Park and Preserve, Alaska using a 26-year dataset of 387 radiocollared wolves. Harvest of breeding wolves is a highly contentious conservation and management issue worldwide, with unknown population-level consequences. Breeder loss preceded 77% of cases (n = 53) of pack dissolution from 1986 to 2012. Packs were more likely to dissolve if a female or both breeders were lost and pack size was small. Harvest of breeders increased the probability of pack dissolution, likely because the timing of harvest coincided with the breeding season of wolves. Rates of denning and successful recruitment were uniformly high for packs that did not experience breeder loss; however, packs that lost breeders exhibited lower denning and recruitment rates. Breeder mortality and pack dissolution had no significant effects on immediate or longer term population dynamics. Our results indicate the importance of breeding individuals is context dependent. The impact of breeder loss on social group persistence, reproduction and population growth may be greatest when average group sizes are small and mortality occurs during the breeding season. This study highlights the importance of reproductive individuals in maintaining group cohesion in social species, but at the population level socially complex species may be resilient

  4. Initiating the D&D Project for the EBR-II

    SciTech Connect

    Rick Demmer

    2010-08-01

    A novel decommissioning project is underway to close the Experimental Breeder Reactor-II (EBR-II) “fast” reactor at the Idaho National Laboratory (INL), Materials and Fuels Complex (MFC) facility near Idaho Falls, ID. The facility was placed in cold shutdown in 1994 and work began on the removal of the metallic sodium coolant. The bulk of the sodium was drained and treated beginning in 2001. The residual sodium heel was chemically passivated to render it less reactive in 2005 using a novel carbon dioxide treatment. Approximately 700 kg of metallic sodium and 3500 kg of sodium bicarbonate remain in the facility. A RCRA Waste Treatment Permit, issued in 2002 by the State of Idaho Department of Environmental Quality, requires annual progress toward closure of the facility, and that all regulated materials be removed or deactivated, and the waste products removed by 2022. The baseline sodium removal technology would result in about 100,000 gallons of low-level waste solution requiring treatment along with separate handling of the large components (intermediate heat exchanger, rotating plug, etc) outside of the primary tank.

  5. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    NASA Astrophysics Data System (ADS)

    Bahri, Che Nor Aniza Che Zainul; Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-01

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  6. Advantages of liquid fluoride thorium reactor in comparison with light water reactor

    SciTech Connect

    Bahri, Che Nor Aniza Che Zainul Majid, Amran Ab.; Al-Areqi, Wadeeah M.

    2015-04-29

    Liquid Fluoride Thorium Reactor (LFTR) is an innovative design for the thermal breeder reactor that has important potential benefits over the traditional reactor design. LFTR is fluoride based liquid fuel, that use the thorium dissolved in salt mixture of lithium fluoride and beryllium fluoride. Therefore, LFTR technology is fundamentally different from the solid fuel technology currently in use. Although the traditional nuclear reactor technology has been proven, it has perceptual problems with safety and nuclear waste products. The aim of this paper is to discuss the potential advantages of LFTR in three aspects such as safety, fuel efficiency and nuclear waste as an alternative energy generator in the future. Comparisons between LFTR and Light Water Reactor (LWR), on general principles of fuel cycle, resource availability, radiotoxicity and nuclear weapon proliferation shall be elaborated.

  7. Litter loss triggers estrus in a nonsocial seasonal breeder

    PubMed Central

    Steyaert, Sam MJG; Swenson, Jon E; Zedrosser, Andreas

    2014-01-01

    Sexually selected infanticide (SSI) is often presumed to be rare among seasonal breeders, because it would require a near immediate return to estrus after the loss of an entire litter during the mating season. We evaluated changes in reproductive strategies and the reproductive fate of females that experienced litter loss during the mating season in a seasonal breeder with strong evidence for SSI, the brown bear. First, we used a long-term demographic dataset (1986–2011) to document that a large majority of females (>91%) that lose their entire litter during the mating season in fact do enter estrus, mate, and give birth during the subsequent birthing season. Second, we used high-resolution movement data (2005–2011) to evaluate how females changed reproductive strategies after losing their entire litter during the mating season. We hypothesized that females would shift from the sedentary lifestyle typical for females with cubs-of-the-year to a roam-to-mate behavior typical for receptive females in no more than a few (∼3) days after litter loss. We found that females with cubs-of-the-year moved at about 1/3 of the rate and in a less bimodal diurnal pattern than receptive females during the mating season. The probability of litter loss was positively related with movement rate, suggesting that being elusive and sedentary is a strategy to enhance cub survival rather than a relic of cub mobility itself. The movement patterns of receptive females and females after litter loss were indistinguishable within 1–2 days after the litter loss, and we illustrate that SSI can significantly reduce the female interbirth interval (50–85%). Our results suggest that SSI can also be advantageous for males in seasonally breeding mammals. We propose that infanticide as a male reproductive strategy is more prevalent among mammals with reproductive seasonality than observed or reported. PMID:24558586

  8. Evaluation of the Use of Existing RELAP5-3D Models to Represent the Actinide Burner Test Reactor

    SciTech Connect

    C. B. Davis

    2007-02-01

    The RELAP5-3D code is being considered as a thermal-hydraulic system code to support the development of the sodium-cooled Actinide Burner Test Reactor as part of Global Nuclear Energy Partnership. An evaluation was performed to determine whether the control system could be used to simulate the effects of non-convective mechanisms of heat transport in the fluid that are not currently represented with internal code models, including axial and radial heat conduction in the fluid and subchannel mixing. The evaluation also determined the relative importance of axial and radial heat conduction and fluid mixing on peak cladding temperature for a wide range of steady conditions and during a representative loss-of-flow transient. The evaluation was performed using a RELAP5-3D model of a subassembly in the Experimental Breeder Reactor-II, which was used as a surrogate for the Actinide Burner Test Reactor. An evaluation was also performed to determine if the existing centrifugal pump model could be used to simulate the performance of electromagnetic pumps.

  9. Criticality and Safety Parameter Studies of a 3-MW TRIGA MARK-II Research Reactor and Validation of the Generated Cross-Section Library and Computational Method

    SciTech Connect

    Bhuiyan, S.I.; Mondal, M.A.W.; Sarker, M.M.; Rahman, M.; Shahdatullah, M.S.; Huda, M.Q.; Chakrobortty, T.K.; Khan, M.J.H

    2000-05-15

    This study deals with the analysis of some neutronics and safety parameters of the current core of a 3-MW TRIGA MARK-II research reactor and validation of the generated macroscopic cross-section library and calculational techniques by benchmarking with experimental, operational, and available Safety Analysis Report (SAR) values. The overall strategy is: (a) generation of the problem-dependent cross-section library from basic Evaluated Nuclear Data Files such as ENDF/B-VI and JENDL-3.2 with NJOY94.10+, (b) use of the WIMSD-5 package to generate a few-group neutron macroscopic cross section for all of the materials in the core and its immediate neighborhood, (c) use the three-dimensional CITATION code to perform the global analysis of the core, and (d) checking of the validity of the CITATION diffusion code with the MCNP4B2 Monte Carlo code. The ultimate objective is to establish methods for reshuffling the current core configuration to upgrade the thermal flux at irradiation locations for increased isotope production. The computational methods, tools and techniques, customization of cross-section libraries, various models for cells and supercells, and many associated utilities are standardized and established/validated for the overall neutronic analysis. The excess reactivity, neutron flux, power distribution, power peaking factors, determination of the hot spot, and fuel temperature reactivity coefficients {alpha}{sub f} in the temperature range of 45 to 1000 deg. C are studied. All the analyses are performed using the 4- and 7-group libraries of the macroscopic cross sections generated from the 69-group WIMSD-5 library. The 7-group calculations yield comparatively better agreement with the experimental value of k{sub eff} and the other core parameters. The CITATION test runs using different cross-section sets based on the different models applied in the WIMSD-5 calculations show a strong influence of those models on the final integral parameter. Some of the cells

  10. Updated reference design of a liquid metal cooled tandem mirror fusion breeder

    SciTech Connect

    Berwald, D.H.; Whitley, R.H.; Garner, J.K.; Gromada, R.J.; McCarville, T.J.; Moir, R.W.; Lee, J.D.; Bandini, B.R.; Fulton, F.J.; Wong, C.P.C.; Maya, I.; Hoot, C.G.; Schultz, K.R.; Miller, L.G.; Beeston, J.M.; Harris, B.L.; Westman, R.A.; Ghoniem, N.M.; Orient, G.; Wolfer, M.; DeVan, J.H.; Torterelli, P.

    1985-09-01

    Detailed studies of key techinical issues for liquid metal cooled fusion breeder (fusion-fission hybrid blankets) have been performed during the period 1983-4. Based upon the results of these studies, the 1982 reference liquid metal cooled tandem mirror fusion breeder blanket design was updated and is described. The updated reference blankets provides increased breeding and lower technological risk in comparison with the original reference blanket. In addition to the blanket design revisions, a plant concept, cost, and fuel cycle economics assessment is provided. The fusion breeder continues to promise an economical source of fissile fuel for the indefinite future.

  11. Effect of various sources of organic carbon and high nitrite and nitrate concentrations on the selection of denitrifying bacteria. II. Continuous cultures in packed bed reactors.

    PubMed

    Błaszczyk, M

    1983-01-01

    The effect of different organic compounds, nitrites and nitrates at the concentration of 1,000 mg N/l on the quantitative and strain-specific selection of denitrifying bacteria was determined in anaerobic packed bed reactors. Both the source of carbon and nitrogen form influenced strain specificity and the frequency of occurrence of denitrifying bacteria. The frequency of denitrifying bacteria within packed bed reactor ranged in different media from 11% (glucose and nitrates) to 100% (methanol and ethanol with nitrates). A single species selection was observed in the presence of nitrites within packed bed reactor: Pseudomonas aeruginosa in medium with acetate. Pseudomonas stutzeri in medium with ethanol, Pseudomonas mendocina in medium with methanol and Pseudomonas fluorescens in medium with glucose. When nitrates were present in packed bed reactor, the dominating bacteria were: P. stutzeri in medium with acetate, P. fluorescens in medium with ethanol, Paracoccus denitrificans in medium with methanol and Alcaligenes faecalis in medium with glucose. PMID:6194668

  12. MLW, TRU, LLW, MIXED, HAZARDOUS WASTES AND ENVIRONMENTAL RESTORATION. WASTE MANAGEMENT/ENERGY SECURITY AND A CLEAN ENVIRONMENT. DFR Decommissioning: the Breeder Fuel Processing

    SciTech Connect

    Bonnet, C.; Potier, P.; Ashton, Brian Morris

    2003-02-27

    The Dounreay site, in North Scotland, was opened in 1955 and a wide range of nuclear facilities have been built and operated there by UKAEA (The United Kingdom Atomic Energy Authority) for the development of atomic energy research. The Dounreay Fast Reactor (DFR) was built between 1955 and 1957, and operated until 1977 for demonstration purposes and for producing electricity. Today, its decommissioning is a key part of the whole Dounreay Site Restoration Plan that integrates the major decommissioning activities such as the fuel treatment and the waste management. The paper presents the contract strategy and provides an overview of the BFR project which consists in the removal of the breeder elements from the reactor and their further treatment. It mainly provides particular details of the Retrieval and Processing Facilities design.

  13. A comparison of radioactive waste from first generation fusion reactors and fast fission reactors with actinide recycling

    SciTech Connect

    Koch, M.; Kazimi, M.S.

    1991-04-01

    Limitations of the fission fuel resources will presumably mandate the replacement of thermal fission reactors by fast fission reactors that operate on a self-sufficient closed fuel cycle. This replacement might take place within the next one hundred years, so the direct competitors of fusion reactors will be fission reactors of the latter rather than the former type. Also, fast fission reactors, in contrast to thermal fission reactors, have the potential for transmuting long-lived actinides into short-lived fission products. The associated reduction of the long-term activation of radioactive waste due to actinides makes the comparison of radioactive waste from fast fission reactors to that from fusion reactors more rewarding than the comparison of radioactive waste from thermal fission reactors to that from fusion reactors. Radioactive waste from an experimental and a commercial fast fission reactor and an experimental and a commercial fusion reactor has been characterized. The fast fission reactors chosen for this study were the Experimental Breeder Reactor 2 and the Integral Fast Reactor. The fusion reactors chosen for this study were the International Thermonuclear Experimental Reactor and a Reduced Activation Ferrite Helium Tokamak. The comparison of radioactive waste parameters shows that radioactive waste from the experimental fast fission reactor may be less hazardous than that from the experimental fusion reactor. Inclusion of the actinides would reverse this conclusion only in the long-term. Radioactive waste from the commercial fusion reactor may always be less hazardous than that from the commercial fast fission reactor, irrespective of the inclusion or exclusion of the actinides. The fusion waste would even be far less hazardous, if advanced structural materials, like silicon carbide or vanadium alloy, were employed.

  14. Mechanical properties test data of Alloy 718 for liquid metal fast breeder reactor applications

    SciTech Connect

    Korth, G.E.

    1983-01-01

    Mechanical property test data are reported for Alloy 718 with two heat treatments: conventional heat treatment (CHT) for base metal and Idaho National Engineering Laboratory (INEL) heat treatment (IHT) for base and weld metal. Tests were conducted in air from 24 to 704{degree}C and include elastic properties (Young's modulus, shear modulus, Poisson's ratio), tensile properties, creep-rupture properties, fatigue properties, creep-fatigue properties, and Charpy impact behavior. Effects of long term thermal aging at 538, 593, 649, and 704{degree}C for times to 25,000 h are also reported for CHT material (tensile, creep-rupture, fatigue, and Charpy), and IHT material (tensile, and Charpy). 18 refs., 63 figs., 36 tabs.

  15. Thermal-performance study of liquid metal fast breeder reactor insulation

    SciTech Connect

    Shiu, Kelvin K.

    1980-09-01

    Three types of metallic thermal insulation were investigated analytically and experimentally: multilayer reflective plates, multilayer honeycomb composite, and multilayer screens. Each type was subjected to evacuated and nonevacuated conditions, where thermal measurements were made to determine thermal-physical characteristics. A variation of the separation distance between adjacent reflective plates of multilayer reflective plates and multilayer screen insulation was also experimentally studied to reveal its significance. One configuration of the multilayer screen insulation was further selected to be examined in sodium and sodium oxide environments. The emissivity of Type 304 stainless steel used in comprising the insulation was measured by employing infrared technology. A comprehensive model was developed to describe the different proposed types of thermal insulation. Various modes of heat transfer inherent in each type of insulation were addressed and their relative importance compared. Provision was also made in the model to allow accurate simulation of possible sodium and sodium oxide contamination of the insulation. The thermal-radiation contribution to heat transfer in the temperature range of interest for LMFBR's was found to be moderate, and the suppression of natural convection within the insulation was vital in preserving its insulating properties. Experimental data were compared with the model and other published results. Moreover, the three proposed test samples were assessed and compared under various conditions as viable LMFBR thermal insulations.

  16. AB INITIO STUDY OF ADVANCED METALLIC NUCLEAR FUELS FOR FAST BREEDER REACTORS

    SciTech Connect

    Landa, A; Soderlind, P; Grabowski, B; Turchi, P A; Ruban, A V; Vitos, L

    2012-04-23

    Density-functional formalism is applied to study the ground state properties of {gamma}-U-Zr and {gamma}-U-Mo solid solutions. Calculated heats of formation are compared with CALPHAD assessments. We discuss how the heat of formation in both alloys correlates with the charge transfer between the alloy components. The decomposition curves for {gamma}-based U-Zr and U-Mo solid solutions are derived from Ising-type Monte Carlo simulations. We explore the idea of stabilization of the {delta}-UZr{sub 2} compound against the {alpha}-Zr (hcp) structure due to increase of Zr d-band occupancy by the addition of U to Zr. We discuss how the specific behavior of the electronic density of states in the vicinity of the Fermi level promotes the stabilization of the U{sub 2}Mo compound. The mechanism of possible Am redistribution in the U-Zr and U-Mo fuels is also discussed.

  17. Seismic design technology for breeder reactor structures. Volume 1. Special topics in earthquake ground motion

    SciTech Connect

    Reddy, D.P.

    1983-04-01

    This report is divided into twelve chapters: seismic hazard analysis procedures, statistical and probabilistic considerations, vertical ground motion characteristics, vertical ground response spectrum shapes, effects of inclined rock strata on site response, correlation of ground response spectra with intensity, intensity attenuation relationships, peak ground acceleration in the very mean field, statistical analysis of response spectral amplitudes, contributions of body and surface waves, evaluation of ground motion characteristics, and design earthquake motions. (DLC)

  18. Contribution of Clinch River Breeder Reactor plant design and development to the LMFBR fuel cycle

    SciTech Connect

    Riley, D.R.; Dickson, P.W.

    1981-01-01

    This paper describes how the CRBRP development and CRBRP focus of the LMFBR base technology program have led to advances in the state of the art in physics, thermal-hydraulics, structural analysis, core restraint, seismic analysis, and analysis of hypothetical core-disruptive accident energetics, all of which have been incorporated through disciplined engineering into the final CRBRP design. The total development in the US of fuels and materials, the analytical advances made on CRBRP design, and the incorporation of the latest experimental results into that design have put the US technology in general and the CRBRP design in particular at the forefront of technology. This has placed the US in a position to develop the most favorable LMFBR fuel cycle.

  19. Tubular gage for a liquid-metal-cooled fast breeder reactor

    DOEpatents

    Hutter, Ernest; Tuma, Leroy A.

    1977-06-14

    Spring-loaded plungers are arranged about a housing for insertion into a polygonal tube, one plunger for each side of the tube. Each plunger has a locking cam and sliding wedge mechanism which can overcome the spring force associated with the plunger and lock it in any position. The wedges are operated by a rod moveable axially in the housing. Several housings with their associated plungers can be stacked. The stack is lowered into the polygonal tube with all of the plungers locked in a fully inward position. When the stack is in the tube, each wedge is moved to release its locking cam, allowing each of the plungers to spring outward against an inner side of the tube. Each housing will thus gage the internal dimensions of the tube at its elevation. The plungers are locked in position, the entire stack is rotated to bring the plungers into the corners described by the intersections of the flat sides, and the stack is removed from the tube whereupon the dimensions across opposite locked plungers may be read by a micrometer.

  20. US Liquid Metal Fast Breeder Reactor man-machine interface program

    SciTech Connect

    Vaurio, J.K.; Change, S.A.

    1982-01-01

    The US LMFBR Man-Machine Interface Program is supportive to and an integral part of the LMFBR Safety Program. This paper describes the goal and objectives of the program, and the necessary research and development efforts with a logical structure for the orderly and timely implementation of the prgoram. Current status and near-term and long-term priority activities are also summarized.

  1. Piping support system for liquid-metal fast-breeder reactor

    DOEpatents

    Brussalis, Jr., William G.

    1984-01-01

    A pipe support consisting of a rigid link pivotally attached to a pipe and an anchor, adapted to generate stress or strain in the link and pipe due to pipe thermal movement, which stress or strain can oppose further pipe movement and generally provides pipe support. The pipe support can be used in multiple combinations with other pipe supports to form a support system. This support system is most useful in applications in which the pipe is normally operated at a constant elevated or depressed temperature such that desired stress or strain can be planned in advance of pipe and support installation. The support system is therefore especially useful in steam stations and in refrigeration equipment.

  2. Development of the DIPRES process for the fast breeder reactor fuel cycle

    SciTech Connect

    Collins, E D; Jackson, M D; Griffin, C W; Rasmussen, D E; Norman, R E

    1984-01-01

    In 1979 the Consolidated Fuel Reprocessing Program (CFRP) at ORNL initiated a program for the development of advanced conversion processes with potential for simplifying and improving the conversion/pellet fabrication flowsheet for recycle plutonium. An evaluation of advanced conversion processes led to the selection of DIPRES (DIrect PREss Spheriodized) for development because it has the largest potential for process and product improvements. DIPRES utilizes a gel sphere conversion process and product to provide a spherical feed material for pellet fabrication. The free-flowing nature of the spherical conversion product allows it to be fed directly to pellet presses (i.e., direct press feed) in place of conventional, mechanically blended powder feed. This is advantageous for remote fabrication. The DIPRES feed is prepared by an internal gelation process.

  3. Seismic design technology for breeder reactor structures. Volume 4. Special topics in piping and equipment

    SciTech Connect

    Reddy, D.P.

    1983-04-01

    This volume is divided into five chapters: experimental verification of piping systems, analytical verification of piping restraint systems, seismic analysis techniques for piping systems with multisupport input, development of floor spectra from input response spectra, and seismic analysis procedures for in-core components. (DLC)

  4. Applications of plasma core reactors to terrestrial energy systems

    NASA Technical Reports Server (NTRS)

    Latham, T. S.; Biancardi, F. R.; Rodgers, R. J.

    1974-01-01

    Plasma core reactors offer several new options for future energy needs in addition to space power and propulsion applications. Power extraction from plasma core reactors with gaseous nuclear fuel allows operation at temperatures higher than conventional reactors. Highly efficient thermodynamic cycles and applications employing direct coupling of radiant energy are possible. Conceptual configurations of plasma core reactors for terrestrial applications are described. Closed-cycle gas turbines, MHD systems, photo- and thermo-chemical hydrogen production processes, and laser systems using plasma core reactors as prime energy sources are considered. Cycle efficiencies in the range of 50 to 65 percent are calculated for closed-cycle gas turbine and MHD electrical generators. Reactor advantages include continuous fuel reprocessing which limits inventory of radioactive by-products and thorium-U-233 breeder configurations with about 5-year doubling times.-

  5. Results with the electron cyclotron resonance charge breeder for the {sup 252}Cf fission source project (Californium Rare Ion Breeder Upgrade) at Argonne Tandem Linac Accelerator System

    SciTech Connect

    Vondrasek, R.; Kondrashev, S.; Pardo, R.; Scott, R.; Zinkann, G. P.

    2010-02-15

    The construction of the Californium Rare Ion Breeder Upgrade, a new radioactive beam facility for the Argonne Tandem Linac Accelerator System (ATLAS), is nearing completion. The facility will use fission fragments from a 1 Ci {sup 252}Cf source; thermalized and collected into a low-energy particle beam by a helium gas catcher. In order to reaccelerate these beams, an existing ATLAS electron cyclotron resonance (ECR) ion source was redesigned to function as an ECR charge breeder. Thus far, the charge breeder has been tested with stable beams of rubidium and cesium achieving charge breeding efficiencies of 9.7% into {sup 85}Rb{sup 17+} and 2.9% into {sup 133}Cs{sup 20+}.

  6. Nucleon-meson transport capability for accelerator-breeder target design. [CALOR

    SciTech Connect

    Gabriel, T.A.; Alsmiller, R.G. Jr.

    1982-01-01

    A state-of-the-art code system for nucleon-meson-lepton transport which has direct applicability to accelerator breeders is presented. Some pertinent data that have been obtained using this system are discussed and compared with experimental data.

  7. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    SciTech Connect

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

  8. Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

    SciTech Connect

    Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.; Johnson, Gerald D.; Brown, W. F.; Paxton, Michael M.; Puigh, Raymond J.; Eiholzer, Cheryl R.; Martinez, C.; Blotter, M. A.

    2000-02-28

    A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

  9. The prevalence of subclinical endometritis and intrauterine infections in repeat breeder cows.

    PubMed

    Pothmann, H; Prunner, I; Wagener, K; Jaureguiberry, M; de la Sota, R L; Erber, R; Aurich, C; Ehling-Schulz, M; Drillich, M

    2015-05-01

    The objectives of this study were to assess the prevalence of subclinical endometritis and the presence of common uterine pathogens in repeat breeder cows. A total of 121 cows with three or more consecutive artificial inseminations without conception and no clinical signs of disease were defined as repeat breeder cows and were enrolled in this trial. Intrauterine samples were collected with the cytobrush technique to determine the prevalence of subclinical endometritis and bacteriologic infections. Blood samples were analyzed for concentrations of progesterone and estradiol in plasma to assess ovarian activity. Furthermore, breed, parity, history of calving and postpartum uterine infection, clinical findings of transrectal palpation, and backfat thickness were analyzed as potential factors for the prevalence of subclinical endometritis in repeat breeder cows. The prevalence of subclinical endometritis in repeat breeder cows was 12.7%; but common uterine pathogens, Escherichia coli and Trueperella pyogenes, were found in only one and three cows, respectively. Ovarian activity was determined in 95.0% of all cows. Recorded variables had no effect on the prevalence of subclinical endometritis in repeat breeder cows. In conclusion, subclinical endometritis and uterine infections linked to common pathogens were playing a minor role as a cause for repeat breeder cows in this study. Alternative reasons for failure to conceive in these cows are discussed. PMID:25670153

  10. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    SciTech Connect

    1984-04-01

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  11. CORAL: a stepping stone for establishing the Indian fast reactor fuel reprocessing technology

    SciTech Connect

    Venkataraman, M.; Natarajan, R.; Raj, Baldev

    2007-07-01

    The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR) spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)

  12. Thermal response of a pin-type fusion reactor blanket during steady and transient reactor operation

    SciTech Connect

    Grotz, S.; Ghoniem, N.M.

    1986-02-01

    The thermal analysis of the blanket examines both the steady-state and transient reactor operations. The steady-state analysis covers full power and fractional power operation whereas the transient analysis examines the effects of power ramps and blanket preheat. The blanket configuration chosen for this study is a helium cooled solid breeder design. We first discuss the full power, steady-state temperature fields in the first wall, beryllium rods, and breeder rods. Next we examine the effects of fractional power on coolant flow and temperature field distributions. This includes power plateaus of 10%, 20%, 50%, 80%, and 100% of full power. Also examined are the restrictions on the rates of power ramping between plateaus. Finally we discuss the power and time requirements for pre-heating the primary from cold iron conditions up to startup temperature (250/sup 0/C).

  13. Nonlinear, inelastic fast reactor subassembly interaction analyses

    SciTech Connect

    Sutherland, W.H.; Bard, F.E.

    1983-01-01

    Liquid Metal Fast Breeder Reactor (LMFBR) core structural design is complicated by the trade-offs associated with keeping the subassemblies closely packed for the neutronic considerations and accommodating the volumetric changes associated with irradiation swelling. The environmental variation across the reactor core results in temperature and neutron flux gradients across the subassemblies which in turn cause the subassemblies to bow as well as dilate and grow volumetrically. These deformations in a tightly packed reactor core cause the subassemblies to interact and can potentially result in excessive withdrawal loads during the refueling operations. ABADAN, a general purpose, nonlinear, inelastic, multi-dimensional finite element structural analysis computer code, was developed for the express purpose of solving large nonlinear problems as typified by the above interaction problems. For the subassembly interaction problem ABADAN has been applied to the solution of an interacting radial row of Fast Flux Test Facility (FFTF) fuel assemblies.

  14. Reactor control rod timing system. [LMFBR

    DOEpatents

    Wu, P.T.K.

    1980-03-18

    A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  15. Application of PCT to the EBR II ceramic waste form.

    SciTech Connect

    Ebert, W. L.; Lewis, M. A.; Johnson, S. G.

    2002-01-10

    We are evaluating the use of the Product Consistency Test (PCT) developed to monitor the consistency of borosilicate glass waste forms for application to the multiphase ceramic waste form (CWF) that will be used to immobilize waste salts generated during the electrometallurgical conditioning of spent sodium-bonded nuclear fuel from the Experimental Breeder Reactor No. 2 (EBR II). The CWF is a multiphase waste form comprised of about 70% sodalite, 25% borosilicate glass binder, and small amounts of halite and oxide inclusions. It must be qualified for disposal as a non-standard high-level waste (HLW) form. One of the requirements in the DOE Waste Acceptance System Requirements Document (WASRD) for HLW waste forms is that the consistency of the waste forms be monitored.[1] Use of the PCT is being considered for the CWF because of the similarities of the dissolution behaviors of both the sodalite and glass binder phases in the CWF to borosilicate HLW glasses. This paper provides (1) a summary of the approach taken in selecting a consistency test for CWF production and (2) results of tests conducted to measure the precision and sensitivity of the PCT conducted with simulated CWF.

  16. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    SciTech Connect

    Corradin, Michael; Anderson, M.; Muci, M.; Hassan, Yassin; Dominguez, A.; Tokuhiro, Akira; Hamman, K.

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  17. Microwave cavity studies for thermal testing of ceramic breeder materials

    SciTech Connect

    Kuston, R.L.

    1987-01-01

    Dielectric heating of proposed ceramic tritium breeder material to study the thermomechanical and thermal-hydraulic properties of the material has been previously suggested. Recent computer studies using codes capable of modeling three-dimensional EM cavities with enclosed dielectric material have been used to determine the size limitations of cavity designs at 200 MHz. The sample can be as large as 0.44 /times/ 0.72m in the plane that is transverse to the direction of neutron flux. The uniformity of volumetric heating over the transverse plane is constant to within a few percent. The sample can be as long as 10cm in the direction of the heat flux and match the expected exponential decay of heat generation, exp( /minus/z/lambda), to within +/minus/8%. The design of the chamber is decribed, including the sample region, additional dielectric loading blocks on two sides of the sample region that are required to generate the field uniformity in the transverse plane, and a description of the matching-section portion of the cavity which provides the correct geometry to cause the cavity to resonate at 200 MHz with the right z dependence to stimulate the exponentially-decaying heat profile in the sample region. The matching section consists of two dielectric slabs, one on each wall of the chamber, and an air or free space region in the center of the matching section. The coupling loop is located near the wall end of the matching section in the free space region. 7 refs., 2 figs.

  18. Influence of canthaxanthin on broiler breeder reproduction, chick quality, and performance.

    PubMed

    Zhang, W; Zhang, K Y; Ding, X M; Bai, S P; Hernandez, J M; Yao, B; Zhu, Q

    2011-07-01

    To investigate the effect of canthaxanthin supplied via a maternal route on the production of both breeder hens and chickens, 270 Chinese Three-Yellow breeder hens were randomly divided into 2 groups consisting of 135 birds each (5 replicates of 27) for study. The breeder hens were fed either a basal diet or the basal diet supplemented with 6 mg of canthaxanthin/kg for 24 wk. At the end of the 24-wk breeder experiment, all hatching eggs laid in 5 consecutive days of each group were collected and incubated. For each breeder group, 100 newly hatched chicks (5 replicates of 20) were reared under environmentally controlled conditions for 21 d. Canthaxanthin supplementation resulted in the following outcomes: an enhancement of the serum total antioxidant capacity (TAC) of breeder hens (P = 0.029), a significant increase in the yolk colorimetric score of Roche Yolk Color Fan (RYCF; P < 0.001), and a significant improvement of the antioxidant status of the egg yolk (P < 0.05). The chicks that hatched from eggs laid by breeder hens fed the canthaxanthin supplementation diet demonstrated a higher pigmentation colorimetric score of RYCF for their shank skin (P < 0.05), and the antioxidant capacity of the newly hatched chicks was significantly increased (P < 0.05). Both of these positive effects on shank skin pigmentation colorimetric score of RYCF and antioxidant capacity were observed for at least 7 d posthatching, and the chicks that hatched from canthaxanthin-enriched eggs showed a lower mortality (0 vs. 4%) during the first 21 d posthatching. These findings support the hypothesis that canthaxanthin supplementation of the maternal diet enhances the protective capacity of tissues against oxidative stress in vivo, which might be beneficial for poultry producers. PMID:21673167

  19. BEATRIX-II, phase II: Data summary report

    SciTech Connect

    Slagle, O.D.; Hollenberg, G.W.

    1996-05-01

    The BEATRIX-II experimental program was an International Energy Agency sponsored collaborative effort between Japan, Canada, and the United States to evaluate the performance of ceramic solid breeder materials in a fast-neutron environment at high burnup levels. This report addresses the Phase II activities, which included two in situ tritium-recovery canisters: temperature-change and temperature-gradient. The temperature-change canister contained a Li{sub 2}O ring specimen that had a nearly uniform temperature profile and was capable of temperature changes between 530 and 640{degrees}C. The temperature-gradient canister contained a Li{sub 2}ZrO{sub 3} pebble bed operating under a thermal gradient of 440 to 1100{degrees}C. Postirradiation examination was carried out to characterize the Phase II in situ specimens and a series of nonvented capsules designed to address the compatibility of beryllium with lithium-ceramic solid-breeder materials. The results of the BEATRIX-II, Phase II, irradiation experiment provided an extensive data base on the in situ tritium-release characteristics of Li{sub 2}O and Li{sub 2}ZrO{sub 3} for lithium burnups near 5%. The composition of the sweep gas was found to be a critical parameter in the recovery of tritium from both Li{sub 2}O and Li{sub 2}ZrO{sub 3}. Tritium inventories measured confirmed that Li{sub 2}O and Li{sub 2}ZrO{sub 3} exhibited very low tritium retention during the Phase II irradiation. Tritium inventories in Li{sub 2}ZrO{sub 3} after Phase II tended to be larger than those found for Li{sub 2}ZrO{sub 3} in other in situ experiments, but the larger values may reflect the larger generation rates in BEATRIX-II. A series of 20 capsules was irradiated to determine the compatibility of lithium ceramics and beryllium under conditions similar to a fusion blanket. It is concluded that Li{sub 2}O and Li{sub 2}ZrO{sub 3} should remain leading candidates for use in a solid-breeder fusion-blanket application.

  20. Effects of breeder age on mineral contents and weight of yolk sac, embryo development, and hatchability in Pekin ducks.

    PubMed

    Onbasilar, E E; Erdem, E; Hacan, O; Yalçin, S

    2014-02-01

    The current study was carried out to investigate the effects of breeder age on egg composition, changes of embryo, yolk sac, and yolk minerals during incubation and hatchability in Pekin ducks. A total of 495 freshly laid eggs were obtained from the same flock of Pekin ducks, aged 28, 34, and 40 wk, and were reared in accordance with the management guide of the duck breeders (Star 53-Grimaud Freres). At each breeder age, egg measurements were made on a random subsample of unincubated eggs. Embryo and yolk sac measurements were made on embryonic day (E) 12, E16, E20, and E25. On d 28 of incubation, the healthy ducklings were removed and sex of chicks was determined. All chicks were weighed and hatching results were determined. Egg weight and yolk percentages increased; however, albumen percentages, shell thickness, and yolk index decreased as the flock aged. Shell percentages, shell breaking strength, albumen index, and haugh units were not affected by breeder age. Also, breeder age affected the Mg, P, K, Ca, Cu, and Zn levels in the yolk, except for Na level on day of setting, and breeder age affected the mineral consumed by embryo during incubation. However, on E25, the levels of examined minerals, except for P level in the yolk sac, were not statistically different in duck breeder age groups. Relative yolk sac and embryo weights of eggs obtained from different breeder ages varied from E16 to E25; however, embryo length was different in breeder age groups from E12 to E20. Hatching weight was affected by breeder age and sex. Hatching results were not different among breeder age groups. This study indicates that breeder age is important for some egg characteristics, relative yolk sac weight, some contents of minerals in the yolk, embryonic growth during incubation, and duckling weight. PMID:24570471

  1. BEATRIX-II: In situ tritium recovery from a fast neutron irradiation of solid breeder materials

    SciTech Connect

    Puigh, R J; Hollenberg, G W; Kurasawa, T; Watanabe, H; Hastings, I J; Miller, J M; Berk, S E; Bauer, R E; Baker, D E

    1988-09-01

    An in situ tritium recovery experiment is being fabricated for the irradiation of Li/sub 2/O in the Fast Flux Test Facility located at the Hanford Site, Richland, Washington, United States of America. Two in situ tritium recovery canisters will be irradiated with lithium atom burnups to 4%. One canister will provide fundamental data on tritium release as a function of temperature, gas composition, and flow rate. The other canister will provide integrated performance data from solid pellet specimens with large (450/degree/C) radial temperature gradients. 10 refs., 5 figs., 1 tab.

  2. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  3. Development of tritium breeding blankets for DT-burning fusion reactors

    SciTech Connect

    Clemmer, R.G.

    1980-01-01

    This study examines the status of understanding of blanket tritium recovery and the performance of potentially viable tritium breeding materials under conditions anticipated in a DT-fueled fusion reactor environment. The existing physicochemical, thermophysical, and ceramographic data for candidate liquid and solid breeders are reviewed and appropriate operating conditions defined. It is shown that selection of a breeding material and an appropriate tritium recovery method can impose significant constraints upon blanket design, particularly when considerations of breeder/coolant/structure compatibility and temperature limitations are taken into account.

  4. Research Reactor Benchmarks

    SciTech Connect

    Ravnik, Matjaz; Jeraj, Robert

    2003-09-15

    A criticality benchmark experiment performed at the Jozef Stefan Institute TRIGA Mark II research reactor is described. This experiment and its evaluation are given as examples of benchmark experiments at research reactors. For this reason the differences and possible problems compared to other benchmark experiments are particularly emphasized. General guidelines for performing criticality benchmarks in research reactors are given. The criticality benchmark experiment was performed in a normal operating reactor core using commercially available fresh 20% enriched fuel elements containing 12 wt% uranium in uranium-zirconium hydride fuel material. Experimental conditions to minimize experimental errors and to enhance computer modeling accuracy are described. Uncertainties in multiplication factor due to fuel composition and geometry data are analyzed by sensitivity analysis. The simplifications in the benchmark model compared to the actual geometry are evaluated. Sample benchmark calculations with the MCNP and KENO Monte Carlo codes are given.

  5. Light Water Breeder end-of-life component examinations at Shippingport Atomic Power Station and module visual and dimensional examinations at Expended Core Facility (LWBR Development Program)

    SciTech Connect

    Wargo, J.E.

    1987-10-01

    This report presents highlights of visual and dimensional examinations of the Light Water Breeder Reactor fuel assemblies and selected core components following five years of power operation in which the core achieved 29,047 effective full power hours. Each type of fuel assembly (seed, blanket, and reflector) is described, and the end-of-life conditions are documented in photographs and data plots. Fuel modules were examined immediately after removal from the reactor vessel at the Shippingport Atomic Power Station and after shipment to the Expended Core Facility at the Naval Reactors Facility in Idaho. Further inspection was performed on one seed and one reflector assembly after their external support shells were removed. Module length changes and bow data are presented for selected assemblies. Structural component examinations include magnetic particle testing and ultrasonic test inspection of the LWBR reactor vessel closure head. Visual inspections were also performed on compression sleeves and guide tube extensions which formed part of the guide path for the movable fuel assemblies. 4 refs., 103 figs., 5 tabs.

  6. Liquid-metal-cooled reactor

    DOEpatents

    Hutter, E.

    A perforated depressor plate extending across the bottom of the instrument tree of a fast breeder reactor cooperates with a circular cylindrical metal bellows forming a part of the upper adapter of each core assembly and bearing on the bottom of the depressor plate to restrict flow of coolant between core assemblies, thereby reducing significantly the pressure differential between the coolant inside the core assemblies and the coolant outside of the core assemblies. Openings in the depressor plate are slightly smaller than the top of the upper adapter so the depressor plate will serve as a backup mechanical holddown for the core. In addition, coolant mixing devices and locating devices are provided attached to the depressor plate.

  7. Fatigue Testing of Metallurgically-Bonded EBR-II Superheater Tubes

    SciTech Connect

    Terry C. Totemeier

    2006-12-01

    Fatigue crack growth tests were performed on 2¼Cr-1Mo steel specimens machined from ex-service Experimental Breeder ReactorII (EBR-II) superheater duplex tubes. The tubes had been metallurgically bonded with a 100 µm thick Ni interlayer; the specimens incorporated this bond layer. Tests were performed at room temperature in air and at 400°C in air and humid Ar; cracks were grown at varied levels of constant ?K. Crack growth tests at a range of ?K were also performed on specimens machined from the shell of the superheater. In all conditions the presence of the Ni interlayer was found to result in a net retardation of growth as the crack passed through the interlayer. The mechanism of retardation was identified as a disruption of crack planarity and uniformity after passing through the porous interlayer. Full crack arrest was only observed in a single test performed at near-threshold ?K level (12 MPa?m) at 400°C. In this case the crack tip was blunted by oxidation of the base steel at the steel-interlayer interface.

  8. Visual imagery and the user model applied to fuel handling at EBR-II

    SciTech Connect

    Brown-VanHoozer, S.A.

    1995-06-01

    The material presented in this paper is based on two studies involving visual display designs and the user`s perspective model of a system. The studies involved a methodology known as Neuro-Linguistic Programming (NLP), and its use in expanding design choices which included the ``comfort parameters`` and ``perspective reality`` of the user`s model of the world. In developing visual displays for the EBR-II fuel handling system, the focus would be to incorporate the comfort parameters that overlap from each of the representation systems: visual, auditory and kinesthetic then incorporate the comfort parameters of the most prominent group of the population, and last, blend in the other two representational system comfort parameters. The focus of this informal study was to use the techniques of meta-modeling and synesthesia to develop a virtual environment that closely resembled the operator`s perspective of the fuel handling system of Argonne`s Experimental Breeder Reactor - II. An informal study was conducted using NLP as the behavioral model in a v reality (VR) setting.

  9. Fatigue Testing of Metallurgically-Bonded EBR-II Superheater Tubes

    SciTech Connect

    T.C. Totemeier; D.M. Wachs; D.L. Porter

    2008-05-01

    Fatigue crack growth and impact tests were performed on 2¼Cr-1Mo steel specimens machined from ex-service Experimental Breeder ReactorII (EBR-II) superheater duplex tubes. The tubes had been metallurgically bonded with a 100 µm thick Ni layer; the specimens incorporated this bond layer. Impact tests were performed at temperatures from –50 to 400°C; cracks propagating from the V-notch were arrested by delamination at the bond layer for all tests with one exception at –50°C. Fatigue crack growth tests were performed at room temperature in air and at 400°C in air and humid Ar; cracks were grown at varied levels of constant ?K. In all conditions the presence of the Ni bond layer was found to result in a net retardation of growth as the crack passed through the layer. The mechanism of retardation was identified as a disruption of crack planarity and uniformity after passing through the porous bond layer. Full crack arrest was only observed in a single test performed at near-threshold ?K level (12 MPa?m) at 400°C. In this case the crack tip was blunted by oxidation of the base steel at the steel-nickel interface.

  10. Development of fusion blanket technology for the DEMO reactor.

    PubMed

    Colling, B R; Monk, S D

    2012-07-01

    The viability of various materials and blanket designs for use in nuclear fusion reactors can be tested using computer simulations and as parts of the test blanket modules within the International Thermonuclear Experimental Reactor (ITER) facility. The work presented here focuses on blanket model simulations using the Monte Carlo simulation package MCNPX (Computational Physics Division Los Alamos National Laboratory, 2010) and FISPACT (Forrest, 2007) to evaluate the tritium breeding capability of a number of solid and liquid breeding materials. The liquid/molten salt breeders are found to have the higher tritium breeding ratio (TBR) and are to be considered for further analysis of the self sufficiency timing. PMID:22112596

  11. Startup of the FFTF sodium cooled reactor. [Acceptance Test Program

    SciTech Connect

    Redekopp, R.D.; Umek, A.M.

    1981-03-01

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed.

  12. Nuclear reactor for breeding U.sup.233

    DOEpatents

    Bohanan, Charles S.; Jones, David H.; Raab, Jr., Harry F.; Radkowsky, Alvin

    1976-01-01

    A light-water-cooled nuclear reactor capable of breeding U.sup.233 for use in a light-water breeder reactor includes physically separated regions containing U.sup.235 fissile material and U.sup.238 fertile material and Th.sup.232 fertile material and Pu.sup.239 fissile material, if available. Preferably the U.sup.235 fissile material and U.sup.238 fertile material are contained in longitudinally movable seed regions and the Pu.sup.239 fissile material and Th.sup.232 fertile material are contained in blanket regions surrounding the seed regions.

  13. Behaviour of palladium(II), platinum(IV), and rhodium(III) in artificial and natural waters: influence of reactor surface and geochemistry on metal recovery.

    PubMed

    Cobelo-Garcia, Antonio; Turner, Andrew; Millward, Geoffrey E; Couceiro, Fay

    2007-03-01

    The recovery of dissolved platinum group elements (PGE: Pd(II), Pt(IV) and Rh(III)) added to Milli-Q water, artificial freshwater and seawater and filtered natural waters has been studied, as a function of pH and PGE concentration, in containers of varying synthetic composition. The least adsorptive and/or precipitative loss was obtained for borosilicate glass under most of the conditions employed, whereas the greatest loss was obtained for low-density polyethylene. Of the polymeric materials tested, the adsorptive and/or precipitative loss of PGE was lowest for fluorinated ethylene propylene (Teflon). The loss of Pd(II) in freshwater was significant due to its affinity for surface adsorption and its relatively low solubility. The presence of natural dissolved organic matter increases the recovery of Pd(II) but enhances the loss of Pt(IV). The loss of Rh(III) in seawater was significant and was mainly due to precipitation, whereas Pd(II) recovery was enhanced, compared to freshwater, because of its complexation with chloride. The results have important implications regarding protocols employed for sample preservation and controlled laboratory experiments used in the study of the speciation and biogeochemical behaviour of PGE. PMID:17386666

  14. Optimization of electron-cyclotron-resonance charge-breeder ions : Final CRADA Report.

    SciTech Connect

    Pardo, R.; Physics; Far-Tech, Inc.

    2009-10-09

    Measurements of 1+ beam properties and associated performance of ECR Charge Breeder source determined by total efficiency measurement and charge state distributions from the ECR Charge Breeder. These results were communicated to Far-Tech personnel who used them to benchmark the newly developed programs that model ion capture and charge breeding in the ECR Charge Breeder Source. Providing the basic data described above and in the discussion below to Far-Tech allowed them to improve and refine their calculational tools for ECR ion sources. These new tools will be offered for sale to industry and will also provide important guidance to other research labs developing Charge Breeding ion sources for radioactive beam physics research.

  15. Full-length U-xPu-10Zr (x = 0, 8, 19 wt.%) fast reactor fuel test in FFTF

    NASA Astrophysics Data System (ADS)

    Porter, D. L.; Tsai, Hanchung

    2012-08-01

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt.%) metallic fast reactor test with commercial-length (91.4-cm active fuel-column length) conducted to date. With few remaining test reactors, there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning-of-life (BOL) peak cladding temperature of the hottest pin was 608 °C, cooling to 522 °C at end-of-life (EOL). Selected fuel pins were examined non-destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta-gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3-cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ˜0.7 X/L axial location along the fuel column. This resulted from a higher production of rare-earth fission products at this location and a higher ΔT between fuel center and cladding than at core center, together providing more rare earths at the cladding and more FCCI. This behavior could

  16. Full-length U-xPu-10Zr (x=0, 8, 19 wt%) Fast Reactor Fuel Test in FFTF

    SciTech Connect

    D. L. Porter; H.C. Tsai

    2012-08-01

    The Integral Fast Reactor-1 (IFR-1) experiment performed in the Fast Flux Test Facility (FFTF) was the only U-Pu-10Zr (Pu-0, 8 and 19 wt%) metallic fast reactor test with commercial-length (91.4 cm active fuel column length) conducted to date. With few remaining test reactors there is little opportunity for performing another test with a long active fuel column. The assembly was irradiated to the goal burnup of 10 at.%. The beginning of life (BOL) peak cladding temperature of the hottest pin was 608?C, cooling to 522?C at end of life (EOL). Selected fuel pins were examined non destructively using neutron radiography, precision axial gamma scanning, and both laser and spiral contact cladding profilometry. Destructive exams included plenum gas pressure, volume, and gas composition determinations on a number of pins followed by optical metallography, electron probe microanalysis (EPMA), and alpha and beta gamma autoradiography on a single U-19Pu-10Zr pin. The post-irradiation examinations (PIEs) showed very few differences compared to the short-pin (34.3 cm fuel column) testing performed on fuels of similar composition in Experimental Breeder Reactor-II (EBR-II). The fuel column grew axially slightly less than observed in the short pins, but with the same pattern of decreasing growth with increasing Pu content. There was a difference in the fuel-cladding chemical interaction (FCCI) in that the maximum cladding penetration by interdiffusion with fuel/fission products did not occur at the top of the fuel column where the cladding temperature is highest, as observed in EBR-II tests. Instead, the more exaggerated fission-rate profile of the FFTF pins resulted in a peak FCCI at ~0.7 X/L axial location along the fuel column. This resulted from a lower production of rare earth fission products higher in the fuel column as well as a much smaller delta-T between fuel center and cladding, and therefore less FCCI, despite the higher cladding temperature. This behavior could

  17. 63. ARAII. Tenton crane in SL1 reactor building transports the ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    63. ARA-II. Ten-ton crane in SL-1 reactor building transports the reactor head. February 24, 1958. Ineel photo no. 58-879. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  18. BNCT for locally recurrent head and neck cancer: preliminary clinical experience from a phase I/II trial at Tsing Hua Open-Pool Reactor.

    PubMed

    Wang, L W; Wang, S J; Chu, P Y; Ho, C Y; Jiang, S H; Liu, Y W H; Liu, Y H; Liu, H M; Peir, J J; Chou, F I; Yen, S H; Lee, Y L; Chang, C W; Liu, C S; Chen, Y W; Ono, K

    2011-12-01

    To introduce our preliminary experience of treating locally and regionally recurrent Head and Neck cancer patients at Tsing Hua Open-Pool Reactor in Taiwan, four patients (M/F=3/1, median age 68 Y/O) were enrolled. BNCT with BPA (400 mg/kg) injected in 2 phases and prescription dose of 12-35 Gy (Eq.)/fraction for 2 fractions at 30 day interval can be given with sustained blood boron concentration and tolerable early toxicities for recurrent H & N cancer. PMID:21478023

  19. Fractionated BNCT for locally recurrent head and neck cancer: experience from a phase I/II clinical trial at Tsing Hua Open-Pool Reactor.

    PubMed

    Wang, Ling-Wei; Chen, Yi-Wei; Ho, Ching-Yin; Hsueh Liu, Yen-Wan; Chou, Fong-In; Liu, Yuan-Hao; Liu, Hong-Ming; Peir, Jinn-Jer; Jiang, Shiang-Huei; Chang, Chi-Wei; Liu, Ching-Sheng; Wang, Shyh-Jen; Chu, Pen-Yuan; Yen, Sang-Hue

    2014-06-01

    To introduce our experience of treating locally and regionally recurrent head and neck cancer patients with BNCT at Tsing Hua Open-Pool Reactor in Taiwan, 12 patients (M/F=10/2, median age 55.5 Y/O) were enrolled and 11 received two fractions of treatment. Fractionated BNCT at 30-day interval with adaptive planning according to changed T/N ratios was feasible, effective and safe for selected recurrent head and neck cancer in this trial. PMID:24369888

  20. 64. ARAII. Interior view of SL1 reactor building with reactor ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    64. ARA-II. Interior view of SL-1 reactor building with reactor head in place in center foreground. March 21, 1958. Ineel photo no. 58-1360. Photographer: Jack L. Anderson. - Idaho National Engineering Laboratory, Army Reactors Experimental Area, Scoville, Butte County, ID

  1. Performance of the Argonne National Laboratory electron cyclotron resonance charge breeder

    SciTech Connect

    Vondrasek, R.; Kolomiets, A.; Levand, A.; Pardo, R.; Savard, G.; Scott, R.

    2011-05-15

    An electron cyclotron resonance charge breeder for the Californium rare ion breeder upgrade (CARIBU), a new radioactive beam facility for the Argonne Tandem Linac Accelerator System (ATLAS), has been constructed and commissioned. Charge breeding efficiencies up to 15.6% have been realized for stable beams with a typical breeding time of 10 ms/charge state. The CARIBU system has been undergoing commissioning tests utilizing a 100 mCi {sup 252}Cf fission source. A charge breeding efficiency of 14.8 {+-} 5% has been achieved for the first radioactive beam of {sup 143}Cs{sup 27+}.

  2. BDDR, a new CEA technological and operating reactor database

    SciTech Connect

    Soldevilla, M.; Salmons, S.; Espinosa, B.

    2013-07-01

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

  3. Challenges and Opportunities Associated with Simultaneous Energy Cane and Sugarcane Genetic Improvement -- Results of a Survey of International Sugarcane Breeders

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Following Brazil's dramatic success in utilizing sugarcane (Saccharum spp.) for large-scale ethanol production, and with a growing interest in energy crops worldwide, sugarcane breeders have been charged with genetically improving cane as an energy crop. We conducted a survey of sugarcane breeders i...

  4. The Impact of Yeast Culture Residue on the Suppression of Dietary Aflatoxin on the Performance of Broiler Breeder Hens

    Technology Transfer Automated Retrieval System (TEKTRAN)

    A study was conducted to examine the effect of yeast culture residue (YCR) on the suppression of aflatoxicosis in broiler breeder hens. One hundred twenty, 35-wk-old, Cobb broiler breeder hens of the same cross were fed diets supplemented with aflatoxin (AF) (0 or 3 mg/kg) and YCR (0 or 2 lb/ton) s...

  5. Improved computational model (AQUIFAS) for activated sludge, integrated fixed-film activated sludge, and moving-bed biofilm reactor systems, part II: multilayer biofilm diffusional model.

    PubMed

    Sen, Dipankar; Randall, Clifford W

    2008-07-01

    Research was undertaken to develop a diffusional model of the biofilm that can be applied in lieu of a semi-empirical model to upgrade an activated sludge system to an integrated fixed-film activated sludge (IFAS) or moving-bed biofilm reactor (MBBR) system. The model has been developed to operate with up to 12 cells (reactors) in series, with biofilm media incorporated to one or more of the zone cells, except the anaerobic zone cells. The values of the kinetic parameters for the model were measured using pilot-scale activated sludge, IFAS, and MBBR systems. The biofilm is divided into 12 layers and has a stagnant liquid layer. Diffusion and substrate utilization are calculated for each layer. The equations are solved simultaneously using a finite difference technique. The biofilm flux model is then linked to the activated sludge model. Advanced features include the ability to compute the biofilm thickness and the effect of biofilm thickness on performance. The biofilm diffusional model is also used to provide information and create a table of biofilm yields at different substrate concentrations that can be used in the semi-empirical model. PMID:18710146

  6. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  7. Food residue recycling by swine breeders in a developing economy: a case study in Da Nang, Viet Nam.

    PubMed

    Kato, Takaaki; Pham, Dung Thi Xuan; Hoang, Hai; Xue, Yonghai; Tran, Quang Van

    2012-12-01

    This study provides a detailed description of food residue collection by swine breeders in Da Nang, Viet Nam. In January 2011, the study surveyed 30 swine breeders in two villages with respect to locations, methods, prices, quantities, and prospects for food residue collection. The sampled swine breeders regularly visited 55 locations in central Da Nang to collect raw food residue. They then transferred the food residue to their piggeries, boiled it, and fed it to their swine. A regression analysis revealed that the total amount of food residue collected by a farm depends on the number of swine in the farm and the number of collections made per day. Swine breeders in Da Nang were estimated to collect 26.3 metric tons of organic waste per day, which amounted to 4.1% of domestic waste collected by the local government. Among the sampled swine breeders, 93% answered that they would continue using food residue for the next five years. PMID:22883688

  8. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1959-10-27

    A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

  9. Completed Decommissioning of the Research Reactor TRIGA Heidelberg We are specialised in Decommissioning a Research Reactor in Germany now

    SciTech Connect

    Juenger-Graef, B.; Hoever, K.; Moser, T.; Berthold, M.; Blenski, H.J.

    2006-07-01

    This paper describes the decommissioning of the TRIGA Heidelberg II reactor which was used until 1999, and of the TRIGA Heidelberg I reactor, which was for the last 20 years in a safe containment. (authors)

  10. Dynamic analysis of large suspended LMFBR reactor vessels

    SciTech Connect

    Ma, D.C.; Gvildys, J.; Chang, Y.W.

    1983-01-01

    Large breeder reactor vessels are often designed under the top-suspended condition. Since the vessel contains a large volume of liquid sodium as reactor coolant, the structural integrity of the vessel bottom head and its effect on the vessel dynamic response are of great importance to the safety and reliability of the reactor systems. This paper presents a dynamic analysis of the large suspended reactor vessel subjected to the horizontal earthquake excitation with the emphasis on the effect of bottom head vibration on fluid pressure and sloshing response. Unlike the conventional lumped mass method, the present analysis treats the liquid sodium as a continuum medium. As a result, the important effects ignored in the lumped mass method such as fluid coupling, fluid-structure interaction, interaction between sloshing and vessel vibration, etc. can be accounted into the analysis.

  11. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  12. CONVECTION REACTOR

    DOEpatents

    Hammond, R.P.; King, L.D.P.

    1960-03-22

    An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

  13. Comparative breeding characteristics of fusion and fast reactors.

    PubMed

    Fortescue, P

    1977-06-17

    Expressions are developed to allow ready comparison of a hybrid fission-fusion plant and a fast breeder with respect to the number of thermal reactors that their fissile production could support, both for their feed requirements and for the new inventory needs of an expanding industry. These relations are expressed in terms of the neutron multiplication factor obtained in the fusion blanket, and the analogous quantities represented by the conversion ratios of the fast and thermal fission associated with the comparison. Results are presented graphically both for the steady state and for industries of arbitrary growth rate, and include the influence of tritium production requirements. Even a modest blanket neutron multiplication factor could enable the hybrid fusion system greatly to outperform the fast breeder on this simple basis of material balances. PMID:17831749

  14. Breazeale Reactor Modernization Program

    SciTech Connect

    Davison, C. C.

    2003-04-16

    The Penn State Breazeale Nuclear Reactor is the longest operating licensed research reactor in the nation. The facility has played a key role in educating scientists, engineers and in providing facilities and services to researchers in many different disciplines. In order to remain a viable and effective research and educational institution, a multi-phase modernization project was proposed. Phase I was the replacement of the 25-year old reactor control and safety system along with associated wiring and hardware. This phase was fully funded by non-federal funds. Tasks identified in Phases II-V expand upon and complement the work done in Phase I to strategically implement state-of-the-art technologies focusing on identified national needs and priorities of the future.

  15. Complex dynamic behavior in the bromate-oxalic acid-acetone-Mn(II) oscillating reaction in a continuous stirred tank reactor (CSTR)

    NASA Astrophysics Data System (ADS)

    Silva, Lucyane C.; Faria, Roberto B.

    2007-05-01

    The oscillating reaction bromate-oxalic acid-acetone-Mn(II)-sulfuric acid was observed for the first time in a CSTR at 20 °C. Depending on the bromate concentrations and flow rate, the system showed large amplitude oscillations, two kinds of mixed mode oscillations, quasiperiodicity and bursts of large amplitude oscillations, all mapped in a phase diagram. More complex behavior was favored at low bromate concentrations. The system without acetone was discovered to oscillate too, but the more complex patterns were not seen, indicating that acetone is implied in their formation.

  16. Sperm production and testicular development of broiler breeder males reared on shortened growth cycles

    Technology Transfer Automated Retrieval System (TEKTRAN)

    Feed restriction is an important tool used in the rearing of broiler breeders to control growth and maintain body weight. Feed restriction during the growing phase typically provides 60-80% less than what birds would consume if provided feed ad libitum, resulting in a perceived animal welfare issu...

  17. Sequencing analysis of Mycoplasma gallisepticum wild strains in vaccinated chicken breeder flocks.

    PubMed

    Khalifa, Rabab; Eissa, Sabry; El-Hariri, Mahmoud; Refai, Mohamed

    2014-01-01

    Mycoplasma gallisepticum (MG) infection is still of continuing economic concern in commercial broiler breeder chicken flocks in Egypt. MG infection continues to emerge despite the application of vaccination programs in breeder flocks. This prompted flock surveillance including MG isolation and molecular characterization of the circulating MG strains. The present study was concerned with 15 broiler breeder flocks of different ages (5-51 weeks). Three flocks were apparently healthy and 12 flocks were diseased. The aim of the study was to characterize the MG strains recovered from tracheal swabs. Four positive MG DNA extracts identified by rt-PCR and confirmed by isolation were subjected to sequencing of the mgc2 gene and intergenic spacer region (IGSR). The current molecular study demonstrated the presence of 3 different wild-type MG strains (RabE1-08, RabE2-09 and RabE3-09) in vaccinated diseased flocks, while the fourth strain (RabE4-08), which was isolated from a nonvaccinated apparently healthy breeder flock, scored 100% of homology and similarity to the F-strain vaccine by the sequence analysis of mgc2 and IGSR. It can be assumed that the vaccine F strain, which is supposed to replace field strains not only failed to do that, but also infected nonvaccinated flocks. Accordingly, there is a need to revise the control program including vaccine strategy in parallel with biosecurity measures. PMID:24525899

  18. Mating behavior and fertility of broiler breeder males reared on shortened growth cycles

    Technology Transfer Automated Retrieval System (TEKTRAN)

    One of the more difficult tasks when raising broiler breeder cockerels is controlling weight gain in the rearing house without inflicting excess stress. This is a period of time for the young male when many portions of reproductive system are in the formative stages and, if neglected, can have lif...

  19. Materials data base and design equations for the UCLA solid breeder blanket

    SciTech Connect

    Sharafat, S.; Amodeo, R.; Ghoniem, N.M.

    1986-02-01

    The materials and properties investigated for this blanket study are listed. The phenomenological equations and mathematical fits for all materials and properties considered are given. Efforts to develop a swelling equation based on the few experimental data points available for breeder materials are described. The sintering phenomena for ceramics is investigated.

  20. Ionization efficiency studies with charge breeder and conventional electron cyclotron resonance ion source

    SciTech Connect

    Koivisto, H. Tarvainen, O.; Toivanen, V.; Komppula, J.; Kronholm, R.; Lamy, T.; Angot, J.; Delahaye, P.; Maunoury, L.; Patti, G.; Standylo, L.; Steczkiewicz, O.; Choinski, J.

    2014-02-15

    Radioactive Ion Beams play an increasingly important role in several European research facility programs such as SPES, SPIRAL1 Upgrade, and SPIRAL2, but even more for those such as EURISOL. Although remarkable advances of ECRIS charge breeders (CBs) have been achieved, further studies are needed to gain insight on the physics of the charge breeding process. The fundamental plasma processes of charge breeders are studied in the frame of the European collaboration project, EMILIE, for optimizing the charge breeding. Important information on the charge breeding can be obtained by conducting similar experiments using the gas mixing and 2-frequency heating techniques with a conventional JYFL 14 GHz ECRIS and the LPSC-PHOENIX charge breeder. The first experiments were carried out with noble gases and they revealed, for example, that the effects of the gas mixing and 2-frequency heating on the production of high charge states appear to be additive for the conventional ECRIS. The results also indicate that at least in the case of noble gases the differences between the conventional ECRIS and the charge breeder cause only minor impact on the production efficiency of ion beams.

  1. Breeder survey, tools, and resources to visualize diversity and pedigree relationships at MaizeGDB

    Technology Transfer Automated Retrieval System (TEKTRAN)

    In collaboration with maize researchers, the MaizeGDB Team prepared a survey to identify breeder needs for visualizing pedigrees, diversity data, and haplotypes, and distributed it to the maize community on behalf of the Maize Genetics Executive Committee (Summer 2015). We received 48 responses from...

  2. The Role of Feeding Regimens in Regulating Metabolism of Sexually Mature Broiler Breeders

    Technology Transfer Automated Retrieval System (TEKTRAN)

    A trial was conducted to determine the effects of different rearing feed regimens on plasma hormone and metabolite levels and hepatic lipid metabolism and gene expression on sexually mature broiler breeders. Cobb 500 birds were divided into two groups at 4 weeks of age and fed either everyday (ED) ...

  3. Cross sections for actinide burner reactors

    SciTech Connect

    Difilippo, F.C.

    1991-01-01

    Recent studies have shown the feasibility of burning higher actinides (i.e., transuranium (TRU) elements excluding plutonium) in ad hoc designed reactors (Actinide Burner Reactors: ABR) which, because of their hard neutron spectra, enhance the fission of TRU. The transmutation of long-lived radionuclides into stable or short-lived isotopes reduces considerably the burden of handling high-level waste from either LWR or Fast Breeder Reactors (FBR) fuels. Because of the large concentrations of higher actinides in these novel reactor designs the Doppler effect due to TRU materials is the most important temperature coefficient from the point of view of reactor safety. Here we report calculations of energy group-averaged capture and fission cross sections as function of temperature and dilution for higher actinides in the resolved and unresolved resonance regions. The calculations were done with the codes SAMMY in the resolved region and URR in the unresolved regions and compared with an independent calculation. 4 refs., 2 figs., 2 tabs.

  4. Current status and directions for fast reactor reprocessing

    SciTech Connect

    Burch, W.D.

    1983-01-01

    The development of fast breeder reactors (FBRs) for commercial electric power production has been under way in several countries for more than 20 years. In the United States as elsewhere, early work was centered on small reactors to prove the feasibility of concepts and later was followed by larger reactors to test engineering features and to develop fuel technology. In the early 1970s, with the perceived crisis in electrical generation expected late in this century, major efforts were mounted to plan and carry out comprehensive development programs to ensure the capability to develop and begin using this new form of nuclear power by the end of this century. This comprehensive effort included the first serious efforts directed toward the supporting fuel cycle activities. However, because of the effects of the oil price rise and resulting conservation, a slowdown of industrial growth, and cut-backs in energy needs, there has been a decline in program activities. Unlike the fuel cycle for light-water reactors (LWRs), where supply and the back-end recycle and/or waste disposal activities can largely be uncoupled, recovery and recycle of fissile materials in spent fuel must be accomplished in one or two years in a practical breeder system. 3 references.

  5. Pyroprocessing of Oxidized Sodium-Bonded Fast Reactor Fuel -- an Experimental Study of Treatment Options for Degraded EBR-II Fuel

    SciTech Connect

    S. D. Herrmann; L. A. Wurth; N. J. Gese

    2013-09-01

    An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electrometallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li2O at 650 °C with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide. The experimental study illustrated how zirconium oxide and sodium oxide present different challenges to a lithium-based electrolytic reduction system for conversion of select metal oxides to metal.

  6. Development of pyro-processing technology for thorium-fuelled molten salt reactor

    SciTech Connect

    Uhlir, J.; Straka, M.; Szatmary, L.

    2012-07-01

    The Molten Salt Reactor (MSR) is classified as the non-classical nuclear reactor type based on the specific features coming out from the use of liquid fuel circulating in the MSR primary circuit. Other uniqueness of the reactor type is based on the fact that the primary circuit of the reactor is directly connected with the on-line reprocessing technology, necessary for keeping the reactor in operation for a long run. MSR is the only reactor system, which can be effectively operated within the {sup 232}Th- {sup 233}U fuel cycle as thorium breeder with the breeding factor significantly higher than one. The fuel cycle technologies proposed as ford the fresh thorium fuel processing as for the primary circuit fuel reprocessing are pyrochemical and mainly fluoride. Although these pyrochemical processes were never previously fully verified, the present-day development anticipates an assumption for the successful future deployment of the thorium-fuelled MSR technology. (authors)

  7. NEUTRONIC REACTOR

    DOEpatents

    Fraas, A.P.; Mills, C.B.

    1961-11-21

    A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

  8. REACTOR COOLING

    DOEpatents

    Quackenbush, C.F.

    1959-09-29

    A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

  9. Pyroprocessing of oxidized sodium-bonded fast reactor fuel - An experimental study of treatment options for degraded EBR-II fuel

    SciTech Connect

    Hermann, S.D.; Gese, N.J.; Wurth, L.A.

    2013-07-01

    An experimental study was conducted to assess pyrochemical treatment options for degraded EBR-II fuel. As oxidized material, the degraded fuel would need to be converted back to metal to enable electrorefining within an existing electro-metallurgical treatment process. A lithium-based electrolytic reduction process was studied to assess the efficacy of converting oxide materials to metal with a particular focus on the impact of zirconium oxide and sodium oxide on this process. Bench-scale electrolytic reduction experiments were performed in LiCl-Li{sub 2}O at 650 C. degrees with combinations of manganese oxide (used as a surrogate for uranium oxide), zirconium oxide, and sodium oxide. In the absence of zirconium or sodium oxide, the electrolytic reduction of MnO showed nearly complete conversion to metal. The electrolytic reduction of a blend of MnO-ZrO{sub 2} in LiCl - 1 wt% Li{sub 2}O showed substantial reduction of manganese, but only 8.5% of the zirconium was found in the metal phase. The electrolytic reduction of the same blend of MnO-ZrO{sub 2} in LiCl - 1 wt% Li{sub 2}O - 6.2 wt% Na{sub 2}O showed substantial reduction of manganese, but zirconium reduction was even less at 2.4%. This study concluded that ZrO{sub 2} cannot be substantially reduced to metal in an electrolytic reduction system with LiCl - 1 wt% Li{sub 2}O at 650 C. degrees due to the perceived preferential formation of lithium zirconate. This study also identified a possible interference that sodium oxide may have on the same system by introducing a parasitic and cyclic reaction of dissolved sodium metal between oxidation at the anode and reduction at the cathode. When applied to oxidized sodium-bonded EBR-II fuel (e.g., U-10Zr), the prescribed electrolytic reduction system would not be expected to substantially reduce zirconium oxide, and the accumulation of sodium in the electrolyte could interfere with the reduction of uranium oxide, or at least render it less efficient.

  10. The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

    SciTech Connect

    Smith, R.R.

    1986-01-01

    This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

  11. Effects of environmental temperature and dietary energy on energy partitioning coefficients of female broiler breeders.

    PubMed

    Pishnamazi, A; Renema, R A; Paul, D C; Wenger, I I; Zuidhof, M J

    2015-10-01

    With increasing disparity between broiler breeder target weights and broiler growth potential, maintenance energy requirements have become a larger proportion of total broiler breeder energy intake. Because energy is partitioned to growth and egg production at a lower priority than maintenance, accurate prediction of maintenance energy requirements is important for practical broiler breeder feed allocation decisions. Environmental temperature affects the maintenance energy requirement by changing rate of heat loss to the environment. In the ME system, heat production (energy lost) is part of the maintenance requirement (ME). In the current study, a nonlinear mixed model was derived to predict ME partitioning of broiler breeder hens under varied temperature conditions. At 21 wk of age, 192 Ross 708 hens were individually caged within 6 controlled environmental chambers. From 25 to 41 wk, 4 temperature treatments (15°C, 19°C, 23°C, and 27°C) were randomly assigned to the chambers for 2-week periods. Half of the birds in each chamber were fed a high-energy (HE; 2,912 kcal/kg) diet, and half were fed a low-energy (LE; 2,790 kcal/kg) diet. The nonlinear mixed regression model included a normally distributed random term representing individual hen maintenance, a quadratic response to environmental temperature, and linear ADG and egg mass (EM) coefficients. The model assumed that energy requirements for BW gain and egg production were not influenced by environmental temperature because hens were homeothermic, and the cellular processes for associated biochemical processes occurred within a controlled narrow core body temperature range. Residual feed intake (RFI) and residual ME (RME) were used to estimate efficiency. A quadratic effect of environmental temperature on broiler breeder MEm was predicted ( < 0.0001), with a minimum energy expenditure at 24.3°C. Predicted ME at 21°C was 92.5 kcal/kg; requirements for gain and EM were 2.126 and 1.789 kcal/g, respectively

  12. Optimizing a neutron-beam focusing device for the direct geometry time-of-flight spectrometer TOFTOF at the FRM II reactor source

    NASA Astrophysics Data System (ADS)

    Rasmussen, N. G.; Simeoni, G. G.; Lefmann, K.

    2016-04-01

    A dedicated beam-focusing device has been designed for the direct geometry thermal-cold neutron time-of-flight spectrometer TOFTOF at the neutron facility FRM II (Garching, Germany). The prototype, based on the compressed Archimedes' mirror concept, benefits from the adaptive-optics technology (adjustable supermirror curvature) and the compact size (only 0.5 m long). We have simulated the neutron transport across the entire guide system. We present a detailed computer characterization of the existing device, along with the study of the factors mostly influencing the future improvement. We have optimized the simulated prototype as a function of the neutron wavelength, accounting also for all relevant features of a real instrument like the non-reflecting side edges. The results confirm the "chromatic" displacement of the focal point (flux density maximum) at fixed supermirror curvature, and the ability of a variable curvature to keep the focal point at the sample position. Our simulations are in excellent agreement with theoretical predictions and the experimentally measured beam profile. With respect to the possibility of a further upgrade, we find that supermirror coatings with m-values higher than 3.5 would have only marginal influence on the optimal behaviour, whereas comparable spectrometers could take advantage of longer focusing segments, with particular impact for the thermal region of the neutron spectrum.

  13. Characterization of Irradiated Metal Waste from the Pyrometallurgical Treatment of Used EBR-II Fuel

    SciTech Connect

    B.R. Westphal; K.C. Marsden; W.M. McCartin; S.M. Frank; D.D. Keiser, Jr.; T.S. Yoo; D. Vaden; D.G. Cummings; K.J. Bateman; J. J. Giglio; T. P. O'Holleran; P. A. Hahn; M. N. Patterson

    2013-03-01

    As part of the pyrometallurgical treatment of used Experimental Breeder Reactor-II fuel, a metal waste stream is generated consisting primarily of cladding hulls laden with fission products noble to the electrorefining process. Consolidation by melting at high temperature [1873 K (1600 degrees C)] has been developed to sequester the noble metal fission products (Zr, Mo, Tc, Ru, Rh, Te, and Pd) which remain in the iron-based cladding hulls. Zirconium from the uranium fuel alloy (U-10Zr) is also deposited on the hulls and forms Fe-Zr intermetallics which incorporate the noble metals as well as residual actinides during processing. Hence, Zr has been chosen as the primary indicator for consistency of the metal waste. Recently, the first production-scale metal waste ingot was generated and sampled to monitor Zr content for Fe-Zr intermetallic phase formation and validation of processing conditions. Chemical assay of the metal waste ingot revealed a homogeneous distribution of the noble metal fission products as well as the primary fuel constituents U and Zr. Microstructural characterization of the ingot confirmed the immobilization of the noble metals in the Fe-Zr intermetallic phase.

  14. Characterization of Irradiated Metal Waste from the Pyrometallurgical Treatment of Used EBR-II Fuel

    NASA Astrophysics Data System (ADS)

    Westphal, Brian R.; Frank, S. M.; McCartin, W. M.; Cummings, D. G.; Giglio, J. J.; O'Holleran, T. P.; Hahn, P. A.; Yoo, T. S.; Marsden, K. C.; Bateman, K. J.; Patterson, M. N.

    2015-01-01

    As part of the pyrometallurgical treatment of used Experimental Breeder Reactor-II fuel, a metal waste stream is generated consisting primarily of cladding hulls laden with fission products noble to the electrorefining process. Consolidation by melting at high temperature [1873 K (1600 °C)] has been developed to sequester the noble metal fission products (Zr, Mo, Tc, Ru, Rh, Te, and Pd) which remain in the iron-based cladding hulls. Zirconium from the uranium fuel alloy (U-10Zr) is also deposited on the hulls and forms Fe-Zr intermetallics which incorporate the noble metals as well as residual actinides during processing. Hence, Zr has been chosen as the primary indicator for consistency of the metal waste. Recently, the first production-scale metal waste ingot was generated and sampled to monitor Zr content for Fe-Zr intermetallic phase formation and validation of processing conditions. Chemical assay of the metal waste ingot revealed a homogeneous distribution of the noble metal fission products as well as the primary fuel constituents U and Zr. Microstructural characterization of the ingot confirmed the immobilization of the noble metals in the Fe-Zr intermetallic phase.

  15. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

    A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

  16. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.; Johnson, H.W.

    1961-04-01

    BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

  17. Effect of lycopene on semen quality, fertility and native immunity of broiler breeder.

    PubMed

    Mangiagalli, M G; Martino, P A; Smajlovic, T; Guidobono Cavalchini, L; Marelli, S P

    2010-02-01

    1. The effect of drinking water supplementation with lycopene on the semen quality, fertility and immunity of broiler breeders was evaluated. 2. Broiler breeder males were individually caged from 25 to 42 weeks old and divided into two group: L group, treated birds (lycopene 0.5 g/l) and C group, control birds. Laying hens were divided into two groups and artificially inseminated. 3. Semen variables were evaluated and daily fertility recorded. Serum bactericidal activity was tested. 4. Semen production and viability were affected by lycopene supplementation. Serum bactericidal activity was better in L than in C group. The fertility rate curve of the L group displayed a positive trend. PMID:20390581

  18. Clinical outbreak of Bordetella avium infection in two turkey breeder flocks.

    PubMed

    Kelly, B J; Ghazikhanian, G Y; Mayeda, B

    1986-01-01

    An acute upper respiratory disease was observed in two broad-breasted white (BBW) turkey primary breeder flocks. Associated clinical signs included sneezing, depression, and a deep dry cough originating from large conducting airways. Morbidity reached approximately 15-20% of the hens in an affected house. None of the turkeys died, and total feed consumption was not affected. A minimal effect upon egg production was noticed. Sera from an acutely affected flock exhibited a marked rise in titer to Bordetella avium compared with preinfection sera samples. In Case 1, B. avium was isolated in pure culture from affected birds. In Case 2, B. avium was diagnosed by serological results and clinical signs; bacteriological examination was not attempted. The findings presented here are consistent with an acute clinical outbreak of B. avium-induced turkey rhinotracheitis (turkey coryza) in BBW turkey breeder hens. PMID:3729868

  19. Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

    NASA Astrophysics Data System (ADS)

    Zhu, Qingjun; Li, Jia; Liu, Songlin

    2016-07-01

    In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the water-cooled ceramic breeder blanket for CFETR. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2014GB122000, and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

  20. Characterization of advertisements for puppies sold online: determinants of cost and a comparison with parent club breeders.

    PubMed

    Voris, H C; Wittum, T E; Rajala-Schultz, P J; Lord, L K

    2011-07-01

    The Internet is an increasingly common way for consumers to purchase puppies. Yet very little information is available about the types of puppies sold via the Internet. In addition these sales are not subject to United States Depart of Agriculture (USDA) regulation. The objectives of the study were to describe puppies sold via the Internet, to assess the characteristics that contribute to the cost of a puppy, and to compare puppies sold via the Internet with puppies sold by American Kennel Club (AKC) Parent Club breeders. Over 14 weeks in 2008, Yorkshire Terrier, Shih Tzu, English Bulldog, Boxer, and Labrador Retriever puppies for sale on two large-scale online puppy sales sites were categorized based on their Internet advertisements. Data were collected in three categories: puppy characteristics, health characteristics, and policies (such as spay/neuter requirement, health guarantee, and return policy). After the survey was completed, 25 AKC Parent Club breeders and 25 other breeders who advertised via one of the puppy sales websites were randomly selected and interviewed over the phone. Small breed puppies were most frequently advertised with 35.2% (1228/3485) of advertisements for Yorkshire Terriers and 23.0% (802/3485) for Shih Tzus. Almost one quarter of Internet breeders 768/3474 (22.2%) advertised four or more different dog breeds. Champion bloodlines increased the cost of a puppy of all breeds. AKC Parent Club breeders 21/25 (84%) were more likely to mention breed-specific health screening tests when compared to Internet breeders 7/25 (28%). Consumers should apply the same standards for purchasing from a breeder found through a puppy sales site as they would for purchasing from a local breeder. Breeders who advertise at one of the large-scale puppy sales websites are less knowledgeable about breed-specific health issues compared to an AKC Parent Club breeder. Internet breeders are less likely to perform these screening tests on their breeding dogs and may