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Sample records for advanced liquid-metal reactor

  1. Reference site selection report for the advanced liquid metal reactor at the Idaho National Engineering Laboratory

    SciTech Connect

    Sivill, R.L.

    1990-03-01

    This Reference Site Selection Report was prepared by EG G, Idaho Inc., for General Electric (GE) to provide information for use by the Department of Energy (DOE) in selecting a Safety Test Site for an Advanced Liquid Metal Reactor. Similar Evaluation studies are planned to be conducted at other potential DOE sites. The Power Reactor Innovative Small Module (PRISM) Concept was developed for ALMR by GE. A ALMR Safety Test is planned to be performed on a DOE site to demonstrate features and meet Nuclear Regulatory Commission Requirements. This study considered possible locations at the Idaho National Engineering Laboratory that met the ALMR Prototype Site Selection Methodology and Criteria. Four sites were identified, after further evaluation one site was eliminated. Each of the remaining three sites satisfied the criteria and was graded. The results were relatively close. Thus concluding that the Idaho National Engineering Laboratory is a suitable location for an Advanced Liquid Metal Reactor Safety Test. 23 refs., 13 figs., 9 tabs.

  2. Summary of advanced LMR (Liquid Metal Reactor) evaluations: PRISM (Power Reactor Inherently Safe Module) and SAFR (Sodium Advanced Fast Reactor)

    SciTech Connect

    Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Kennett, R.J.; Cheng, H.S.; Kroeger, P.G. )

    1989-10-01

    In support of the US Nuclear Regulatory Commission (NRC), Brookhaven National Laboratory (BNL) has performed independent analyses of two advanced Liquid Metal Reactor (LMR) concepts. The designs, sponsored by the US Department of Energy (DOE), the Power Reactor Inherently Safe Module (PRISM) (Berglund, 1987) and the Sodium Advanced Fast Reactor (SAFR) (Baumeister, 1987), were developed primarily by General Electric (GE) and Rockwell International (RI), respectively. Technical support was provided to DOE, RI, and GE, by the Argonne National Laboratory (ANL), particularly with respect to the characteristics of the metal fuels. There are several examples in both PRISM and SAFR where inherent or passive systems provide for a safe response to off-normal conditions. This is in contrast to the engineered safety systems utilized on current US Light Water Reactor (LWR) designs. One important design inherency in the LMRs is the inherent shutdown'', which refers to the tendency of the reactor to transition to a much lower power level whenever temperatures rise significantly. This type of behavior was demonstrated in a series of unscrammed tests at EBR-II (NED, 1986). The second key design feature is the passive air cooling of the vessel to remove decay heat. These systems, designated RVACS in PRISM and RACS in SAFR, always operate and are believed to be able to prevent core damage in the event that no other means of heat removal is available. 27 refs., 78 figs., 3 tabs.

  3. Performance analysis of a mixed nitride fuel system for an advanced liquid metal reactor

    SciTech Connect

    Lyon, W.F.; Baker, R.B.; Leggett, R.D.

    1990-11-01

    The conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design has been performed. As a first step, an intensive literature survey was completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins were designed and analyzed using the SIEX computer code. The analysis predicted that the nitride fuel consistently operated at peak temperatures and cladding strain levels that compared quite favorably with competing fuel designs. These results, along with data available in the literature on nitride fuel performance, indicate that a nitride fuel system should offer enhanced capabilities for advanced liquid metal reactors. 13 refs., 10 figs., 2 tabs.

  4. Safety aspects of the US advanced LMR (liquid metal reactor) design

    SciTech Connect

    Pedersen, D.R.; Gyorey, G.L.; Marchaterre, J.F.; Rosen, S.; General Electric Co., San Jose, CA; Argonne National Lab., IL; USDOE Assistant Secretary for Nuclear Energy, Washington, DC )

    1989-01-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the plant design program at General Electric based on the PRISM (Power Reactor Innovative Small Module) concept, and the Integral Fast Reactor program (IFR) at Argonne National Laboratory (ANL). The goal of the US program is to produce a standard, commercial ALMR, including the associated fuel cycle. This paper discusses the US regulatory framework for design of an ALMR, safety aspects of the IFR program at ANL, the IFR fuel cycle and actinide recycle, and the ALMR plant design program at GE. 6 refs., 5 figs.

  5. Robotics application for in-service inspection of the ALMR. [Advanced Liquid Metal Reactor (ALMR)

    SciTech Connect

    Kwant, W.; Ramsour, N.L. . Nuclear Energy Div.); Sweeney, F.J. )

    1993-01-01

    The US Advanced Liquid Metal Reactor (ALMR) Program is developing and licensing a reactor system that is compact for factory fabrication and modular construction. The design includes provisions for in-service inspection to verify performance and safety capabilities throughout the life of the plant. A DOE sponsored robotics team, comprised of members from the universities of Florida, Michigan, Tennessee, Texas and from Oak Ridge National Laboratory, is developing advanced inspection equipment using robotics for nuclear application. This equipment is compact and remotely operated and particularly suited for inspection of the ALMR. Extensive 3D simulations are used to refine and demonstrate the inspection methods. This paper focuses on inspection methods for the reactor vessel and the reactor vessel auxiliary cooling system (RVACS). Inspection capabilities are included for visual inspection of the reactor vessel outer surface and volumetric inspection of the welds. The robotics team is devising a compact crawler design with the capabilities to perform these inspections. Similarly, various robot concepts are being evaluated for accomplishing the RVACS visual inspection and cleaning procedures.

  6. EXPERIMENTAL LIQUID METAL FUEL REACTOR

    DOEpatents

    Happell, J.J.; Thomas, G.R.; Denise, R.P.; Bunts, J.L. Jr.

    1962-01-23

    A liquid metal fuel nuclear fission reactor is designed in which the fissionable material is dissolved or suspended in a liquid metal moderator and coolant. The liquid suspension flows into a chamber in which a critical amount of fissionable material is obtained. The fluid leaves the chamber and the heat of fission is extracted for power or other utilization. The improvement is in the support arrangement for a segrnented graphite core to permit dif ferential thermal expansion, effective sealing between main and blanket liquid metal flows, and avoidance of excessive stress development in the graphite segments. (AEC)

  7. Design of the reactor vessel inspection robot for the advanced liquid metal reactor

    SciTech Connect

    Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

    1994-06-01

    A consortium of four universities and Oak Ridge National Laboratory designed a prototype wall-crawling robot to perform weld inspection in an advanced nuclear reactor. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a mock non-hostile environment and shown to perform as expected, as detailed in this report.

  8. Liquid metal cooled nuclear reactor plant system

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

  9. Transient bowing of core assemblies in advanced liquid metal fast reactors

    SciTech Connect

    Kamal, S.A.; Orechwa, Y.

    1986-01-01

    Two alternative core restraint concepts are considered for a conceptual design of a 900 MWth liquid metal fast reactor core with a heterogeneous layout. The two concepts, known as limited free bowing and free flowering, are evaluated based on core bowing criteria that emphasize the enhancement of inherent reactor safety. The core reactivity change during a postulated loss of flow transient is calculated in terms of the lateral displacements and displacement-reactivity-worths of the individual assemblies. The NUBOW-3D computer code is utilized to determine the assembly deformations and interassembly forces that arise when the assemblies are subjected to temperature gradients and irradiation induced creep and swelling during the reactor operation. The assembly ducts are made of the ferritic steel HT-9 and remain in the reactor core for four-years at full power condition. Whereas both restraint systems meet the bowing criteria, a properly designed limited free bowing system appears to be more advantageous than a free flowering system from the point of view of enhancing the reactor inherent safety.

  10. Advanced liquid-metal-reactor development at ANL during the 1980s

    SciTech Connect

    Wade, D.C. )

    1990-01-01

    The fundamental long-term rationale for the liquid-metal reactor (LMR) remains unchanged - to provide for resource extension by consuming the more abundant isotope of uranium. The design goals for the next generation of breeder reactors include a concerted effort to provide engineering solutions that can positively impact on the technical issues and on the public perception issues that harry the current generation of commercial power reactors. The work at Argonne National Laboratory since late 1983 has approached these design goals for the next-generation-reactor enterprise as a whole and has based the approach on a closed, fissile-self-sufficient, transuranic-self-consuming fuel cycle that employs a uranium/plutonium metallic alloy fuel form. Pyrometallurgically based reprocessing and remote injection casting fuel refabrication form the basis on which the closed fuel cycle is expected to provide low fuel cycle costs even on an incremental dedicated fuel cycle facility basis. Passive safety features that decouple reactor shutdown and decay heat removal from reliance on balance-of-plant equipment form the basis for reduced capital costs. Preapproved licensing form the basis for shorter and more predictable construction cycles. Incomplete fission product separation and incomplete uranium/plutonium separation of the product streams provide for a deterrent to subnational diversion. Finally, the waste management strategy is based on pyrometallurgical reprocessing in which all transuranics can be made to follow the plutonium-rich product stream.

  11. Three-dimensional thermal-hydraulic analysis of an advanced liquid metal reactor design by the COMMIX computer code

    SciTech Connect

    Shin, Y.W.

    1991-01-01

    The emphasis in the development of advanced liquid metal reactors (LMRs) is on inherent safety and economics. One such feature is the adoption of thermal radiation and natural-convection cooling of the reactor to handle decay heat following a reactor shutdown. The decay heat removal feature of the LMR design under investigation here involves an in-vessel overflow of hot-pool sodium next to the reactor vessel (RV) in such a way that in the event of a reactor heat-up due to decay heat, the RV temperature is elevated and thereby the rate of heat removal from the reactor to the ambient air is increased. The purpose is to limit the temperature rise due to the decay heat. The objective of this study is to evaluate the performance of the simple passive decay heat removal feature of an advanced LMR design based on radiation and natural convection. The evaluation was carried out by performing calculations using the COMMIX Code for two cases, one with the passive heat removal features and the other without the features, and comparing the results. 2 refs., 6 figs., 1 tab.

  12. Selection of the reference steam generator for the advanced liquid metal reactor

    SciTech Connect

    Loewen, Eric P.; Boardman, Chuck

    2007-07-01

    In February 2006 President Bush announced the Advanced Energy Initiative, which included the Department of Energy's (DOE) Global Nuclear Energy Partnership (GNEP). GNEP has seven broad goals; one of the major elements being to develop and deploy advanced nuclear fuel recycling technology that includes consuming spent nuclear fuel in an Advanced Recycling Reactor (ARR). DOE is contemplating accelerating the deployment of these technologies to achieve the construction of a commercial scale application of these technologies. DOE now defines this approach as 'two simultaneous tracks: (1) deployment of commercial scale facilities for which advanced technologies are available now or in the near future, and (2) further research and development of transmutation fuels technologies'. GEHitachi Nuclear Energy Americas LLC (GHNEA) believes an integrated technical solution is achievable in the near term to accelerate the commercial demonstration of GNEP infrastructure. The GHNEA ARR concept involves a single integrated recycling facility sized to service a single reactor module ARR capable of destroying light water and fast reactor sourced actinides. This paper describes the bases and rationale behind the selection of the helical coil steam generator (HCSG) as the reference steam generator concept for the ALMR and S-PRISM reactor concepts. (authors)

  13. 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.

  14. Testimony of Fred R. Mynatt before the Energy Research and Development Subcommittee of the Committee on Science, Space, and Technology, US House of Representatives. [Advanced fuel technology, gas-cooled reactor technology, and liquid metal-cooled reactor technology programs

    SciTech Connect

    Mynatt, F.R.

    1987-03-18

    This report provides a description of the statements submitted for the record to the committee on Science, Space, and Technology of the United States House of Representatives. These statements describe three principal areas of activity of the Advanced Reactor Technology Program of the Department of Energy (DOE). These areas are advanced fuel cycle technology, modular high-temperature gas-cooled reactor technology, and liquid metal-cooled reactor. The areas of automated reactor control systems, robotics, materials and structural design shielding and international cooperation were included in these statements describing the Oak Ridge National Laboratory's efforts in these areas. (FI)

  15. Liquid metal pump for nuclear reactors

    DOEpatents

    Allen, H.G.; Maloney, J.R.

    1975-10-01

    A pump for use in pumping high temperature liquids at high pressures, particularly liquid metals used to cool nuclear reactors is described. It is of the type in which the rotor is submerged in a sump but is fed by an inlet duct which bypasses the sump. A chamber, kept full of fluid, surrounds the pump casing into which fluid is bled from the pump discharge and from which fluid is fed to the rotor bearings and hence to the sump. This equalizes pressure inside and outside the pump casing and reduces or eliminates the thermal shock to the bearings and sump tank.

  16. Small Liquid Metal Cooled Reactor Safety Study

    SciTech Connect

    Minato, A; Ueda, N; Wade, D; Greenspan, E; Brown, N

    2005-11-02

    The Small Liquid Metal Cooled Reactor Safety Study documents results from activities conducted under Small Liquid Metal Fast Reactor Coordination Program (SLMFR-CP) Agreement, January 2004, between the Central Research Institute of the Electric Power Industry (CRIEPI) of Japan and the Lawrence Livermore National Laboratory (LLNL)[1]. Evaluations were completed on topics that are important to the safety of small sodium cooled and lead alloy cooled reactors. CRIEPI investigated approaches for evaluating postulated severe accidents using the CANIS computer code. The methods being developed are improvements on codes such as SAS 4A used in the US to analyze sodium cooled reactors and they depend on calibration using safety testing of metal fuel that has been completed in the TREAT facility. The 4S and the small lead cooled reactors in the US are being designed to preclude core disruption from all mechanistic scenarios, including selected unprotected transients. However, postulated core disruption is being evaluated to support the risk analysis. Argonne National Laboratory and the University of California Berkeley also supported LLNL with evaluation of cores with small positive void worth and core designs that would limit void worth. Assessments were also completed for lead cooled reactors in the following areas: (1) continuing operations with cladding failure, (2) large bubbles passing through the core and (3) recommendations concerning reflector control. The design approach used in the US emphasizes reducing the reactivity in the control mechanisms with core designs that have essentially no, or a very small, reactivity change over the core life. This leads to some positive void worth in the core that is not considered to be safety problem because of the inability to identify scenarios that would lead to voiding of lead. It is also believed that the void worth will not dominate the severe accident analysis. The approach used by 4S requires negative void worth throughout

  17. 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

  18. Lessons Learned about Liquid Metal Reactors from FFTF Experience

    SciTech Connect

    Wootan, David W.; Casella, Andrew M.; Omberg, Ronald P.; Burke, Thomas M.; Grandy, Christopher

    2016-09-20

    The Fast Flux Test Facility (FFTF) is the most recent liquid-metal reactor (LMR) to operate in the United States, from 1982 to 1992. FFTF is located on the DOE Hanford Site near Richland, Washington. The 400-MWt sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission test reactor was designed specifically to irradiate Liquid Metal Fast Breeder Reactor (LMFBR) fuel and components in prototypical temperature and flux conditions. FFTF played a key role in LMFBR development and testing activities. The reactor provided extensive capability for in-core irradiation testing, including eight core positions that could be used with independent instrumentation for the test specimens. In addition to irradiation testing capabilities, FFTF provided long-term testing and evaluation of plant components and systems for LMFBRs. The FFTF was highly successful and demonstrated outstanding performance during its nearly 10 years of operation. The technology employed in designing and constructing this reactor, as well as information obtained from tests conducted during its operation, can significantly influence the development of new advanced reactor designs in the areas of plant system and component design, component fabrication, fuel design and performance, prototype testing, site construction, and reactor operations. The FFTF complex included the reactor, as well as equipment and structures for heat removal, containment, core component handling and examination, instrumentation and control, and for supplying utilities and other essential services. The FFTF Plant was designed using a “system” concept. All drawings, specifications and other engineering documentation were organized by these systems. Efforts have been made to preserve important lessons learned during the nearly 10 years of reactor operation. A brief summary of Lessons Learned in the following areas will be discussed: Acceptance and Startup Testing of FFTF FFTF Cycle Reports

  19. Safety and control of modular liquid-metal reactors

    SciTech Connect

    Sackett, J.I. ); Sevy, R.H.; Wei, T.Y.C. )

    1989-01-01

    As part of recent development efforts on advanced reactor designs Argonne National Laboratory has proposed the integral fast reactor (IFR) concept. The IFR concept is currently being applied to modular-sized reactors, which would be built in multiple power packs together with an integrated fuel-cycle facility. It has been amply demonstrated that the concept, as applied to modular designs, has significant advantages in regard to anticipated transients without scram (ATWS). Attention is now focused on whether or not those advantages derived from the IFR traits can be translated to the operational/design-basis-accident class of transients. Inherent operability, in which reactor power control is effected through the use of primary pumps and balance-of-plant (BOP) swings rather than through the active motion of control rods, is a proposal to utilize the enhanced inherent feedback response of the IFR to improve the operating characteristics of liquid-metal reactors (LMRs). The scheme has associated with it potential advantages in the areas of plant control and design simplification. This study on inherent operability in modular LMRs therefore has implications for both operational and ATWS events. Current intentions are to analytically explore possibilities of applying various schemes to advanced LMRs with the aid of the SASSYS system code and then to test viable alternatives in the Experimental Breeder Reactor II (EBR-II) plant under the auspices of the inherent safety operability testing program.

  20. Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1989-01-01

    The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs.

  1. Liquid metal cooled nuclear reactors with passive cooling system

    DOEpatents

    Hunsbedt, Anstein; Fanning, Alan W.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of cooling medium flow circuits which cooperate to remove and carry heat away from the fuel core upon loss of the normal cooling flow circuit to areas external thereto.

  2. Advance Liquid Metal Reactor Discrete Dynamic Event Tree/Bayesian Network Analysis and Incident Management Guidelines (Risk Management for Sodium Fast Reactors)

    SciTech Connect

    Denman, Matthew R.; Groth, Katrina M.; Cardoni, Jeffrey N.; Wheeler, Timothy A.

    2015-04-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self-correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the system's design to manage the accident. Inherently and passively safe designs are laudable, but nonetheless extreme boundary conditions can interfere with the design attributes which facilitate inherent safety, thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayesian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The authors would like to acknowledge the U.S. Department of Energy's Office of Nuclear Energy for funding this research through Work Package SR-14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at Argonne National Laboratory, Oak Ridge National Laboratory, and Idaho National Laboratory for their continue d contributions to the advanced reactor PRA mission area.

  3. Passive cooling safety system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.; Hui, Marvin M.; Berglund, Robert C.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  4. Indirect passive cooling system for liquid metal cooled nuclear reactors

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1990-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel. The passive cooling system includes a closed primary fluid circuit through the partitions surrounding the reactor vessel and a partially adjoining secondary open fluid circuit for carrying transferred heat out into the atmosphere.

  5. 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.

  6. Present status of liquid metal research for a fusion reactor

    NASA Astrophysics Data System (ADS)

    Tabarés, Francisco L.

    2016-01-01

    Although the use of solid materials as targets of divertor plasmas in magnetic fusion research is accepted as the standard solution for the very challenging issue of power and particle handling in a fusion reactor, a generalized feeling that the present options chosen for ITER will not represent the best choice for a reactor is growing up. The problems found for tungsten, the present selection for the divertor target of ITER, in laboratory tests and in hot plasma fusion devices suggest so. Even in the absence of the strong neutron irradiation expected in a reactor, issues like surface melting, droplet ejection, surface cracking, dust generation, etc., call for alternative solutions in a long pulse, high efficient fusion energy-producing continuous machine. Fortunately enough, decades of research on plasma facing materials based on liquid metals (LMs) have produced a wealth of appealing ideas that could find practical application in the route to the realization of a commercial fusion power plant. The options presently available, although in a different degree of maturity, range from full coverage of the inner wall of the device with liquid metals, so that power and particle exhaust together with neutron shielding could be provided, to more conservative combinations of liquid metal films and conventional solid targets basically representing a sort of high performance, evaporative coating for the alleviation of the surface degradation issues found so far. In this work, an updated review of worldwide activities on LM research is presented, together with some open issues still remaining and some proposals based on simple physical considerations leading to the optimization of the most conservative alternatives.

  7. Hydrogen permeation resistant layers for liquid metal reactors

    SciTech Connect

    McGuire, J.C.

    1980-03-01

    Reviewing the literature in the tritium diffusion field one can readily see a wide divergence in results for both the response of permeation rate to pressure, and the effect of oxide layers on total permeation rates. The basic mechanism of protective oxide layers is discussed. Two coatings which are less hydrogen permeable than the best naturally occurring oxide are described. The work described is part of an HEDL-ANL cooperative research program on Tritium Permeation in Liquid Metal Cooled Reactors. This includes permeation work on hydrogen, deuterium, and tritium with the hydrogen-deuterium research leading to the developments presented.

  8. Method of shielding a liquid-metal-cooled reactor

    DOEpatents

    Sayre, Robert K.

    1978-01-01

    The primary heat transport system of a nuclear reactor -- particularly for a liquid-metal-cooled fast-breeder reactor -- is shielded and protected from leakage by establishing and maintaining a bed of a powdered oxide closely and completely surrounding all components thereof by passing a gas upwardly therethrough at such a rate as to slightly expand the bed to the extent that the components of the system are able to expand without damage and yet the particles of the bed remain close enough so that the bed acts as a guard vessel for the system. Preferably the gas contains 1 to 10% oxygen and the gas is passed upwardly through the bed at such a rate that the lower portion of the bed is a fixed bed while the upper portion is a fluidized bed, the line of demarcation therebetween being high enough that the fixed bed portion of the bed serves as guard vessel for the system.

  9. Water storage of liquid-metal fast-breeder-reactor fuel

    SciTech Connect

    Meacham, S.A.

    1982-01-01

    The purpose of this paper is to present a general overview of a concept proposed for receiving and storing liquid metal fast breeder reactor (LMFBR) spent fuel. This work was done as part of the Consolidated Fuel Reprocessing Program (CFRP) at the Oak Ridge National Laboratory (ORNL). The CFRP has as its major objective the development of technology for reprocessing advanced nuclear reactor fuels. The program plans that research and development will be carried through to a sufficient scale, using irradiated spent fuel under plant operating conditions, to establish a basis for confident projection of reprocessing capability to support a breeder industry.

  10. 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.

  11. Method for passive cooling liquid metal cooled nuclear reactors, and system thereof

    DOEpatents

    Hunsbedt, Anstein; Busboom, Herbert J.

    1991-01-01

    A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting from fuel decay during reactor shutdown. The passive cooling system comprises a plurality of partitions surrounding the reactor vessel in spaced apart relation forming intermediate areas for circulating heat transferring fluid which remove and carry away heat from the reactor vessel.

  12. Advances by the Integral Fast Reactor Program

    SciTech Connect

    Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

    1991-01-01

    The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs.

  13. Resistive sensor and electromagnetic actuator for feedback stabilization of liquid metal walls in fusion reactors

    NASA Astrophysics Data System (ADS)

    Mirhoseini, S. M. H.; Volpe, F. A.

    2016-12-01

    Liquid metal walls in fusion reactors will be subject to instabilities, turbulence, induced currents, error fields and temperature gradients that will make them locally bulge, thus entering in contact with the plasma, or deplete, hence exposing the underlying solid substrate. To prevent this, research has begun to actively stabilize static or flowing free-surface liquid metal layers by locally applying forces in feedback with thickness measurements. Here we present resistive sensors of liquid metal thickness and demonstrate \\mathbf{j}× \\mathbf{B} actuators, to locally control it.

  14. Experimental analysis of liquid-metal reactor scram rod kinematic characteristics

    NASA Astrophysics Data System (ADS)

    Konovalenko, F. D.; Kondrashov, S. I.

    2017-01-01

    This article represents the results of computational and experimental research of liquid-metal research reactor control rod kinematics. In this research liquid-metal coolant (sodium) was simulated by water. Investigation of control rod scram-mode movement duration and investigation of velocity of movable parts near the bump of damper are the purposes of this research. Also mathematic simulation of control rod movement in scram mode was performed. Computational results for some modes of water circulation comply with experimental results well. Results of this work will be used for tests of scram rod drive of above-named research reactor. It will significantly simplify the scram rod drive testing stand construction.

  15. Nuclear fuel for liquid metal cooled nuclear reactors

    SciTech Connect

    Duncombe, E.; Adamson, J.; Gratton, C.P.

    1983-11-22

    In a cluster of nuclear fuel rods cooled by liquid metal an obstruction to coolant flow results in overheating in the wake of the obstruction. By the provision of open ended heat transfer tubes in the flow channels, a guaranteed supply of coolant is maintained and this supply holds the temperature to below saturation. Heat transfer via the tubes is highly efficient and ensures that a sufficient temperature rise occurs at the cluster exit to provoke a response from the outlet temperature transducer sensing average temperature.

  16. Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

  17. 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.

  18. 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…

  19. Passive cooling system for top entry liquid metal cooled nuclear reactors

    DOEpatents

    Boardman, Charles E.; Hunsbedt, Anstein; Hui, Marvin M.

    1992-01-01

    A liquid metal cooled nuclear fission reactor plant having a top entry loop joined satellite assembly with a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during shutdown, or heat produced during a mishap. This satellite type reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary cooling system when rendered inoperative.

  20. Liquid metal reactor deactivation as applied to the experimental breeder reactor - II.

    SciTech Connect

    Earle, O. K.; Michelbacher, J. A.; Pfannenstiel, D. F.; Wells, P. B.

    1999-05-28

    The Experimental Breeder Reactor-II (EBR-II) at Argonne National Laboratory-West (ANL-W) was shutdown in September, 1994. This sodium cooled reactor had been in service since 1964, and by the US Department of Energy (DOE) mandate, was to be placed in an industrially and radiologically safe condition for ultimate decommissioning. The deactivation of a liquid metal reactor presents unique concerns. The first major task associated with the project was the removal of all fueled assemblies. In addition, sodium must be drained from systems and processed for ultimate disposal. Residual quantities of sodium remaining in systems must be deactivated or inerted to preclude future hazards associated with pyrophoricity and generation of potentially explosive hydrogen gas. A Sodium Process Facility (SPF) was designed and constructed to react the elemental sodium from the EBR-II primary and secondary systems to sodium hydroxide for disposal. This facility has a design capacity to allow the reaction of the complete inventory of sodium at ANL-W in less than two years. Additional quantities of sodium from the Fermi-1 reactor are also being treated at the SPF.

  1. Method of detecting leakage of reactor core components of liquid metal cooled fast reactors

    DOEpatents

    Holt, Fred E.; Cash, Robert J.; Schenter, Robert E.

    1977-01-01

    A method of detecting the failure of a sealed non-fueled core component of a liquid-metal cooled fast reactor having an inert cover gas. A gas mixture is incorporated in the component which includes Xenon-124; under neutron irradiation, Xenon-124 is converted to radioactive Xenon-125. The cover gas is scanned by a radiation detector. The occurrence of 188 Kev gamma radiation and/or other identifying gamma radiation-energy level indicates the presence of Xenon-125 and therefore leakage of a component. Similarly, Xe-126, which transmutes to Xe-127 and Kr-84, which produces Kr-85.sup.m can be used for detection of leakage. Different components are charged with mixtures including different ratios of isotopes other than Xenon-124. On detection of the identifying radiation, the cover gas is subjected to mass spectroscopic analysis to locate the leaking component.

  2. Investigation of Liquid Metal Embrittlement of Materials for use in Fusion Reactors

    NASA Astrophysics Data System (ADS)

    Kennedy, Daniel; Jaworski, Michael

    2014-10-01

    Liquid metals can provide a continually replenished material for the first wall and extraction blankets of fusion reactors. However, research has shown that solid metal surfaces will experience embrittlement when exposed to liquid metals under stress. Therefore, it is important to understand the changes in structural strength of the solid metal materials and test different surface treatments that can limit embrittlement. Research was conducted to design and build an apparatus for exposing solid metal samples to liquid metal under high stress and temperature. The apparatus design, results of tensile testing, and surface imaging of fractured samples will be presented. This work was supported in part by the U.S. Department of Energy, Office of Science, Office of Workforce Development for Teachers and Scientists (WDTS) under the Science Undergraduate Laboratory Internships Program (SULI).

  3. Liquid metal systems development: reactor vessel support structure evaluation

    SciTech Connect

    McEdwards, J.A.

    1981-01-01

    Results of an evaluation of support structures for the reactor vessel are reported. The U ring, box ring, integral ring, tee ring and tangential beam supports were investigated. The U ring is the recommended vessel support structure configuration.

  4. Sodium leak detection system for liquid metal cooled nuclear reactors

    DOEpatents

    Modarres, Dariush

    1991-01-01

    A light source is projected across the gap between the containment vessel and the reactor vessel. The reflected light is then analyzed with an absorption spectrometer. The presence of any sodium vapor along the optical path results in a change of the optical transmissivity of the media. Since the absorption spectrum of sodium is well known, the light source is chosen such that the sensor is responsive only to the presence of sodium molecules. The optical sensor is designed to be small and require a minimum of amount of change to the reactor containment vessel.

  5. Assessment of the performance potential of the martensitic alloy HT-9 for liquid-metal fast-breeder-reactor applications

    SciTech Connect

    Straalsund, J.L.; Gelles, D.S.

    1983-05-01

    Martensitic stainless steels appear to provide attractive alternatives to austenitic stainless steels for liquid metal fast breeder reactors (LMFBR). The United States National Cladding/Duct (NCD) Materials Development Program has selected Sandvik alloy HT-9 (12CrMoW) as one of six prime candidate alloys for advanced in-core structural materials having very high peak burnup capabilities. The NCD program, since 1974, has been accumulating engineering data for HT-9. Properties include swelling, irradiation creep and microstructure as a function of fluence, postirradiation mechanical properties, thermal creep, sodium compatibility and hardware fabrication technology. Tests results are presented.

  6. 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.

  7. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1995-01-01

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo's structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated.

  8. Passive air cooling of liquid metal-cooled reactor with double vessel leak accommodation capability

    DOEpatents

    Hunsbedt, A.; Boardman, C.E.

    1995-04-11

    A passive and inherent shutdown heat removal method with a backup air flow path which allows decay heat removal following a postulated double vessel leak event in a liquid metal-cooled nuclear reactor is disclosed. The improved reactor design incorporates the following features: (1) isolation capability of the reactor cavity environment in the event that simultaneous leaks develop in both the reactor and containment vessels; (2) a reactor silo liner tank which insulates the concrete silo from the leaked sodium, thereby preserving the silo`s structural integrity; and (3) a second, independent air cooling flow path via tubes submerged in the leaked sodium which will maintain shutdown heat removal after the normal flow path has been isolated. 5 figures.

  9. High Temperature Ultrasonic Transducers for In-Service Inspection of Liquid Metal Fast Reactors

    SciTech Connect

    Griffin, Jeffrey W.; Posakony, Gerald J.; Harris, Robert V.; Baldwin, David L.; Jones, Anthony M.; Bond, Leonard J.

    2011-12-31

    In-service inspection of liquid metal (sodium) fast reactors requires the use of ultrasonic transducers capable of operating at high temperatures (>200°C), high gamma radiation fields, and the chemically reactive liquid sodium environment. In the early- to mid-1970s, the U.S. Atomic Energy Commission supported development of high-temperature, submersible single-element transducers, used for scanning and under-sodium imaging in the Fast Flux Test Facility and the Clinch River Breeder Reactor. Current work is building on this technology to develop the next generation of high-temperature linear ultrasonic transducer arrays for under-sodium viewing and in-service inspections.

  10. Under-Sodium Viewing: A Review of Ultrasonic Imaging Technology for Liquid Metal Fast Reactors

    SciTech Connect

    Griffin, Jeffrey W.; Peters, Timothy J.; Posakony, Gerald J.; Chien, Hual-Te; Bond, Leonard J.; Denslow, Kayte M.; Sheen, Shuh-Haw; Raptis, Paul

    2009-03-27

    This current report is a summary of information obtained in the "Information Capture" task of the U.S. DOE-funded "Under Sodium Viewing (USV) Project." The goal of the multi-year USV project is to design, build, and demonstrate a state-of-the-art prototype ultrasonic viewing system tailored for periodic reactor core in-service monitoring and maintenance inspections. The study seeks to optimize system parameters, improve performance, and re-establish this key technology area which will be required to support any new U.S. liquid-metal cooled fast reactors.

  11. Reactor plasma facing component designs based on liquid metal concepts supported in porous systems

    NASA Astrophysics Data System (ADS)

    Tabarés, F. L.; Oyarzabal, E.; Martin-Rojo, A. B.; Tafalla, D.; de Castro, A.; Soleto, A.

    2017-01-01

    The use of liquid metals (LMs) as plasma facing components in fusion devices was proposed as early as 1970 for a field reversed concept and inertial fusion reactors. The idea was extensively developed during the APEX Project, at the turn of the century, and it is the subject at present of the biennial International Symposium on Lithium Applications (ISLA), whose fourth meeting took place in Granada, Spain at the end of September 2015. While liquid metal flowing concepts were specially addressed in USA research projects, the idea of embedding the metal in a capillary porous system (CPS) was put forwards by Russian teams in the 1990s, thus opening the possibility of static concepts. Since then, many ideas and accompanying experimental tests in fusion devices and laboratories have been produced, involving a large fraction of countries within the international fusion community. Within the EUROFusion Roadmap, these activities are encompassed into the working programs of the plasma facing components (PFC) and divertor tokamak test (DTT) packages. In this paper, a review of the state of the art in concepts based on the CPS set-up for a fusion reactor divertor target, aimed at preventing the ejection of the liquid metal by electro-magnetic (EM) forces generated under plasma operation, is described and required R+D activities on the topic, including ongoing work at CIEMAT specifically oriented to filling the remaining gaps, are stressed.

  12. Investigation of heat transfer in liquid-metal flows under fusion-reactor conditions

    NASA Astrophysics Data System (ADS)

    Poddubnyi, I. I.; Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, V. G.; Sviridov, E. V.; Leshukov, A. Yu.; Aleskovskiy, K. V.; Obukhov, D. M.

    2016-12-01

    The effect discovered in studying a downward liquid-metal flow in vertical pipe and in a channel of rectangular cross section in, respectively, a transverse and a coplanar magnetic field is analyzed. In test blanket modules (TBM), which are prototypes of a blanket for a demonstration fusion reactor (DEMO) and which are intended for experimental investigations at the International Thermonuclear Experimental Reactor (ITER), liquid metals are assumed to fulfil simultaneously the functions of (i) a tritium breeder, (ii) a coolant, and (iii) neutron moderator and multiplier. This approach to testing experimentally design solutions is motivated by plans to employ, in the majority of the currently developed DEMO blanket projects, liquid metals pumped through pipes and/or rectangular channels in a transvers magnetic field. At the present time, experiments that would directly simulate liquid-metal flows under conditions of ITER TBM and/or DEMO blanket operation (irradiation with thermonuclear neutrons, a cyclic temperature regime, and a magnetic-field strength of about 4 to 10 T) are not implementable for want of equipment that could reproduce simultaneously the aforementioned effects exerted by thermonuclear plasmas. This is the reason why use is made of an iterative approach to experimentally estimating the performance of design solutions for liquid-metal channels via simulating one or simultaneously two of the aforementioned factors. Therefore, the investigations reported in the present article are of considerable topical interest. The respective experiments were performed on the basis of the mercury magneto hydrodynamic (MHD) loop that is included in the structure of the MPEI—JIHT MHD experimental facility. Temperature fields were measured under conditions of two- and one-sided heating, and data on averaged-temperature fields, distributions of the wall temperature, and statistical fluctuation features were obtained. A substantial effect of counter thermo gravitational

  13. 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.

  14. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    DOEpatents

    Jones, Robert D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

  15. Diversion analysis and safeguards measures for liquid metal fast breeder reactors

    SciTech Connect

    Persiani, P.J.

    1981-10-01

    The general objective of the study is to perform a diversion analysis and an assessment of the available safeguards methods and systems for verifying inventory and flow of nuclear material in accessible and inaccessible areas of liquid-metal fast breeder reactor, LMFBR, systems. The study focuses primarily on the assembly-handling operations, assembly storage facilities, and reactor operations facilities relating to existing and/or near-term planned experimental, demonstration and prototypal reactor plants. The safeguards systems and methods presented are considered to be feasible for development and for implementation within the resource limitation of the IAEA and are considered to be consistent with the objectives, requirements, and constraints of the IAEA as outlined in the IAEA documents INFCIRC/153 and INFCIRC/66-Rev-2.

  16. Control rod heterogeneity effects in liquid-metal fast breeder reactors: Method developments and experimental validation

    SciTech Connect

    Carta, M.; Granget, G.; Palmiotti, G.; Salvatores, M.; Soule, R.

    1988-11-01

    The control rod worth assessment in a large liquid-metal fast breeder reactor is strongly dependent on the actual arrangement of the absorber pins inside the control rod subassemblies. The so-called heterogeneity effects (i.e., the effects on the rod reactivity of the actual rod internal geometry versus homogenization of the absorber atoms over all the subassembly volume) have been evaluated, using explicit and variational methods to derive appropriate cross sections. An experimental program performed at the MASURCA facility has been used to validate these methods.

  17. Pattern recognition techniques applied to acoustic detection of liquid-metal fast breeder reactor cooling defects

    SciTech Connect

    Brunet, M.; Dubuisson, B.

    1983-08-01

    In the event of a partial or total blockage of a liquid-metal fast breeder reactor core subassembly, a boiling zone may be created. Acoustic signals from such a zone could provide a means of early detection of accident conditions. A three-step method, based on pattern recognition techniques, is described and used to analyze data from three experiments that simulate core cooling fault conditions. This method is shown to be capable of detecting the abnormal situation in each of the experiments analyzed.

  18. Performance of metal and oxide fuel cores during accidents in large liquid-metal-cooled reactors

    SciTech Connect

    Royl, P.H.; Kussmaul, G. ); Cahalan, J.E.; Wigeland, R.A. ); Friedel, G. ); Moreau, J. ); Perks, M. )

    1992-02-01

    This paper reports on a cooperative effort among European and U.S. analysts, which is an assessment of the comparative safety performance of metal and oxide fuels during accidents in a 3500-MW (thermal), pool-type, liquid-metal-cooled reactor (LMR) is performed. The study focuses on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower, and the unprotected loss-of-heat-sink (ULOHS). Core designs with a similar power output that have been previously analyzed in Europe under ULOF accident conditions are also included in this comparison. Emphasis is placed on identification of design features that provide passive, self-limiting responses to postulated accident conditions and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than do oxide-fueled reactors of the same design.

  19. Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

    SciTech Connect

    Cahalan, J.; Wigeland, R. ); Friedel, G. , Bergisch Gladbach ); Kussmaul, G.; Royl, P. ); Moreau, J. ); Perks, M.

    1990-01-01

    In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs.

  20. Time-series investigation of anomalous thermocouple responses in a liquid-metal-cooled reactor

    SciTech Connect

    Gross, K.C.; Planchon, H.P.; Poloncsik, J.

    1988-03-24

    A study was undertaken using SAS software to investigate the origin of anomalous temperature measurements recorded by thermocouples (TCs) in an instrumented fuel assembly in a liquid-metal-cooled nuclear reactor. SAS macros that implement univariate and bivariate spectral decomposition techniques were employed to analyze data recorded during a series of experiments conducted at full reactor power. For each experiment, data from physical sensors in the tests assembly were digitized at a sampling rate of 2/s and recorded on magnetic tapes for subsequent interactive processing with CMS SAS. Results from spectral and cross-correlation analyses led to the identification of a flow rate-dependent electromotive force (EMF) phenomenon as the origin of the anomalous TC readings. Knowledge of the physical mechanism responsible for the discrepant TC signals enabled us to device and justify a simple correction factor to be applied to future readings.

  1. A liquid-metal reactor core demonstration experiment using HT-9

    SciTech Connect

    Bridges, A.E.; Waltar, A.E.; Leggett, R.D.; Baker, R.B. ); Ethridge, J.L. )

    1993-06-01

    The use of the ferritic/martensitic HT-9 alloy as the cladding and duct material for the attainment of the high fuel burnup levels critical to the viability of an economical liquid-metal reactor fuel system. The CDE, a partial core loading of fuel and blanket assemblies in the US Department of Energy's Fast Flux Test Facility, has successfully attained its irradiation exposure goal of 3 yr. Consisting of ten fuel and six blanket assemblies in a heterogeneous core configuration, the CDE has clearly demonstrated the capability of the advanced fuel and blanket designs to attain high burnups and fast fluences. Each CDE fuel assembly consisted of 169 large-diameter fuel pins comprising mixed-oxide annular fuel pellets in sealed HT-9 cladding tubes. Each CDE blanket assembly consisted of 91 large-diameter pins comprising solid depleted uranium dioxide pellets in sealed HT-9 cladding tubes. The maximum-exposure CDE fuel assembly reached a peak pellet burnup of 163,900 MWd/ton metal (M) and a peak fast fluence (E > 0.1 MeV) of 23.3 [times] 10[sup 22] n/cm[sup 2]. The maximum-exposure CDE blanket assembly reached a peak pellet burnup of 43 100 MWd/ton M and a peak fast fluence (E . 0.1 MeV) of 22.8 [times] 10[sup 22] n/cm[sup 2]. Lead test fuel assemblies built to CDE specifications continue their successful irradiation and have attained burnups of > 238,000 MWd/ton M with accumulated fast fluences (E > 0.1 MeV) of > 38 [times] 10[sup 22] n/cm[sup 2]. In-core measurements of HT-9 ducts and withdrawal loads of the assemblies indicate that duct distortion will not be a factor that limits the lifetime of the fuel or blanket assemblies. Comparison of the measured and predicted coolant outlet temperatures from the peak CDE fuel and blanket assemblies indicate the irradiation of the CDE has proceeded as planned. The CDE represents a tremendous success in demonstrating the lifetime capabilities of this advanced oxide system using the HT-9 ferritic alloy for structural materials.

  2. Jacking mechanism for upper internals structure of a liquid metal nuclear reactor

    DOEpatents

    Gillett, James E.; Wineman, Arthur L.

    1984-01-01

    A jacking mechanism for raising the upper internals structure of a liquid metal nuclear reactor which jacking mechanism uses a system of gears and drive shafts to transmit force from a single motor to four mechanically synchronized ball jacks to raise and lower support columns which support the upper internals structure. The support columns have a pin structure which rides up and down in a slot in a housing fixed to the reactor head. The pin has two locking plates which can be rotated around the pin to bring bolt holes through the locking plates into alignment with a set of bolt holes in the housing, there being a set of such housing bolt holes corresponding to both a raised and a lowered position of the support column. When the locking plate is so aligned, a surface of the locking plate mates with a surface in the housing such that the support column is then supported by the locking plate and not by the ball jacks. Since the locking plates are to be installed and bolted to the housing during periods of reactor operation, the ball jacks need not be sized to react the large forces which occur or potentially could occur on the upper internals structure of the reactor during operation. The locking plates react these loads. The ball jacks, used only during refueling, can be smaller, which enable conventionally available equipment to fulfill the precision requirements for the task within available space.

  3. Experimental and design experience with passive safety features of liquid metal reactors

    SciTech Connect

    Lucoff, D.M.; Waltar, A.E.; Sackett, J.I.; Salvatores, M.; Aizawa, K.

    1992-10-01

    Liquid metal cooled reactors (LMRs) have already been demonstrated to be robust machines. Many reactor designers now believe that it is possible to include in this technology sufficient passive safety that LMRs would be able to survive loss of flow, loss of heat sink, and transient overpower events, even if the plant protective system fails completely and do so without damage to the core. Early whole-core testing in Rapsodie, EBR-II. and FFTF indicate such designs may be possible. The operational safety testing program in EBR-II is demonstrating benign response of the reactor to a full range of controls failures. But additional testing is needed if transient core structural response under major accident conditions is to be properly understood. The proposed international Phase IIB passive safety tests in FFTF, being designed with a particular emphasis on providing, data to understand core bowing extremes, and further tests planned in EBR-II with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions.

  4. The role of SASSYS-1 in LMR (Liquid Metal Reactor) safety analysis

    SciTech Connect

    Dunn, F.E.; Wei, T.Y.C.

    1988-01-01

    The SASSYS-1 liquid metal reactor systems analysis computer code is currently being used as the principal tool for analysis of reactor plant transients in LMR development projects. These include the IFR and EBR-II Projects at Argonne National Laboratory, the FFTF project at Westinghouse-Hanford, the PRISM project at General Electric, the SAFR project at Rockwell International, and the LSPB project at EPRI. The SASSYS-1 code features a multiple-channel thermal-hydraulics core representation coupled with a point kinetics neutronics model with reactivity feedback, all combined with detailed one-dimensional thermal-hydraulic models of the primary and intermediate heat transport systems, including pipes, pumps, plena, valves, heat exchangers and steam generators. In addition, SASSYS-1 contains detailed models for active and passive shutdown and emergency heat rejection systems and a generalized plant control system model. With these models, SASSYS-1 provides the capability to analyze a wide range of transients, including normal operational transients, shutdown heat removal transients, and anticipated transients without scram events. 26 refs., 16 figs.

  5. Forced-to-natural convection transition tests in parallel simulated liquid metal reactor fuel assemblies

    SciTech Connect

    Levin, A.E. ); Montgomery, B.H. )

    1990-01-01

    The Thermal-Hydraulic Out of Reactor Safety (THORS) Program at Oak Ridge National Laboratory (ORNL) had as its objective the testing of simulated, electrically heated liquid metal reactor (LMR) fuel assemblies in an engineering-scale, sodium loop. Between 1971 and 1985, the THORS Program operated 11 simulated fuel bundles in conditions covering a wide range of normal and off-normal conditions. The last test series in the Program, THORS-SHRS Assembly 1, employed two parallel, 19-pin, full-length, simulated fuel assemblies of a design consistent with the large LMR (Large Scale Prototype Breeder -- LSPB) under development at that time. These bundles were installed in the THORS Facility, allowing single- and parallel-bundle testing in thermal-hydraulic conditions up to and including sodium boiling and dryout. As the name SHRS (Shutdown Heat Removal System) implies, a major objective of the program was testing under conditions expected during low-power reactor operation, including low-flow forced convection, natural convection, and forced-to-natural convection transition at various powers. The THORS-SHRS Assembly 1 experimental program was divided up into four phases. Phase 1 included preliminary and shakedown tests, including the collection of baseline steady-state thermal-hydraulic data. Phase 2 comprised natural convection testing. Forced convection testing was conducted in Phase 3. The final phase of testing included forced-to-natural convection transition tests. Phases 1, 2, and 3 have been discussed in previous papers. The fourth phase is described in this paper. 3 refs., 2 figs.

  6. Diversion analysis and safeguards measures for liquid-metal fast-breeder reactors. Paper IAEA-SM-260/6

    SciTech Connect

    Persiani, P.J.; Ermakov, S.V.

    1982-01-01

    The objective of the study is to perform a diversion analysis and an assessment of the available safeguards methods and systems for verifying inventory and flow of nuclear material in accessible and inaccessible areas of liquid-metal fast breeder reactor, LMFBR, systems. The study focuses primarily on the assembly-handling operations, assembly storage facilities, and reactor operations facilities relating to experimental, demonstration and prototypal reactor plants. The safeguards systems and methods presented are considered to be feasible for development and implementation within the resource limitation of the IAEA and are considered to be consistent with the objectives, requirements, and constraints of the IAEA documents INFCIRC/153 and INFCIRC/66-Rev-2.

  7. The thermodynamics of pyrochemical processes for liquid metal reactor fuel cycles

    SciTech Connect

    Johnson, I.

    1987-01-01

    The thermodynamic basis for pyrochemical processes for the recovery and purification of fuel for the liquid metal reactor fuel cycle is described. These processes involve the transport of the uranium and plutonium from one liquid alloy to another through a molten salt. The processes discussed use liquid alloys of cadmium, zinc, and magnesium and molten chloride salts. The oxidation-reduction steps are done either chemically by the use of an auxiliary redox couple or electrochemically by the use of an external electrical supply. The same basic thermodynamics apply to both the salt transport and the electrotransport processes. Large deviations from ideal solution behavior of the actinides and lanthanides in the liquid alloys have a major influence on the solubilities and the performance of both the salt transport and electrotransport processes. Separation of plutonium and uranium from each other and decontamination from the more noble fission product elements can be achieved using both transport processes. The thermodynamic analysis is used to make process design computations for different process conditions.

  8. Experiments on liquid-metal fast breeder reactor aerosol source terms after severe accidents

    SciTech Connect

    Berthoud, G.; Longest, A.W.; Wright, A.L.; Schutz, W.P.

    1988-05-01

    In the extremely unlikely event of a liquid-metal fast breeder reactor core disruptive accident, expanding core material or sodium vapor inside the sodium pool may cause leaks in the vessel head and transport of radioactive material, mostly aerosols, in one large bubble or several smaller bubbles under energetic conditions to the cover gas and through leaks to the inner containment (''instantaneous source term''). Out-of-pile experiments on bubble expansion from a pressurized source inside a liquid (water or sodium) and related phenomena like heat transfer, condensation, entrainment, rise, and aerosol transport were carried out in France and the United States and are continuing in the Federal Republic of Germany. Parameters and results of these experiments are described and discussed, mainly concerning the aerosol problem. It appears that several mechanisms exist for a very efficient removal of particles from the bubble. Retention factors larger than 10,000 were found in most cases. In addition, a short survey is given of French and German experiments on fuel and fission product release from evaporating or burning sodium pools (delayed source term).

  9. 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.

  10. Effect of yttrium additions on void swelling in Liquid Metal Fast Breeder Reactor candidate cladding alloys

    SciTech Connect

    Hopson, R.D.

    1981-10-01

    Candidate Liquid Metal Fast Breeder Reactor cladding alloys AL1 (Fe-26% Ni-9% Cr) and AL2 (Fe-35% Ni-12% Cr) without and with the addition of 0.1% yttrium were bombarded by 4 MeV/sup 56/Fe/sup 2 +/ ions without and with simultaneous bombardment by 0.4 MeV /sup 4/He/sup +/ ions. These bombardments were conducted at various irradiation temperatures to determine the effect of yttrium on void swelling. The addition of yttrium decreased peak swelling for 4 MeV /sup 56/Fe/sup 2 +/ ion bombarded AL1 and AL2 by 28% and 20%, respectively. In all cases where similar sample comparisons were made (i.e., undoped with undoped and doped with doped) and where bombardment conditions were similar (i.e., single with single beam and dual with dual beam), AL1 showed less peak swelling than did AL2. Simultaneously implanting helium during heavy-ion bombardment increased peak swelling in undoped and doped AL1 by factors of 2.3 and 2.6, respectively.

  11. Eddy current position indicating apparatus for measuring displacements of core components of a liquid metal nuclear reactor

    DOEpatents

    Day, Clifford K.; Stringer, James L.

    1977-01-01

    Apparatus for measuring displacements of core components of a liquid metal fast breeder reactor by means of an eddy current probe. The active portion of the probe is located within a dry thimble which is supported on a stationary portion of the reactor core support structure. Split rings of metal, having a resistivity significantly different than sodium, are fixedly mounted on the core component to be monitored. The split rings are slidably positioned around, concentric with the probe and symmetrically situated along the axis of the probe so that motion of the ring along the axis of the probe produces a proportional change in the probes electrical output.

  12. 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

  13. Bellows-Type Accumulators for Liquid Metal Loops of Space Reactor Power Systems

    NASA Astrophysics Data System (ADS)

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2006-01-01

    In many space nuclear power systems, the primary and/or secondary loops use liquid metal working fluids, and require accumulators to accommodate the change in the liquid metal volume and maintain sufficient subcooling to avoid boiling. This paper developed redundant and light-weight bellows-type accumulators with and without a mechanical spring, and compared the operating condition and mass of the accumulators for different types of liquid metal working fluids and operating temperatures: potassium, NaK-78, sodium and lithium loops of a total capacity of 50 liters and nominal operating temperatures of 840 K, 860 K, 950 K and 1340 K, respectively. The effects of using a mechanical spring and different structural materials on the design, operation and mass of the accumulators are also investigated. The structure materials considered include SS-316, Hastelloy-X, C-103 and Mo-14Re. The accumulator without a mechanical spring weighs 23 kg and 40 kg for a coolant subcooling of 50 K and 100 K, respectively, following a loss of the fill gas. The addition of a mechanical spring comes with a mass penalty, in favor of higher redundancy and maintaining a higher liquid metal subcooling.

  14. Bellows-Type Accumulators for Liquid Metal Loops of Space Reactor Power Systems

    SciTech Connect

    Tournier, Jean-Michel; El-Genk, Mohamed S.

    2006-01-20

    In many space nuclear power systems, the primary and/or secondary loops use liquid metal working fluids, and require accumulators to accommodate the change in the liquid metal volume and maintain sufficient subcooling to avoid boiling. This paper developed redundant and light-weight bellows-type accumulators with and without a mechanical spring, and compared the operating condition and mass of the accumulators for different types of liquid metal working fluids and operating temperatures: potassium, NaK-78, sodium and lithium loops of a total capacity of 50 liters and nominal operating temperatures of 840 K, 860 K, 950 K and 1340 K, respectively. The effects of using a mechanical spring and different structural materials on the design, operation and mass of the accumulators are also investigated. The structure materials considered include SS-316, Hastelloy-X, C-103 and Mo-14Re. The accumulator without a mechanical spring weighs 23 kg and 40 kg for a coolant subcooling of 50 K and 100 K, respectively, following a loss of the fill gas. The addition of a mechanical spring comes with a mass penalty, in favor of higher redundancy and maintaining a higher liquid metal subcooling.

  15. SAS4A: A computer model for the analysis of hypothetical core disruptive accidents in liquid metal reactors

    SciTech Connect

    Tentner, A.M.; Birgersson, G.; Cahalan, J.E.; Dunn, F.E.; Kalimullah; Miles, K.J.

    1987-01-01

    To ensure that the public health and safety are protected under any accident conditions in a Liquid Metal Fast Breeder Reactor (LMFBR), many accidents are analyzed for their potential consequences. The SAS4A code system, described in this paper, provides such an analysis capability, including the ability to analyze low probability events such as the Hypothetical Core Disruptive Accidents (HCDAs). The SAS4A code system has been designed to simulate all the events that occur in a LMFBR core during the initiating phase of a Hypothetical Core Disruptive Accident. During such postulated accident scenarios as the Loss-of-Flow and Transient Overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling and fuel and cladding melting and relocation. During to the strong neutronic feedback present in a nuclear reactor, these events can significantly influence the reactor power. The SAS4A code system is used in the safety analysis of nuclear reactors, in order to estimate the energetic potential of very low probability accidents. The results of SAS4A simulations are also used by reactor designers in order to build safer reactors and eliminate the possibility of any accident which could endanger the public safety.

  16. Advanced Test Reactor Tour

    SciTech Connect

    Miley, Don

    2011-01-01

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  17. Advanced Test Reactor Tour

    ScienceCinema

    Miley, Don

    2016-07-12

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  18. Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

    SciTech Connect

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

    1994-02-01

    This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

  19. SCoRe - Concepts of Liquid Metal Cooled Space Reactors for Avoidance of Single-Point Failure

    SciTech Connect

    El-Genk, Mohamed; Hatton, Steven; Fox, Charles; Tournier, Jean-Michel

    2005-02-06

    Space nuclear Reactor Power Systems (SRPSs) are being developed to meet electrical power requirements for NASA's planetary exploration missions early next decade. In addition to enjoying some degree of autonomy, these systems need to operate reliably through the end of the mission, which could not be realized solely through a redundancy in the reactor's coolant loop. Besides increasing the total system mass, such hardware redundancy does not eliminate a single-point failure in the reactor and subsequent loss of coolant. This paper presents three concepts of the liquid metal cooled. Sectored, Compact Reactor (SCoRe) for the avoidance of single-point failure. The SCoRe-S, ScoRe-M, and SCoRe-L concepts are for small, medium, and large reactor cores, covering a wide range of electrical power requirements, from 10's of kWe to a few MWe. As a common feature in all SCoRe concepts, the reactor core is divided into six sectors that are neutronically coupled but thermal-hydraulically decoupled. The dividers of the sectors are liquid metal heat pipes, which facilitate cooling a sector experiencing a Loss of Coolant (LOC) by passively transporting the fission power generated in it to the two adjacent sectors without losing the mission. At the same time, the fission power of the reactor is reduced to avoid overheating the fuel in the sector experiencing a LOC. The SCoRe concepts have compact, hexagonal cores surrounded by a relatively thick (10 cm minimum) BeO reflector and axial BeO reflector that is 4 cm thick. The SCoRe is placed directly in front of the radiation shield, thus reducing the shield mass and that of the power system. In SCoRe-S cores, the UN fuel pins are arranged in a triangular lattice while in the SCoRe-M and SCoRe-L cores, the UN fuel pins arranged in a triangular lattice are assembled in 19-pin and 37-pin shrouded bundles, respectively.

  20. Modeling of Flow Blockage in a Liquid Metal-Cooled Reactor Subassembly with a Subchannel Analysis Code

    SciTech Connect

    Jeong, Hae-Yong; Ha, Kwi-Seok; Chang, Won-Pyo; Kwon, Young-Min; Lee, Yong-Bum

    2005-01-15

    The local blockage in a subassembly of a liquid metal-cooled reactor (LMR) is of importance to the plant safety because of the compact design and the high power density of the core. To analyze the thermal-hydraulic parameters in a subassembly of a liquid metal-cooled reactor with a flow blockage, the Korea Atomic Energy Research Institute has developed the MATRA-LMR-FB code. This code uses the distributed resistance model to describe the sweeping flow formed by the wire wrap around the fuel rods and to model the recirculation flow after a blockage. The hybrid difference scheme is also adopted for the description of the convective terms in the recirculating wake region of low velocity. Some state-of-the-art turbulent mixing models were implemented in the code, and the models suggested by Rehme and by Zhukov are analyzed and found to be appropriate for the description of the flow blockage in an LMR subassembly. The MATRA-LMR-FB code predicts accurately the experimental data of the Oak Ridge National Laboratory 19-pin bundle with a blockage for both the high-flow and low-flow conditions. The influences of the distributed resistance model, the hybrid difference method, and the turbulent mixing models are evaluated step by step with the experimental data. The appropriateness of the models also has been evaluated through a comparison with the results from the COMMIX code calculation. The flow blockage for the KALIMER design has been analyzed with the MATRA-LMR-FB code and is compared with the SABRE code to guarantee the design safety for the flow blockage.

  1. Liquid metal hydrogen barriers

    DOEpatents

    Grover, George M.; Frank, Thurman G.; Keddy, Edward S.

    1976-01-01

    Hydrogen barriers which comprise liquid metals in which the solubility of hydrogen is low and which have good thermal conductivities at operating temperatures of interest. Such barriers are useful in nuclear fuel elements containing a metal hydride moderator which has a substantial hydrogen dissociation pressure at reactor operating temperatures.

  2. Uncertainty quantification approaches for advanced reactor analyses.

    SciTech Connect

    Briggs, L. L.; Nuclear Engineering Division

    2009-03-24

    The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

  3. 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.

  4. Development of remote disassembly technology for liquid-metal reactor (LMR) fuel

    SciTech Connect

    Bradley, E.C.; Evans, J.H.; Metz, C.F. III; Weil, B.S.

    1990-01-01

    A major objective of the Consolidated Fuel Reprocessing Program (CFRP) is to develop equipment and demonstrate technology to reprocess fast breeder reactor fuel. Experimental work on fuel disassembly cutting methods began in the 1970s. High-power laser cutting was selected as the preferred cutting method for fuel disassembly. Remotely operated development equipment was designed, fabricated, installed, and tested at Oak Ridge National Laboratory (ORNL). Development testing included remote automatic operation, remote maintenance testing, and laser cutting process development. This paper summarizes the development work performed at ORNL on remote fuel disassembly. 2 refs., 1 fig.

  5. Failed fuel monitoring and surveillance techniques for liquid metal cooled fast reactors

    SciTech Connect

    Lambert, J.D.B.; Mikaili, R.; Gross, K.C.; Strain, R.V.; Aoyama, T.; Ukai, S.; Nomura, S.; Nakae, N.

    1995-05-01

    The Experimental Breeder Reactor II (EBR-II) has been used as a facility for irradiation of LMR fuels and components for thirty years. During this time many tests of experimental fuel were continued to cladding breach in order to study modes of element failure; the methods used to identify such failures are described in a parallel paper. This paper summarizes experience of monitoring the delayed-neutron (DN) and fission-gas (FG) release behavior of a smaller number of elements that continued operation in the run-beyond-cladding-breach (RBCB) mode. The scope of RBCB testing, the methods developed to characterize failures on-line, and examples of DN/FG behavior are described.

  6. Liquid metal enabled pump

    PubMed Central

    Tang, Shi-Yang; Khoshmanesh, Khashayar; Sivan, Vijay; Petersen, Phred; O’Mullane, Anthony P.; Abbott, Derek; Mitchell, Arnan; Kalantar-zadeh, Kourosh

    2014-01-01

    Small-scale pumps will be the heartbeat of many future micro/nanoscale platforms. However, the integration of small-scale pumps is presently hampered by limited flow rate with respect to the input power, and their rather complicated fabrication processes. These issues arise as many conventional pumping effects require intricate moving elements. Here, we demonstrate a system that we call the liquid metal enabled pump, for driving a range of liquids without mechanical moving parts, upon the application of modest electric field. This pump incorporates a droplet of liquid metal, which induces liquid flow at high flow rates, yet with exceptionally low power consumption by electrowetting/deelectrowetting at the metal surface. We present theory explaining this pumping mechanism and show that the operation is fundamentally different from other existing pumps. The presented liquid metal enabled pump is both efficient and simple, and thus has the potential to fundamentally advance the field of microfluidics. PMID:24550485

  7. Feasibility study on ultralong-cycle operation and material performance for compact liquid metal-cooled fast reactors: a review work

    SciTech Connect

    Tak, Taewoo; Choe, Jiwon; Jeong, Yongjin; Lee, Deokjung; Kim, T. K.; Hong, Ser Gi

    2015-11-01

    This paper reviews the feasibility of ultralong-cycle operation on a compact liquid metal-cooled fast reactor (LMR) firstly by assessing the operation of a long-life fast reactor core and secondly by evaluating material performance in respect to both long-cycle operation and compact-size fast reactor. Many kinds of reactor concepts have been proposed, and LMR and small modular reactor (SMR) are the issued leading technologies for generation four (Gen-IV) reactor system development. The breed-and-burn strategy was proposed as a core burning strategy to operate a long cycle, and it has been evaluated in this paper with two reactor concepts: constant axial shape of neutron flux, nuclide densities, and power shape during life of energy and ultralong cycle fast reactor. In addition, Super-Safe, Small, and Simple and small modular fast reactor, compact LMR concepts, have been simulated to evaluate their long-life operation strategies. For the other practical issues, the materials for fuel, coolant, and structure have been identified and some of them are selected to have their performance optimized specifically for compact LMR with a long-cycle operation. It is believed that this comprehensive review will propose a proper direction for future reactor development and will be followed by the next step research for a complete reactor model with the other reactor components.

  8. 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.

  9. Polarized advanced fuel reactors

    SciTech Connect

    Kulsrud, R.M.

    1987-07-01

    The d-/sup 3/He reaction has the same spin dependence as the d-t reaction. It produces no neutrons, so that if the d-d reactivity could be reduced, it would lead to a neutron-lean reactor. The current understanding of the possible suppression of the d-d reactivity by spin polarization is discussed. The question as to whether a suppression is possible is still unresolved. Other advanced fuel reactions are briefly discussed. 11 refs.

  10. LIQUID METAL COMPOSITIONS CONTAINING URANIUM

    DOEpatents

    Teitel, R.J.

    1959-04-21

    Liquid metal compositions containing a solid uranium compound dispersed therein is described. Uranium combines with tin to form the intermetallic compound USn/sub 3/. It has been found that this compound may be incorporated into a liquid bath containing bismuth and lead-bismuth components, if a relatively small percentage of tin is also included in the bath. The composition has a low thermal neutron cross section which makes it suitable for use in a liquid metal fueled nuclear reactor.

  11. Development of an Inspection System for the Reactor Vessel/Containment Vessel of the PRISM and SAFR Liquid Metal Reactors

    SciTech Connect

    1989-02-01

    The integrity of the reactor vessel is of utmost importance in both the PRISM and SAFR concepts. The reactor vessel operates at elevated temperatures and contains molten liquid sodium. To ensure safe operation of the reactor, a periodic, visual inspection of the walls of the containment vessel is required by ASME specifications. This inspection would be conducted during a time when the reactor is shut down for refueling or maintenance. Nuclear Systems Associates, Inc. (NSA) was issued a PRDA contract by the Department of Energy to design, develop, and test a Closed Circuit Television (CCTV) camera system. The purpose of the system is to inspect the welds and wall surface of the Reactor Vessel/Container Vessel for both the PRISM and SAFR type reactors. The system was designed to function at the reactor's normal shutdown temperature, and provide a clear indication of flaws in the wall's weld seams and any cracks that might develop. The project was performed in three phases. The first phase concentrated the efforts on producing a compact camera system with the required resolution, self -contained lighting, and remote control focus and viewing angle. The proposed camera was then tested in a vessel mock-up and found to perform to required specifications at room (cold) temperatures. Simulated flaws, cracks, and a sodium leak were observed with required clarity on both a commercial and blackened stainless steel surfaces. The camera was tested with a single clear glass dome, a single coated glass dome, and a dual-glass dome covering the camera lens and mirror. The second phase of the project was conducted in two parts. The first part involved testing the vessel mock-up at elevated temperatures to verify that the required temperatures can be obtained. The mock-up was constructed with imbedded heaters and both control and indicating thermocouples. Stable operating temperatures over 400°F were achieved. During the second part of this phase, the camera was inserted into the

  12. 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.

  13. A wall-crawling robot for reactor vessel inspection in advanced reactors

    SciTech Connect

    Spelt, P.F.; Crane, C.; Feng, L.; Abidi, M.; Tosunoglu, S.

    1994-06-01

    A consortium of four universities and the Center for Engineering Systems Advanced Research of the Oak Ridge National Laboratory has designed a prototype wall-crawling robot to perform weld inspection in advanced nuclear reactors. Design efforts for the reactor vessel inspection robot (RVIR) concentrated on the Advanced Liquid Metal Reactor because it presents the most demanding environment in which such a robot must operate. The RVIR consists of a chassis containing two sets of suction cups that can alternately grasp the side of the vessel being inspected, providing both locomotion and steering functions. Sensors include three CCD cameras and a weld inspection device based on new shear-wave technology. The restrictions of the inspection environment presented major challenges to the team. These challenges were met in the prototype, which has been tested in a non-radiation, room-temperature mockup of the robot work environment and shown to perform as expected.

  14. Actively convected liquid metal divertor

    NASA Astrophysics Data System (ADS)

    Shimada, Michiya; Hirooka, Yoshi

    2014-12-01

    The use of actively convected liquid metals with j × B force is proposed to facilitate heat handling by the divertor, a challenging issue associated with magnetic fusion experiments such as ITER. This issue will be aggravated even more for DEMO and power reactors because the divertor heat load will be significantly higher and yet the use of copper would not be allowed as the heat sink material. Instead, reduced activation ferritic/martensitic steel alloys with heat conductivities substantially lower than that of copper, will be used as the structural materials. The present proposal is to fill the lower part of the vacuum vessel with liquid metals with relatively low melting points and low chemical activities including Ga and Sn. The divertor modules, equipped with electrodes and cooling tubes, are immersed in the liquid metal. The electrode, placed in the middle of the liquid metal, can be biased positively or negatively with respect to the module. The j × B force due to the current between the electrode and the module provides a rotating motion for the liquid metal around the electrodes. The rise in liquid temperature at the separatrix hit point can be maintained at acceptable levels from the operation point of view. As the rotation speed increases, the current in the liquid metal is expected to decrease due to the v × B electromotive force. This rotating motion in the poloidal plane will reduce the divertor heat load significantly. Another important benefit of the convected liquid metal divertor is the fast recovery from unmitigated disruptions. Also, the liquid metal divertor concept eliminates the erosion problem.

  15. Heat-Powered Pump for Liquid Metals

    NASA Technical Reports Server (NTRS)

    Campana, R. J.

    1986-01-01

    Proposed thermoelectromagnetic pump for liquid metal powered by waste heat; needs no battery, generator, or other external energy source. Pump turns part of heat in liquid metal into pumping energy. In combination with primary pump or on its own, thermoelectric pump circulates coolant between reactor and radiator. As long as there is decay heat to be removed, unit performs function.

  16. Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

    NASA Astrophysics Data System (ADS)

    Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

    2017-01-01

    In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

  17. Metallic fuels for advanced reactors

    NASA Astrophysics Data System (ADS)

    Carmack, W. J.; Porter, D. L.; Chang, Y. I.; Hayes, S. L.; Meyer, M. K.; Burkes, D. E.; Lee, C. B.; Mizuno, T.; Delage, F.; Somers, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor Program, the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. This paper presents an evaluation of metallic alloy fuels. Early US fast reactor developers originally favored metal alloy fuel due to its high fissile density and compatibility with sodium. The goal of fast reactor fuel development programs is to develop and qualify a nuclear fuel system that performs all of the functions of a conventional fast spectrum nuclear fuel while destroying recycled actinides. This will provide a mechanism for closure of the nuclear fuel cycle. Metal fuels are candidates for this application, based on documented performance of metallic fast reactor fuels and the early results of tests currently being conducted in US and international transmutation fuel development programs.

  18. MHD Stability Analysis and Flow Controls of Liquid Metal Free Surface Film Flows as Fusion Reactor PFCs

    NASA Astrophysics Data System (ADS)

    Zhang, Xiujie; Pan, Chuanjie; Xu, Zengyu

    2016-12-01

    Numerical and experimental investigation results on the magnetohydrodynamics (MHD) film flows along flat and curved bottom surfaces are summarized in this study. A simplified modeling has been developed to study the liquid metal MHD film state, which has been validated by the existing experimental results. Numerical results on how the inlet velocity (V), the chute width (W) and the inlet film thickness (d0) affect the MHD film flow state are obtained. MHD stability analysis results are also provided in this study. The results show that strong magnetic fields make the stable V decrease several times compared to the case with no magnetic field, especially small radial magnetic fields (Bn) will have a significant impact on the MHD film flow state. Based on the above numerical and MHD stability analysis results flow control methods are proposed for flat and curved MHD film flows. For curved film flow we firstly proposed a new multi-layers MHD film flow system with a solid metal mesh to get the stable MHD film flows along the curved bottom surface. Experiments on flat and curved MHD film flows are also carried out and some firstly observed results are achieved. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2014GB125003 and 2013GB114002), National Natural Science Foundation of China (No. 11105044)

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

    SciTech Connect

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

    1980-01-01

    Variable width ribbon heating elements have been fabricated which provide a chopped cosine, variable heat flux profile for fuel rod 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 result in the derivation of an analytical expression for the ribbon contours. From this, the ribbons are machined and wound on numerically controlled equipment. Postprocessing and inspection results in a wound, variable width ribbon with the precise dimensional, electrical, and mechanical properties needed for use in fuel pin simulators.

  20. Liquid metal corrosion considerations in alloy development

    SciTech Connect

    Tortorelli, P.F.; DeVan, J.H.

    1984-01-01

    Liquid metal corrosion can be an important consideration in developing alloys for fusion and fast breeder reactors and other applications. Because of the many different forms of liquid metal corrosion (dissolution, alloying, carbon transfer, etc.), alloy optimization based on corrosion resistance depends on a number of factors such as the application temperatures, the particular liquid metal, and the level and nature of impurities in the liquid and solid metals. The present paper reviews the various forms of corrosion by lithium, lead, and sodium and indicates how such corrosion reactions can influence the alloy development process.

  1. 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.

  2. Liquid-metal-cooled, curved-crystal monochromator for Advanced Photon Source bending-magnet beamline 1-BM

    SciTech Connect

    Brauer, S.; Rodricks, B.; Assoufid, L.; Beno, M.A.; Knapp, G.S.

    1996-06-01

    The authors describe a horizontally focusing curved-crystal monochromator that invokes a 4-point bending scheme and a liquid-metal cooling bath. The device has been designed for dispersive diffraction and spectroscopy in the 5--20 keV range, with a predicted focal spot size of {le} 100 {micro}m. To minimize thermal distortions and thermal equilibration time, the 355 x 32 x 0.8 mm crystal will be nearly half submerged in a bath of Ga-In-Sn-Zn alloy. The liquid metal thermally couples the crystal to the water-cooled Cu frame, while permitting the required crystal bending. Calculated thermal profiles and anticipated focusing properties are discussed.

  3. RVACS/RACS (reactor vessel auxiliary cooling system/reactor air cooling system) shutdown heat removal in a modular sized LMR (liquid metal reactor)

    SciTech Connect

    Dunn, F.E.; Wigeland, R.A.; Lo, R.K.

    1988-01-01

    Shutdown heat removal by a RVACS for an unprotected loss of flow case in a modular sized LMR has been analyzed with the SASSYS-1 LMR systems analysis code. For this case it was assumed that all power was lost to the primary and intermediate sodium pumps, and feedwater flow to the steam generators was lost. The control rods failed to scram, but reactivity feedback shut down the power to decay heat levels. The only heat removal was by sodium natural circulation from the core to the vessel wall and by cooling of the vessel wall by radiation and air natural circulation in the Reactor Air Cooling System. The case was run until the system temperatures peaked when the decay heat power level dropped below the heat removal rate.

  4. Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants

    SciTech Connect

    Not Available

    1993-05-13

    Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

  5. Power reactivity decrement components of a homogeneous UPu10Zr-fueled 900-MW(thermal) liquid metal reactor

    SciTech Connect

    Meneghetti, D.; Kucera, D.A.

    1989-01-01

    Linear and Doppler feedback components of the power reactivity decrement (PRD) for a 900-MW(thermal) homogeneous UPu10Zr-fueled sodium-cooled reactor have been calculated. (The PRD is the negative of the reactivity required to bring the reactor from a zero-power hot-critical condition to a given power level.) The components are further separated into power-dependent and power-to-flow-dependent parts. These delineations enhance understanding of the contributions of the components to the feedback process. The delineation also enables the PRDs for other values of coolant flows to be estimated. The linear and Doppler components of the PRD are obtained using the EBRPOCO code, which calculates detailed axially delineated contributions of the components for every subassembly of a loading configuration. Separation of the components into power and power-to-flow parts is made by calculations of the components, assuming infinite thermal conductivities to obtain the power-to-flow values. Subtractions of these from the corresponding PRD quantities give the power-dependent parts. The values of the various feedback components are compared with corresponding quantities reported for an analogous U10Zr-fueled case.

  6. Metal fires and their implications for advanced reactors.

    SciTech Connect

    Nowlen, Steven Patrick; Figueroa, Victor G.; Olivier, Tara Jean; Hewson, John C.; Blanchat, Thomas K.

    2010-10-01

    This report details the primary results of the Laboratory Directed Research and Development project (LDRD 08-0857) Metal Fires and Their Implications for Advance Reactors. Advanced reactors may employ liquid metal coolants, typically sodium, because of their many desirable qualities. This project addressed some of the significant challenges associated with the use of liquid metal coolants, primary among these being the extremely rapid oxidation (combustion) that occurs at the high operating temperatures in reactors. The project has identified a number of areas for which gaps existed in knowledge pertinent to reactor safety analyses. Experimental and analysis capabilities were developed in these areas to varying degrees. In conjunction with team participation in a DOE gap analysis panel, focus was on the oxidation of spilled sodium on thermally massive surfaces. These are spills onto surfaces that substantially cool the sodium during the oxidation process, and they are relevant because standard risk mitigation procedures seek to move spill environments into this regime through rapid draining of spilled sodium. While the spilled sodium is not quenched, the burning mode is different in that there is a transition to a smoldering mode that has not been comprehensively described previously. Prior work has described spilled sodium as a pool fire, but there is a crucial, experimentally-observed transition to a smoldering mode of oxidation. A series of experimental measurements have comprehensively described the thermal evolution of this type of sodium fire for the first time. A new physics-based model has been developed that also predicts the thermal evolution of this type of sodium fire for the first time. The model introduces smoldering oxidation through porous oxide layers to go beyond traditional pool fire analyses that have been carried out previously in order to predict experimentally observed trends. Combined, these developments add significantly to the safety

  7. Prognostics Health Management for Advanced Small Modular Reactor Passive Components

    SciTech Connect

    Meyer, Ryan M.; Ramuhalli, Pradeep; Coble, Jamie B.; Mitchell, Mark R.; Wootan, David W.; Hirt, Evelyn H.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-10-18

    In the United States, sustainable nuclear power to promote energy security is a key national energy priority. Advanced small modular reactors (AdvSMR), which are based on modularization of advanced reactor concepts using non-light-water reactor (LWR) coolants such as liquid metal, helium, or liquid salt may provide a longer-term alternative to more conventional LWR-based concepts. The economics of AdvSMRs will be impacted by the reduced economy-of-scale savings when compared to traditional LWRs and the controllable day-to-day costs of AdvSMRs are expected to be dominated by operations and maintenance costs. Therefore, achieving the full benefits of AdvSMR deployment requires a new paradigm for plant design and management. In this context, prognostic health management of passive components in AdvSMRs can play a key role in enabling the economic deployment of AdvSMRs. In this paper, the background of AdvSMRs is discussed from which requirements for PHM systems are derived. The particle filter technique is proposed as a prognostics framework for AdvSMR passive components and the suitability of the particle filter technique is illustrated by using it to forecast thermal creep degradation using a physics-of-failure model and based on a combination of types of measurements conceived for passive AdvSMR components.

  8. Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.

    SciTech Connect

    Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage

  9. Liquid metal electric pump

    DOEpatents

    Abbin, Joseph P.; Andraka, Charles E.; Lukens, Laurance L.; Moreno, James B.

    1992-01-01

    An electrical pump for pumping liquid metals to high pressures in high temperature environments without the use of magnets or moving mechanical parts. The pump employs a non-porous solid electrolyte membrane, typically ceramic, specific to the liquid metal to be pumped. A DC voltage is applied across the thickness of the membrane causing ions to form and enter the membrane on the electrically positive surface, with the ions being neutralized on the opposite surface. This action provides pumping of the liquid metal from one side of the non-porous solid electrolyte membrane to the other.

  10. Advanced liquid metal reactor fuel and blanket designs using HT9

    SciTech Connect

    Bridges, A.E.; Waltar, A.E.; Leggett, R.D.; Baker, R.B.; Gneiting, B.C.

    1991-08-01

    This paper discusses the results of the Core Demonstration Experiment being irradiated in the US Department of Energy's Fast Flux Test Facility. The CDE clearly demonstrates that mixed-oxide fuel can achieve burnups in excess of 200 MWd/kgM and fast fluences in excess of 30 {times} 10{sup 22} n/cm{sup 2} using the very low swelling ferritic-martensitic alloy, HT9. Supporting data from post-irradiation examination of the ACO-1 experiment, a related fuel test for the CDE, is reported and compared to the existing austenitic database. Additionally, the current status of a follow-on program to test metal fuel using HT9 is reviewed. 22 refs., 6 figs.

  11. Liquid metal enabled microfluidics.

    PubMed

    Khoshmanesh, Khashayar; Tang, Shi-Yang; Zhu, Jiu Yang; Schaefer, Samira; Mitchell, Arnan; Kalantar-Zadeh, Kourosh; Dickey, Michael D

    2017-03-14

    Several gallium-based liquid metal alloys are liquid at room temperature. As 'liquid', such alloys have a low viscosity and a high surface tension while as 'metal', they have high thermal and electrical conductivities, similar to mercury. However, unlike mercury, these liquid metal alloys have low toxicity and a negligible vapor pressure, rendering them much safer. In comparison to mercury, the distinguishing feature of these alloys is the rapid formation of a self-limiting atomically thin layer of gallium oxide over their surface when exposed to oxygen. This oxide layer changes many physical and chemical properties of gallium alloys, including their interfacial and rheological properties, which can be employed and modulated for various applications in microfluidics. Injecting liquid metal into microfluidic structures has been extensively used to pattern and encapsulate highly deformable and reconfigurable electronic devices including electrodes, sensors, antennas, and interconnects. Likewise, the unique features of liquid metals have been employed for fabricating miniaturized microfluidic components including pumps, valves, heaters, and electrodes. In this review, we discuss liquid metal enabled microfluidic components, and highlight their desirable attributes including simple fabrication, facile integration, stretchability, reconfigurability, and low power consumption, with promising applications for highly integrated microfluidic systems.

  12. Research and development on the application of advanced control technologies to advanced nuclear reactor systems: A US national perspective

    SciTech Connect

    White, J.D.; Monson, L.R.; Carrol, D.G.; Dayal, Y.; Argonne National Lab., IL; General Electric Co., San Jose, CA )

    1989-01-01

    Control system designs for nuclear power plants are becoming more advanced through the use of digital technology and automation. This evolution is taking place because of: (1) the limitations in analog based control system performance and maintenance and availability and (2) the promise of significant improvement in plant operation and availability due to advances in digital and other control technologies. Digital retrofits of control systems in US nuclear plants are occurring now. Designs of control and protection systems for advanced LWRs are based on digital technology. The use of small inexpensive, fast, large-capacity computers in these designs is the first step of an evolutionary process described in this paper. Under the sponsorship of the US Department of Energy (DOE), Oak Ridge National Laboratory, Argonne National Laboratory, GE Nuclear Energy and several universities are performing research and development in the application of advances in control theory, software engineering, advanced computer architectures, artificial intelligence, and man-machine interface analysis to control system design. The target plant concept for the work described in this paper is the Power Reactor Inherently Safe Module reactor (PRISM), an advanced modular liquid metal reactor concept. This and other reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. 18 refs., 5 figs.

  13. Liquid metal pump

    DOEpatents

    Pennell, William E.

    1982-01-01

    The liquid metal pump comprises floating seal rings and attachment of the pump diffuser to the pump bowl for isolating structural deflections from the pump shaft bearings. The seal rings also eliminate precision machining on large assemblies by eliminating the need for a close tolerance fit between the mounting surfaces of the pump and the seals. The liquid metal pump also comprises a shaft support structure that is isolated from the pump housing for better preservation of alignment of shaft bearings. The shaft support structure also allows for complete removal of pump internals for inspection and repair.

  14. Evaluation of Enhanced Risk Monitors for Use on Advanced Reactors

    SciTech Connect

    Ramuhalli, Pradeep; Veeramany, Arun; Bonebrake, Christopher A.; Ivans, William J.; Coles, Garill A.; Hirt, Evelyn H.

    2016-09-26

    This study provides an overview of the methodology for integrating time-dependent failure probabilities into nuclear power reactor risk monitors. This prototypic enhanced risk monitor (ERM) methodology was evaluated using a hypothetical probabilistic risk assessment (PRA) model, generated using a simplified design of a liquid-metal-cooled advanced reactor (AR). Component failure data from industry compilation of failures of components similar to those in the simplified AR model were used to initialize the PRA model. Core damage frequency (CDF) over time were computed and analyzed. In addition, a study on alternative risk metrics for ARs was conducted. Risk metrics that quantify the normalized cost of repairs, replacements, or other operations and management (O&M) actions were defined and used, along with an economic model, to compute the likely economic risk of future actions such as deferred maintenance based on the anticipated change in CDF due to current component condition and future anticipated degradation. Such integration of conventional-risk metrics with alternate-risk metrics provides a convenient mechanism for assessing the impact of O&M decisions on safety and economics of the plant. It is expected that, when integrated with supervisory control algorithms, such integrated-risk monitors will provide a mechanism for real-time control decision-making that ensure safety margins are maintained while operating the plant in an economically viable manner.

  15. Advanced Reactor Innovation Evaluation Study (ARIES) Properties Archive

    DOE Data Explorer

    ARIES stands for Advanced Reactor Innovation Evaluation Study. It is a program and a team that explores the commercial potential of fusion as an energy resource. Though it is a multi-institutional program, ARIES is led by the University of California at San Diego. ARIES studies both magnetic fusion energy (MFE) and inertial fusion energy (IFE), using an approach that integrates theory, experiments, and technology. The ARIES team proposes fusion reactor designs and works to understand how technology, materials and plasma physics processes interact and influence each other. A 2005 report to the Fusion Energy Sciences Advisory Committee ("Scientific Challenges, Opportunities, and Priorities for the U.S. Fusion Energy Sciences Program") noted on page 98 an example of the importance of this materials properties aspect: "For instance, effects on plasma edge by various plasma facing materials and effects on various plasma stabilization and control techniques by highly conducting liquid metal blankets are being considered by physicists." This web page is an archive of material properties collected here for the use of the ARIES Fusion Power Plant Studies Team.

  16. Thermodynamics of liquid metal

    SciTech Connect

    Kushnirenko, A.N.

    1988-01-01

    The thermodynamics of a liquid metal based on quantum-mechanical models of the crystal, electronic, and nuclear structures of the metal are derived in this paper. The models are based on such formulations as the Bohr radius, the Boltzmann constant, the Planck Law, the Fermi surface, and the Pauli principle.

  17. Challenges in the Development of Advanced Reactors

    SciTech Connect

    P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson

    2012-08-01

    Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

  18. Liquid metal thermal electric converter

    DOEpatents

    Abbin, Joseph P.; Andraka, Charles E.; Lukens, Laurance L.; Moreno, James B.

    1989-01-01

    A liquid metal thermal electric converter which converts heat energy to electrical energy. The design of the liquid metal thermal electric converter incorporates a unique configuration which directs the metal fluid pressure to the outside of the tube which results in the structural loads in the tube to be compressive. A liquid metal thermal electric converter refluxing boiler with series connection of tubes and a multiple cell liquid metal thermal electric converter are also provided.

  19. Irradiation Facilities at the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-12-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC – formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world’s data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities1. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens.

  20. Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

    SciTech Connect

    Loika, E.F.

    1994-09-22

    This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate.

  1. Advanced Reactors Around the World

    SciTech Connect

    Majumdar, Debu

    2003-09-01

    At the end of 2002, 441 nuclear power plants were operating around the globe and providing 17% of the world's electricity. Although the rate of population growth has slowed, recent United Nations data suggest that two billion more people will be added to the world by 2050. A special report commissioned by the Intergovernmental Panel on Climate Change estimated that electricity demand would grow almost eight-fold from 2000 to 2050 in a high economic grown scenario and more than double in a low-growth scenario. There is also a global aspiration to keep the environment pristine. Because of these reasons, it is expected that a large number of new nuclear reactors may be operating by 2050. Realization of this has created an impetus for the development of a new generation of reactors in several countries. The goal is to make nuclear power cost-competitive with other resources and to enhance safety to a level that no evacuation outside a plant site would be necessary. It should also generate less waste, prevent materials diversion for weapons production, and be sustainable. This article discusses the status of next-generation reactors under development around the world. Specifically highlighted are efforts related to the Generation IV International Forum (GIF) and its six reactor concepts for research and development: Very High Temperature Reactor (VHTR); Gas-Cooled Fast Reactor (GFR); Supercritical Water-Cooled Reactor (SCWR); Sodium-Cooled Fast Reactor (SFR); Lead-Cooled Fast Reactor (LFR); and Molten Salt Reactor (MSR). Also highlighted are nuclear activities specific to Russia and India.

  2. Advances in Tandem Mirror fusion power reactors

    SciTech Connect

    Perkins, L.J.; Logan, B.G.

    1986-05-20

    The Tandem Mirror exhibits several distinctive features which make the reactor embodiment of the principle very attractive: Simple low-technology linear central cell; steady-state operation; high-..beta.. operation; no driven current or disruptions; divertorless operation; direction conversion of end-loss power; low-surface heat loads; and advanced fusion fuel capability. In this paper, we examine these features in connection with two tandem mirror reactor designs, MARS and MINIMARS, and several advanced reactor concepts including the wall-stabilized reactor and the field-reversed mirror. With a novel compact end plug scheme employing octopole stabilization, MINIMARS is expressly designed for short construction times, factory-built modules, and a small (600 MWe) but economic reactor size. We have also configured the design for low radioactive afterheat and inherent/passive safety under LOCA/LOFA conditions, thereby obviating the need for expensive engineered safety systems. In contrast to the complex and expensive double-quadrupole end-cell of the MARS reactor, the compact octopole end-cell of MINIMARS enables ignition to be achieved with much shorter central cell lengths and considerably improves the economy of scale for small (approx.250 to 600 MWe) tandem mirror reactors. Finally, we examine the prospects for realizing the ultimate potential of the tandem mirror with regard to both innovative configurations and novel neutron energy conversion schemes, and stress that advanced fuel applications could exploit its unique reactor features.

  3. Advanced PPA Reactor and Process Development

    NASA Technical Reports Server (NTRS)

    Wheeler, Raymond; Aske, James; Abney, Morgan B.; Miller, Lee A.; Greenwood, Zachary

    2012-01-01

    Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA s Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development work.

  4. Advanced Catalytic Hydrogenation Retrofit Reactor

    SciTech Connect

    Reinaldo M. Machado

    2002-08-15

    Industrial hydrogenation is often performed using a slurry catalyst in large stirred-tank reactors. These systems are inherently problematic in a number of areas, including industrial hygiene, process safety, environmental contamination, waste production, process operability and productivity. This program proposed the development of a practical replacement for the slurry catalysts using a novel fixed-bed monolith catalyst reactor, which could be retrofitted onto an existing stirred-tank reactor and would mitigate many of the minitations and problems associated with slurry catalysts. The full retrofit monolith system, consisting of a recirculation pump, gas/liquid ejector and monolith catalyst, is described as a monolith loop reactor or MLR. The MLR technology can reduce waste and increase raw material efficiency, which reduces the overall energy required to produce specialty and fine chemicals.

  5. 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.

  6. Advances in reactor physics education: Visualization of reactor parameters

    SciTech Connect

    Snoj, L.; Kromar, M.; Zerovnik, G.

    2012-07-01

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

  7. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    SciTech Connect

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  8. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    SciTech Connect

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  9. Advanced High Temperature Reactor Systems and Economic Analysis

    SciTech Connect

    Holcomb, David Eugene; Peretz, Fred J; Qualls, A L

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a large-output [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR's large thermal output enables direct comparison of its performance and requirements with other high output reactor concepts. As high-temperature plants, FHRs can support either high-efficiency electricity generation or industrial process heat production. The AHTR analysis presented in this report is limited to the electricity generation mission. FHRs, in principle, have the potential to be low-cost electricity producers while maintaining full passive safety. However, no FHR has been built, and no FHR design has reached the stage of maturity where realistic economic analysis can be performed. The system design effort described in this report represents early steps along the design path toward being able to predict the cost and performance characteristics of the AHTR as well as toward being able to identify the technology developments necessary to build an FHR power plant. While FHRs represent a distinct reactor class, they inherit desirable attributes from other thermal power plants whose characteristics can be studied to provide general guidance on plant configuration, anticipated performance, and costs. Molten salt reactors provide experience on the materials, procedures, and components necessary to use liquid fluoride salts. Liquid metal reactors provide design experience on using low-pressure liquid coolants, passive decay heat removal, and hot refueling. High temperature gas-cooled reactors provide experience with coated particle fuel and graphite components. Light water reactors (LWRs) show the potentials of transparent, high-heat capacity coolants with low chemical reactivity. Modern coal-fired power plants provide design experience with

  10. Description and identification of difficulties arising from the application of a cleaning process in operating conditions for the treatment of components used on liquid metal fast reactors (LMFR). A technical designed approach to avoid these situations.

    PubMed

    Rodriguez, G; Karpov, A V; Nalimov, Y P

    2001-01-01

    The cleaning process is one of the major maintenance operation for liquid metal fast reactors (LMFRs), both in operation and in their decommissioning stage. Russian and French cleaning processes are briefly described, including problems which have arisen during the processes. It appears that the cause of these problems is always connected to bad draining of the component, resulting in a vigorous reaction between vapour or liquid water and the bulk of sodium. From this discussion, the paper makes major recommendations for the efficient and safe cleaning of sodium wetted components, and proposes several processes which should be developed in order to deal with difficult situations, for example the removal of large amounts of undrainable sodium.

  11. Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect

    BOWEN, W.W.

    1999-11-08

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FFTF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This document reflects the 1 Oct 1999 baseline.

  12. Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect

    GANTT, D.A.

    2000-01-12

    The Advanced Reactors Transition (ART) Resource Loaded Schedule (RLS) provides a cost and schedule baseline for managing the project elements within the ART Program. The Fast Flux Test Facility (FETF) activities are delineated through the end of FY 2000, assuming continued standby. The Nuclear Energy (NE) Legacies and Plutonium Recycle Test Reactor (PRTR) activities are delineated through the end of the deactivation process. This revision reflects the 19 Oct 1999 baseline.

  13. The IAEA CRP on Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste

    SciTech Connect

    Maschek, W.; Chen, X.; Rineiski, A.; Schikorr, M.; Stanculescu, A.; Arien, B.; Malambu, E.; Bai, Y.; Li, J.; Wu, Y.; Zheng, S.; Chabert, C.; Peneliau, Y.; Chebeskov, A.; Dekoussar, V.; Vorotyntsev, M.; da Cruz, D.F.; Devan, K.; Gopalakrishnan, V.; Harish, R.; Mohanakrishnan, P.; Pandikumar, G.; Dulla, S.; Ravetto, P.; Feynberg, O.; Ignatiev, V.; Subbotin, V.; Surenkov, A.; Zakirov, R.; Kophazi, J.; Szieberth, M.; Morita, K.; Srivenkatesan, R.; Taczanowski, S.; Tucek, K.; Wider, H.; Vertes, P.; Uhlir, J.

    2007-07-01

    In 2003, the IAEA has initiated the Coordinated Research Project (CRP) on 'Studies of Advanced Reactor Technology Options for Effective Incineration of Radioactive Waste'. The overall objective of the CRP, performed within the framework of IAEA's Nuclear Energy Department's Technical Working Group on Fast Reactors, is to increase the capability of Member States in developing and applying advanced technologies in the area of long-lived radioactive waste utilization and transmutation. Twenty institutions from 15 Member States and one international organization participated in this CRP. The CRP concentrated on the assessment of the dynamic behavior of various transmutation systems. The reactor systems investigated comprise critical reactors, sub-critical accelerator driven systems with heavy liquid metal and gas cooling, critical molten salt systems, and hybrid fusion/fission systems. Both fertile and fertile-free fuel options have been investigated. Apart from the benchmarking of steady state core configurations (including the investigation of transmutation potential, burn-up behavior and decay heat of minor actinide (MA) bearing fuels), the CRP participants determined the safety coefficients for the individual systems and, in a second stage, performed transient analyses which reflected the generic safety related behavior of the various reactors types. (authors)

  14. Ionic imbalance induced self-propulsion of liquid metals

    PubMed Central

    Zavabeti, Ali; Daeneke, Torben; Chrimes, Adam F.; O'Mullane, Anthony P.; Zhen Ou, Jian; Mitchell, Arnan; Khoshmanesh, Khashayar; Kalantar-zadeh, Kourosh

    2016-01-01

    Components with self-propelling abilities are important building blocks of small autonomous systems and the characteristics of liquid metals are capable of fulfilling self-propulsion criteria. To date, there has been no exploration regarding the effect of electrolyte ionic content surrounding a liquid metal for symmetry breaking that generates motion. Here we show the controlled actuation of liquid metal droplets using only the ionic properties of the aqueous electrolyte. We demonstrate that pH or ionic concentration gradients across a liquid metal droplet induce both deformation and surface Marangoni flow. We show that the Lippmann dominated deformation results in maximum velocity for the self-propulsion of liquid metal droplets and illustrate several key applications, which take advantage of such electrolyte-induced motion. With this finding, it is possible to conceive the propulsion of small entities that are constructed and controlled entirely with fluids, progressing towards more advanced soft systems. PMID:27488954

  15. Ionic imbalance induced self-propulsion of liquid metals.

    PubMed

    Zavabeti, Ali; Daeneke, Torben; Chrimes, Adam F; O'Mullane, Anthony P; Zhen Ou, Jian; Mitchell, Arnan; Khoshmanesh, Khashayar; Kalantar-Zadeh, Kourosh

    2016-08-04

    Components with self-propelling abilities are important building blocks of small autonomous systems and the characteristics of liquid metals are capable of fulfilling self-propulsion criteria. To date, there has been no exploration regarding the effect of electrolyte ionic content surrounding a liquid metal for symmetry breaking that generates motion. Here we show the controlled actuation of liquid metal droplets using only the ionic properties of the aqueous electrolyte. We demonstrate that pH or ionic concentration gradients across a liquid metal droplet induce both deformation and surface Marangoni flow. We show that the Lippmann dominated deformation results in maximum velocity for the self-propulsion of liquid metal droplets and illustrate several key applications, which take advantage of such electrolyte-induced motion. With this finding, it is possible to conceive the propulsion of small entities that are constructed and controlled entirely with fluids, progressing towards more advanced soft systems.

  16. Ionic imbalance induced self-propulsion of liquid metals

    NASA Astrophysics Data System (ADS)

    Zavabeti, Ali; Daeneke, Torben; Chrimes, Adam F.; O'Mullane, Anthony P.; Zhen Ou, Jian; Mitchell, Arnan; Khoshmanesh, Khashayar; Kalantar-Zadeh, Kourosh

    2016-08-01

    Components with self-propelling abilities are important building blocks of small autonomous systems and the characteristics of liquid metals are capable of fulfilling self-propulsion criteria. To date, there has been no exploration regarding the effect of electrolyte ionic content surrounding a liquid metal for symmetry breaking that generates motion. Here we show the controlled actuation of liquid metal droplets using only the ionic properties of the aqueous electrolyte. We demonstrate that pH or ionic concentration gradients across a liquid metal droplet induce both deformation and surface Marangoni flow. We show that the Lippmann dominated deformation results in maximum velocity for the self-propulsion of liquid metal droplets and illustrate several key applications, which take advantage of such electrolyte-induced motion. With this finding, it is possible to conceive the propulsion of small entities that are constructed and controlled entirely with fluids, progressing towards more advanced soft systems.

  17. Method of foaming a liquid metal

    DOEpatents

    Fischer, Albert K.; Johnson, Carl E.

    1980-01-01

    The addition of a small quantity of barium to liquid metal NaK or sodium has been found to promote foam formation and improve bubble retention in the liquid metal. A stable liquid metal foam will provide a more homogeneous liquid metal flow through the channel of a two-phase liquid metal MHD power generator to improve operating efficiency.

  18. Liquid Metal Dynamo Measurements

    NASA Astrophysics Data System (ADS)

    Luh, W. J.; Choi, Y. H.; Hardy, B. S.; Brown, M. R.

    1997-11-01

    Detection of convected magnetic fields in a small-scale liquid metal dynamo is attempted. Initial experiments will focus on the conversion of toroidal to poloidal flux (a version of the ω effect). A precision vector magnetometer will be used to measure the effect of a rotating magnetofluid on a static magnetic field. Water will be used as a control medium and effects will be compared with a conducting medium (liquid sodium or NaK). A small spherical flask (0.16 m diameter) houses 2 liters of fluid, a teflon stirrer creates an asymmetrical flow pattern, and Helmholtz coils generate a constant magnetic field on the order of 10 gauss. The Reynold's number will be of order unity.

  19. Liquid metal thermoacoustic engine

    SciTech Connect

    Swift, G.W.; Migliori, A.; Wheatley, J.C.

    1986-01-01

    We are studying a liquid metal thermoacoustic engine both theoretically and experimentally. This type of engine promises to produce large quantities of electrical energy from heat at modest efficiency with no moving parts. A sound wave is usually thought of as consisting of pressure oscillations, but always attendant to the pressure oscillation are temperature oscillations. The combination produces a rich variety of ''thermoacoustic'' effects. These effects are usually so small that they are never noticed in everyday life; nevertheless under the right circumstances they can be harnessed to produce powerful heat engines, heat pumps, and refrigerators. In our liquid metal thermoacoustic engine, heat flow from a high temperature source to a low temperature sink generates a high-amplitude standing acoustic wave in liquid sodium. This acoustic power is converted to electric power by a simple magnetohydrodynamic effect at the acoustic oscillation frequency. We have developed a detailed thermoacoustic theory applicable to this engine, and find that a reasonably designed liquid sodium engine operating between 700/sup 0/C and 100/sup 0/C should generate about 60 W/cm/sup 2/ of acoustic power at about 1/3 of Carnot's efficiency. Construction of a 3000 W-thermal laboratory model engine has just been completed, and we have exciting preliminary experimental results as of the time of preparation of this manuscript showing, basically, that the engine works. We have also designed and built a 1 kHz liquid sodium magnetohydrodynamic generator and have extensive measurements on it. It is now very well characterized both experimentally and theoretically. The first generator of its kind, it already converts acoustic power to electric power with 40% efficiency. 16 refs., 5 figs.

  20. Advanced Test Reactor National Scientific User Facility

    SciTech Connect

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  1. Reliability assurance for regulation of advanced reactors

    SciTech Connect

    Fullwood, R.; Lofaro, R.; Samanta, P.

    1991-12-31

    The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. This paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics.

  2. Integral fast reactor safety features

    SciTech Connect

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents.

  3. Development of oxygen sensors for use in liquid metal

    SciTech Connect

    Van Nieuwenhove, Rudi; Ejenstam, Jesper; Szakalos, Peter

    2015-07-01

    For generation IV reactor concepts, based on liquid metal cooling, there is a need for robust oxygen sensors which can be used in the core of the reactor since corrosion can only be kept sufficiently low by controlling the dissolved oxygen content in the liquid metal. A robust, ceramic membrane type sensor has been developed at IFE/Halden (Norway) and tested in an autoclave system at KTH (Sweden). The sensor has been tested in lead-bismuth at 550 deg. C and performed well. (authors)

  4. Advanced reactor physics methods for heterogeneous reactor cores

    NASA Astrophysics Data System (ADS)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  5. 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.

  6. POWER GENERATION FROM LIQUID METAL NUCLEAR FUEL

    DOEpatents

    Dwyer, O.E.

    1958-12-23

    A nuclear reactor system is described wherein the reactor is the type using a liquid metal fuel, such as a dispersion of fissile material in bismuth. The reactor is designed ln the form of a closed loop having a core sectlon and heat exchanger sections. The liquid fuel is clrculated through the loop undergoing flssion in the core section to produce heat energy and transferrlng this heat energy to secondary fluids in the heat exchanger sections. The fission in the core may be produced by a separate neutron source or by a selfsustained chain reaction of the liquid fuel present in the core section. Additional auxiliary heat exchangers are used in the system to convert water into steam which drives a turbine.

  7. Mirror Advanced Reactor Study interim design report

    SciTech Connect

    Not Available

    1983-04-01

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

  8. 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

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an

  9. Solute diffusion in liquid metals

    NASA Technical Reports Server (NTRS)

    Bhat, B. N.

    1973-01-01

    A gas model of diffusion in liquid metals is presented. In this model, ions of liquid metals are assumed to behave like the molecules in a dense gas. Diffusion coefficient of solute is discussed with reference to its mass, ionic size, and pair potential. The model is applied to the case of solute diffusion in liquid silver. An attempt was made to predict diffusion coefficients of solutes with reasonable accuracy.

  10. Assessment of Sensor Technologies for Advanced Reactors

    SciTech Connect

    Korsah, Kofi; Ramuhalli, Pradeep; Vlim, R.; Kisner, Roger A.; Britton, Jr, Charles L.; Wootan, D. W.; Anheier, Jr, N. C.; Diaz, A. A.; Hirt, E. H.; Chien, H. T.; Sheen, S.; Bakhtiari, Sasan; Gopalsami, S.; Heifetz, A.; Tam, S. W.; Park, Y.; Upadhyaya, B. R.; Stanford, A.

    2016-10-01

    Sensors and measurement technologies provide information on processes, support operations and provide indications of component health. They are therefore crucial to plant operations and to commercialization of advanced reactors (AdvRx). This report, developed by a three-laboratory team consisting of Argonne National Laboratory (ANL), Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL), provides an assessment of sensor technologies and a determination of measurement needs for AdvRx. It provides the technical basis for identifying and prioritizing research targets within the instrumentation and control (I&C) Technology Area under the Department of Energy’s (DOE’s) Advanced Reactor Technology (ART) program and contributes to the design and implementation of AdvRx concepts.

  11. Instrumentation to Enhance Advanced Test Reactor Irradiations

    SciTech Connect

    J. L. Rempe; D. L. Knudson; K. G. Condie; J. E. Daw; S. C. Taylor

    2009-09-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR will support basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors has been completed. Based on this review, recommendations are made with respect to what instrumentation is needed at the ATR and a strategy has been developed for obtaining these sensors. Progress toward implementing this strategy is reported in this document. It is anticipated that this report will be updated on an annual basis.

  12. Integral fast reactor concept inherent safety features

    SciTech Connect

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1987-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFT development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: 1) a liquid metal (sodium) coolant, 2) a pool-type reactor primary system configuration, 3) an advanced ternary alloy metallic fuel, and 4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS).

  13. Integral Fast Reactor concept inherent safety features

    SciTech Connect

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS).

  14. A simple calculation of control assembly effectiveness in a liquid-metal fast breeder reactor by a transport-diffusion equivalence method

    SciTech Connect

    Benoist, P. ); Carta, M. ); Palmiotti, G. ); Salvatores, M. )

    1989-11-01

    A method to calculate the effectiveness of the control assembly in a fast neutron reactor is proposed. For each type of heterogeneous assembly (control or follower), a polar parameter, taking into account the assembly absorption and the axial leakage of neutrons inside the assembly, is defined. In a similar way, a bipolar parameter, taking into account the reaction of the assembly to a transverse flux gradient, is also defined. These two parameters, deduced from transport theory, are used to determine the absorption cross section and the diffusion coefficient of an equivalent homogeneous control or follower assembly. These new parameters are introduced in a one-group diffusion code, calculating the reactor as a whole with any number of control and follower assemblies. An approximate generalization to multigroup theory is proposed. Numerical comparisons show that this equivalent diffusion method gives results that are much closer to transport results than those obtained by the classical diffusion theory.

  15. Performance and reliability monitoring of advanced reactors

    SciTech Connect

    Robinson, D. G.

    2006-07-01

    Advanced nuclear reactor designs must be designed for long core life and high fuel burn-up. This will pose new challenges for monitoring the state of the highly integrated and largely inaccessible components within the primary reactor vessel. As reactors age, characterizing and predicting the internal operational performance (e.g. vibration/flow characteristics) will be critical for determining maintenance strategies and early identification of safety and reliability issues. This paper discusses a current research program into a new approach for monitoring that provides the capability to combine information from in-situ monitoring sensors within a reactor with maintenance and repair histories to characterize the current probability of a failure event. The proposed approach suggests the application of Bayesian concepts similar to that used in medical monitoring. System-based medical monitoring utilizes hierarchical Bayesian methods such as Markov Chain Monte Carlo (MCMC) and particle filtering methods to deliver estimates of the state of health of a process based on data taken from the process in real-time. Preliminary results from the application of the methodology to the monitoring of corrosion in electronics is presented. (authors)

  16. Beryllium Use in the Advanced Test Reactor

    SciTech Connect

    Glen R. Longhurst

    2007-12-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) began operation in 1967. It makes use of a unique serpentine fuel core design and a beryllium reflector. Reactor control is achieved with rotating beryllium cylinders to which have been fastened plates of hafnium. Over time, the beryllium develops rather high helium content because of nuclear transmutations and begins to swell. The beryllium must be replaced at nominally 10-year intervals. Determination of when the replacement is made is by visual observation using a periscope to examine the beryllium surface for cracking and swelling. Disposition of the irradiated beryllium was once accomplished in the INL’s Radioactive Waste Management Complex, but that is no longer possible. Among contributing reasons are high levels of specific radioactive contaminants including transuranics. The INL is presently considering disposition pathways for this irradiated beryllium, but presently is storing it in the canal adjacent to the reactor. Numerous issues are associated with this situation including (1) Is there a need for ultra-low uranium material? (2) Is there a need to recover tritium from irradiated beryllium either because this is a strategic material resource or in preparation for disposal? (3) Is there a need to remove activation and fission products from irradiated beryllium? (4) Will there be enough material available to meet requirements for research reactors (fission and fusion)? In this paper will be discussed the present status of considerations on these issues.

  17. Advanced Small Modular Reactor Economics Status Report

    SciTech Connect

    Harrison, Thomas J.

    2014-10-01

    This report describes the data collection work performed for an advanced small modular reactor (AdvSMR) economics analysis activity at the Oak Ridge National Laboratory. The methodology development and analytical results are described in separate, stand-alone documents as listed in the references. The economics analysis effort for the AdvSMR program combines the technical and fuel cycle aspects of advanced (non-light water reactor [LWR]) reactors with the market and production aspects of SMRs. This requires the collection, analysis, and synthesis of multiple unrelated and potentially high-uncertainty data sets from a wide range of data sources. Further, the nature of both economic and nuclear technology analysis requires at least a minor attempt at prediction and prognostication, and the far-term horizon for deployment of advanced nuclear systems introduces more uncertainty. Energy market uncertainty, especially the electricity market, is the result of the integration of commodity prices, demand fluctuation, and generation competition, as easily seen in deregulated markets. Depending on current or projected values for any of these factors, the economic attractiveness of any power plant construction project can change yearly or quarterly. For long-lead construction projects such as nuclear power plants, this uncertainty generates an implied and inherent risk for potential nuclear power plant owners and operators. The uncertainty in nuclear reactor and fuel cycle costs is in some respects better understood and quantified than the energy market uncertainty. The LWR-based fuel cycle has a long commercial history to use as its basis for cost estimation, and the current activities in LWR construction provide a reliable baseline for estimates for similar efforts. However, for advanced systems, the estimates and their associated uncertainties are based on forward-looking assumptions for performance after the system has been built and has achieved commercial operation

  18. Plant maintenance and advanced reactors, 2007

    SciTech Connect

    Agnihotri, Newal

    2007-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: A new day for energy in America; Committed to success more than ever, by Andy White, GE--Hitachi Nuclear Energy; Competitive technology for decades, by Steve Tritch, Westinghouse Electric Company; Pioneers of positive community relationship, by Exelon Nuclear; A robust design for 60-years, by Ray Ganthner, Areva; Aiming at no evacuation plants, by Kumiaki Moriya, Hitachi-GE Nuclear Energy, Ltd.; and, Desalination and hydrogen economy, by Dr. I. Khamis, International Atomic Energy Agency. Industry innovation articles in this issue are: Reactor vessel closure head project, by Jeff LeClair, Prairie Island Nuclear Generating Plant; and Submersible remote-operated vehicle, by Michael S. Rose, Entergy's Fitzpatrick Nuclear Station.

  19. Status of the advanced neutron source. [Advanced Neutron Source Reactor

    SciTech Connect

    Hayter, J.B.

    1990-01-01

    Research reactors in the United States are becoming more and more outdated, at a time when neutron scattering is being recognized as an increasingly important technique in areas vital to the US scientific and technological future. The last US research reactor was constructed over 25 years ago, whereas new facilities have been built or are under construction in Japan, Russia and, especially, Western Europe, which now has a commanding lead in this important field. Concern over this situation in the early 1980's by a number of organizations, including the National Academy of Sciences, led to a recommendation that design work start urgently on an advanced US neutron research facility. This recommendation is realized in the Advanced Neutron Source Project. The centerpiece of the Advanced Neutron Source will be a new research reactor of unprecedented flux (>7.5 {times} 10{sup 19} m{sup {minus}2}{center dot}s{sup {minus}1}), equipped with a wide variety of state-of-the-art spectrometers and diffractometers on hot, thermal, and cold neutron beams. Very cold and ultracold neutron beams will also be provided for specialized experiments. This paper will discuss the current status of the design and the plans for scattering instrumentation. 5 refs.

  20. Liquid Metal Pump Technologies for Nuclear Surface Power

    NASA Technical Reports Server (NTRS)

    Polzin, Kurt A.

    2007-01-01

    Multiple liquid metal pump options are reviewed for the purpose of determining the technologies that are best suited for inclusion in a nuclear reactor thermal simulator intended to test prototypical space nuclear system components. Conduction, induction, and thermoelectric electromagnetic pumps are evaluated based on their performance characteristics and the technical issues associated incorporation into a reactor system. The thermoelectric electromagnetic pump is recommended for inclusion in the present system based on favorable quantitative and qualitative measures relative to the other options under consideration.

  1. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    SciTech Connect

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost.

  2. Advanced Burner Reactor Preliminary NEPA Data Study.

    SciTech Connect

    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

    2007-10-15

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  3. Structural materials challenges for advanced reactor systems

    NASA Astrophysics Data System (ADS)

    Yvon, P.; Carré, F.

    2009-03-01

    Key technologies for advanced nuclear systems encompass high temperature structural materials, fast neutron resistant core materials, and specific reactor and power conversion technologies (intermediate heat exchanger, turbo-machinery, high temperature electrolytic or thermo-chemical water splitting processes, etc.). The main requirements for the materials to be used in these reactor systems are dimensional stability under irradiation, whether under stress (irradiation creep or relaxation) or without stress (swelling, growth), an acceptable evolution under ageing of the mechanical properties (tensile strength, ductility, creep resistance, fracture toughness, resilience) and a good behavior in corrosive environments (reactor coolant or process fluid). Other criteria for the materials are their cost to fabricate and to assemble, and their composition could be optimized in order for instance to present low-activation (or rapid desactivation) features which facilitate maintenance and disposal. These requirements have to be met under normal operating conditions, as well as in incidental and accidental conditions. These challenging requirements imply that in most cases, the use of conventional nuclear materials is excluded, even after optimization and a new range of materials has to be developed and qualified for nuclear use. This paper gives a brief overview of various materials that are essential to establish advanced systems feasibility and performance for in pile and out of pile applications, such as ferritic/martensitic steels (9-12% Cr), nickel based alloys (Haynes 230, Inconel 617, etc.), oxide dispersion strengthened ferritic/martensitic steels, and ceramics (SiC, TiC, etc.). This article gives also an insight into the various natures of R&D needed on advanced materials, including fundamental research to investigate basic physical and chemical phenomena occurring in normal and accidental operating conditions, lab-scale tests to characterize candidate materials

  4. ASME Material Challenges for Advanced Reactor Concepts

    SciTech Connect

    Piyush Sabharwall; Ali Siahpush

    2013-07-01

    This study presents the material Challenges associated with Advanced Reactor Concept (ARC) such as the Advanced High Temperature Reactor (AHTR). ACR are the next generation concepts focusing on power production and providing thermal energy for industrial applications. The efficient transfer of energy for industrial applications depends on the ability to incorporate cost-effective heat exchangers between the nuclear heat transport system and industrial process heat transport system. The heat exchanger required for AHTR is subjected to a unique set of conditions that bring with them several design challenges not encountered in standard heat exchangers. The corrosive molten salts, especially at higher temperatures, require materials throughout the system to avoid corrosion, and adverse high-temperature effects such as creep. Given the very high steam generator pressure of the supercritical steam cycle, it is anticipated that water tube and molten salt shell steam generators heat exchanger will be used. In this paper, the ASME Section III and the American Society of Mechanical Engineers (ASME) Section VIII requirements (acceptance criteria) are discussed. Also, the ASME material acceptance criteria (ASME Section II, Part D) for high temperature environment are presented. Finally, lack of ASME acceptance criteria for thermal design and analysis are discussed.

  5. Liquid-metal dip seal with pneumatic spring

    DOEpatents

    Poindexter, Allan M.

    1977-01-01

    An improved liquid-metal dip seal for sealing the annulus between rotating plugs in the reactor vessel head of a liquid-metal fast-breeder nuclear reactor has two legs of differing widths communicating under a seal blade; the wide leg is also in communication with cover gas of the reactor and the narrow leg is also in communication with an isolated plug annulus above the seal. The annulus contains inert gas which acts as a pneumatic spring. Upon increasing cover gas pressure which depresses the level in the wide leg and greatly increases the level in the narrow leg, the pneumatic spring is compressed, and resists further level changes, thus preventing radioactive cover gas from bubbling through the seal.

  6. Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

    SciTech Connect

    Loflin, Leonard; McRimmon, Beth

    2014-12-18

    This report summarizes a project by EPRI to include requirements for small modular light water reactors (smLWR) into the EPRI Utility Requirements Document (URD) for Advanced Light Water Reactors. The project was jointly funded by EPRI and the U.S. Department of Energy (DOE). The report covers the scope and content of the URD, the process used to revise the URD to include smLWR requirements, a summary of the major changes to the URD to include smLWR, and how to use the URD as revised to achieve value on new plant projects.

  7. Advanced reactor vessel steels for reactors with supercritical coolant parameters

    NASA Astrophysics Data System (ADS)

    Markov, S. I.; Dub, V. S.; Lebedev, A. G.; Kuleshova, E. A.; Balikoev, A. G.; Makarycheva, E. V.; Tolstykh, D. S.; Frolov, A. S.; Krikun, E. V.

    2016-09-01

    A set of studies, tests, and technological works is performed to design promising high-strength vessel steels for reactors with supercritical coolant parameters. Compositions and technological parameters are proposed for the production of reference steel (within the limits of the grade composition of 15Kh2NMFA-A steel) and high-nickel steel. These steels are characterized by high properties, including metallurgical quality and service and technological parameters. Steel of the reference composition has high (higher by 15%) strength properties, improved viscoplastic properties, and ductile-brittle transition temperature t c of at most-125°C. The strength properties of the high-nickel steel are higher than those of the existing steels by 40-50% and higher than those of advanced foreign steels by 15-20% at ductile-brittle transition temperature t c of at most-165°C. Moreover, the designed steels are characterized by a low content of harmful impurity elements and nonmetallic inclusions, a fine-grained structure, and a low susceptibility to thermal embrittlement.

  8. LBB application in the US operating and advanced reactors

    SciTech Connect

    Wichman, K.; Tsao, J.; Mayfield, M.

    1997-04-01

    The regulatory application of leak before break (LBB) for operating and advanced reactors in the U.S. is described. The U.S. Nuclear Regulatory Commission (NRC) has approved the application of LBB for six piping systems in operating reactors: reactor coolant system primary loop piping, pressurizer surge, safety injection accumulator, residual heat removal, safety injection, and reactor coolant loop bypass. The LBB concept has also been applied in the design of advanced light water reactors. LBB applications, and regulatory considerations, for pressurized water reactors and advanced light water reactors are summarized in this paper. Technology development for LBB performed by the NRC and the International Piping Integrity Research Group is also briefly summarized.

  9. Advanced ceramic cladding for water reactor fuel

    SciTech Connect

    Feinroth, H.

    2000-07-01

    Under the US Department of Energy's Nuclear Energy Research Initiatives (NERI) program, continuous fiber ceramic composites (CFCCs) are being developed as cladding for water reactor fuel elements. The purpose is to substantially increase the passive safety of water reactors. A development effort was initiated in 1991 to fabricate CFCC-clad tubes using commercially available fibers and a sol-gel process developed by McDermott Technologies. Two small-diameter CFCC tubes were fabricated using pure alumina and alumina-zirconia fibers in an alumina matrix. Densities of {approximately}60% of theoretical were achieved. Higher densities are required to guarantee fission gas containment. This NERI work has just begun, and only preliminary results are presented herein. Should the work prove successful, further development is required to evaluate CFCC cladding and performance, including in-pile tests containing fuel and exploring a marriage of CFCC cladding materials with suitable advanced fuel and core designs. The possibility of much higher temperature core designs, possibly cooled with supercritical water, and achievement of plant efficiencies {ge}50% would be examined.

  10. Corrosion of spent Advanced Test Reactor fuel

    SciTech Connect

    Lundberg, L.B.; Croson, M.L.

    1994-11-01

    The results of a study of the condition of spent nuclear fuel elements from the Advanced Test Reactor (ATR) currently being stored underwater at the Idaho National Engineering Laboratory (INEL) are presented. This study was motivated by a need to estimate the corrosion behavior of dried, spent ATR fuel elements during dry storage for periods up to 50 years. The study indicated that the condition of spent ATR fuel elements currently stored underwater at the INEL is not very well known. Based on the limited data and observed corrosion behavior in the reactor and in underwater storage, it was concluded that many of the fuel elements currently stored under water in the facility called ICPP-603 FSF are in a degraded condition, and it is probable that many have breached cladding. The anticipated dehydration behavior of corroded spent ATR fuel elements was also studied, and a list of issues to be addressed by fuel element characterization before and after forced drying of the fuel elements and during dry storage is presented.

  11. Compact, Lightweight Electromagnetic Pump for Liquid Metal

    NASA Technical Reports Server (NTRS)

    Godfroy, Thomas; Palzin, Kurt

    2010-01-01

    A proposed direct-current electromagnetic pump for circulating a molten alkali metal alloy would be smaller and lighter and would demand less input power, relative to currently available pumps of this type. (Molten alkali metals are used as heat-transfer fluids in high-temperature stages of some nuclear reactors.) The principle of operation of this or any such pump involves exploitation of the electrical conductivity of the molten metal: An electric current is made to pass through the liquid metal along an axis perpendicular to the longitudinal axis of the flow channel, and a magnetic field perpendicular to both the longitudinal axis and the electric current is superimposed on the flowchannel region containing the electric current. The interaction between the electric current and the magnetic field produces the pumping force along the longitudinal axis. The advantages of the proposed pump over other such pumps would accrue from design features that address overlapping thermal and magnetic issues.

  12. 77 FR 76089 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-26

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... Committee on Reactor Safeguards. BILLING CODE 7590-01-P...

  13. Advanced nuclear reactor public opinion project

    SciTech Connect

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  14. Advanced Small Modular Reactor Economics Model Development

    SciTech Connect

    Harrison, Thomas J.

    2014-10-01

    The US Department of Energy Office of Nuclear Energy’s Advanced Small Modular Reactor (SMR) research and development activities focus on four key areas: Developing assessment methods for evaluating advanced SMR technologies and characteristics; and Developing and testing of materials, fuels and fabrication techniques; and Resolving key regulatory issues identified by US Nuclear Regulatory Commission and industry; and Developing advanced instrumentation and controls and human-machine interfaces. This report focuses on development of assessment methods to evaluate advanced SMR technologies and characteristics. Specifically, this report describes the expansion and application of the economic modeling effort at Oak Ridge National Laboratory. Analysis of the current modeling methods shows that one of the primary concerns for the modeling effort is the handling of uncertainty in cost estimates. Monte Carlo–based methods are commonly used to handle uncertainty, especially when implemented by a stand-alone script within a program such as Python or MATLAB. However, a script-based model requires each potential user to have access to a compiler and an executable capable of handling the script. Making the model accessible to multiple independent analysts is best accomplished by implementing the model in a common computing tool such as Microsoft Excel. Excel is readily available and accessible to most system analysts, but it is not designed for straightforward implementation of a Monte Carlo–based method. Using a Monte Carlo algorithm requires in-spreadsheet scripting and statistical analyses or the use of add-ons such as Crystal Ball. An alternative method uses propagation of error calculations in the existing Excel-based system to estimate system cost uncertainty. This method has the advantage of using Microsoft Excel as is, but it requires the use of simplifying assumptions. These assumptions do not necessarily bring into question the analytical results. In fact, the

  15. 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.

  16. 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.

  17. Liquid metal reactor air cooling baffle

    DOEpatents

    Hunsbedt, Anstein

    1994-01-01

    A baffle is provided between a relatively hot containment vessel and a relatively cold silo for enhancing air cooling performance. The baffle includes a perforate inner wall positionable outside the containment vessel to define an inner flow riser therebetween, and an imperforate outer wall positionable outside the inner wall to define an outer flow riser therebetween. Apertures in the inner wall allow thermal radiation to pass laterally therethrough to the outer wall, with cooling air flowing upwardly through the inner and outer risers for removing heat.

  18. Liquid metal reactor air cooling baffle

    DOEpatents

    Hunsbedt, A.

    1994-08-16

    A baffle is provided between a relatively hot containment vessel and a relatively cold silo for enhancing air cooling performance. The baffle includes a perforate inner wall positionable outside the containment vessel to define an inner flow riser therebetween, and an imperforate outer wall positionable outside the inner wall to define an outer flow riser therebetween. Apertures in the inner wall allow thermal radiation to pass laterally therethrough to the outer wall, with cooling air flowing upwardly through the inner and outer risers for removing heat. 3 figs.

  19. Development of a system model for advanced small modular reactors.

    SciTech Connect

    Lewis, Tom Goslee,; Holschuh, Thomas Vernon,

    2014-01-01

    This report describes a system model that can be used to analyze three advance small modular reactor (SMR) designs through their lifetime. Neutronics of these reactor designs were evaluated using Monte Carlo N-Particle eXtended (MCNPX/6). The system models were developed in Matlab and Simulink. A major thrust of this research was the initial scoping analysis of Sandias concept of a long-life fast reactor (LLFR). The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional light water reactors (LWRs) or other SMR designs (e.g. high temperature gas reactor (HTGR)). The system model has subroutines for lifetime reactor feedback and operation calculations, thermal hydraulic effects, load demand changes and a simplified SCO2 Brayton cycle for power conversion.

  20. Generation and characterization of gas bubbles in liquid metals

    SciTech Connect

    Eckert, S.; Gerbeth, G.; Witke, W.

    1996-06-01

    There is an ongoing research performed in the RCR on local transport phenomena in turbulent liquid metal (LM) duct flows exposed to external magnetic fields. In this context so-called MHD flow phenomena can be observed, which are unknown in usual hydraulic engineering. The field of interest covers also the influence of magnetic fields on the behaviour of liquid metal - gas mixtures. Profound knowledge on these LMMHD two-phase flow plays an important role in a variety of technological applications, in particular, in the design of Liquid-Metal MHD generators or for several metallurgical processes employing gas-stirred reactors. However, the highly empirical nature of two-phase flow analysis gives little hope for the prediction of MHD two-phase flows without extensive experimental data. A summary is given about the authors research activities focussing on two directions: (a) Momentum transfer between gas and liquid metal in a bubbly flow regime to investigate the influence of the external magnetic field on the velocity slip ration S (b) Peculiarities of the MHD turbulence to use small gas bubbles as local tracers in order to study the turbulent mass transfer.

  1. Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    A. Joseph Palmer; David A. Petti; S. Blaine Grover

    2014-04-01

    The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

  2. High temperature reactors

    NASA Astrophysics Data System (ADS)

    Dulera, I. V.; Sinha, R. K.

    2008-12-01

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements.

  3. ANDES Measurements for Advanced Reactor Systems

    NASA Astrophysics Data System (ADS)

    Plompen, A. J. M.; Hambsch, F.-J.; Kopecky, S.; Nyman, M.; Rouki, C.; Salvador Castiñeira, P.; Schillebeeckx, P.; Belloni, F.; Berthoumieux, E.; Gunsing, F.; Lampoudis, C.; Calviani, M.; Guerrero, C.; Cano-Ott, D.; Gonzalez Romero, E.; Aïche, M.; Jurado, B.; Mathieu, L.; Derckx, X.; Farget, F.; Rodrigues Tajes, C.; Bacquias, A.; Dessagne, Ph.; Kerveno, M.; Borcea, C.; Negret, A.; Colonna, N.; Goncalves, I.; Penttilä, H.; Rinta-Antila, S.; Kolhinen, V. S.; Jokinen, A.

    2014-05-01

    A significant number of new measurements was undertaken by the ANDES “Measurements for advanced reactor systems” initiative. These new measurements include neutron inelastic scattering from 23Na, Mo, Zr, and 238U, neutron capture cross sections of 238U, 241Am, neutron induced fission cross sections of 240Pu, 242Pu, 241Am, 243Am and 245Cm, and measurements that explore the limits of the surrogate technique. The latter study the feasibility of inferring neutron capture cross sections for Cm isotopes, the neutron-induced fission cross section of 238Pu and fission yields and fission probabilities through full Z and A identification in inverse kinematics for isotopes of Pu, Am, Cm and Cf. Finally, four isotopes are studied which are important to improve predictions for delayed neutron precursors and decay heat by total absorption gamma-ray spectrometry (88Br, 94Rb, 95Rb, 137I). The measurements which are performed at state-of-the-art European facilities have the ambition to achieve the lowest possible uncertainty, and to come as close as is reasonably achievable to the target uncertainties established by sensitivity studies. An overview is presented of the activities and achievements, leaving detailed expositions to the various parties contributing to the conference.

  4. DOE/NE robotics for advanced reactors

    SciTech Connect

    Not Available

    1991-01-01

    ORNL is continuing to transfer technology under the NE Robotics for Advanced Reactors Program to other programs and institutions. The HELIX computer architecture developed under the NE robotics program are currently being transferred to the University of Florida for control of their Articulated Transporter Manipulator (ATMS or snake) robot prototype. The Modular Integrated Control architecture (MICA), also developed under the NE Robotics Program is currently being utilized by the DOE ER WM Robotics Program to control a long-reach manipulator for use in cleanup of underground storage tanks at Hanford. HELIX and MICA were developed at ORNL in response to the need to integrate software and hardware developed on different computer systems by the university participants. The University of Texas is constructing and assembling the actuator module. Several parts have been re-machined. The Tennessee group has performed a series of experiments to verify the performance of their system for object localization, identification, and manipulation using 3D Geometric models (SOLIDGEM). The Michigan group has completed experiments on the radiation resistance of ultrasonic range sensors (URSs). The Florida team continues to model key features of the ALMR. Drawings have been received from GE providing much greater detail on many structures. These details are being incorporated into the IGRIP model of the ALMR. A conceptual design for the Articulated Transporter/Manipulator System (ATMS) has been finalized.

  5. Bearing for liquid metal pump

    DOEpatents

    Dickinson, Robert J.; Wasko, John; Pennell, William E.

    1984-01-01

    A liquid metal pump bearing support comprises a series of tangentially oriented spokes that connect the bearing cylinder to the pump internals structure. The spokes may be arranged in a plurality of planes extending from the bearing cylinder to the pump internals with the spokes in one plane being arranged alternately with those in the next plane. The bearing support structure provides the pump with sufficient lateral support for the bearing structure together with the capability of accommodating differential thermal expansion without adversely affecting pump performance.

  6. Shielding considerations for advanced space nuclear reactor systems

    NASA Astrophysics Data System (ADS)

    Angelo, J. P., Jr.; Buden, D.

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO2) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The status of this advanced heat pipe reactor is reviewed and the radiation environments and shielding requirements for representative manned and unmanned applications are explored.

  7. Shielding considerations for advanced space nuclear reactor systems

    SciTech Connect

    Angelo, J.P. Jr.; Buden, D.

    1982-01-01

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

  8. Conceptual design of the advanced marine reactor MRX

    NASA Astrophysics Data System (ADS)

    1991-02-01

    Design studies on the advanced marine reactors have been done continuously since 1983 at the Japan Atomic Energy Research Institute (JAERI) in order to develop attractive marine reactors for the next generation. At present, two marine reactor concepts are being formulated. One is 100 MWt MRX (Marine Reactor X) for an icebreaker and the other is 300 kWe DRX (Deep-sea Reactor X) for a deep-sea research vessel. They are characterized by an integral type pressurized water reactor (PWR) built-in type control rod drive mechanisms, a water-filled container and a passive decay heat removal system, which realize highly passive safe and compact reactors. This paper is a detailed report including all major results of the MRX design study.

  9. Transformable liquid-metal nanomedicine

    NASA Astrophysics Data System (ADS)

    Lu, Yue; Hu, Quanyin; Lin, Yiliang; Pacardo, Dennis B.; Wang, Chao; Sun, Wujin; Ligler, Frances S.; Dickey, Michael D.; Gu, Zhen

    2015-12-01

    To date, numerous inorganic nanocarriers have been explored for drug delivery systems (DDSs). However, the clinical application of inorganic formulations has often been hindered by their toxicity and failure to biodegrade. We describe here a transformable liquid-metal nanomedicine, based on a core-shell nanosphere composed of a liquid-phase eutectic gallium-indium core and a thiolated polymeric shell. This formulation can be simply produced through a sonication-mediated method with bioconjugation flexibility. The resulting nanoparticles loaded with doxorubicin (Dox) have an average diameter of 107 nm and demonstrate the capability to fuse and subsequently degrade under a mildly acidic condition, which facilitates release of Dox in acidic endosomes after cellular internalization. Equipped with hyaluronic acid, a tumour-targeting ligand, this formulation displays enhanced chemotherapeutic inhibition towards the xenograft tumour-bearing mice. This liquid metal-based DDS with fusible and degradable behaviour under physiological conditions provides a new strategy for engineering theranostic agents with low toxicity.

  10. Liquid-Metal Pump Technologies for Nuclear Surface Power

    NASA Technical Reports Server (NTRS)

    Polzin, K. A.

    2007-01-01

    Multiple liquid-metal pump options are reviewed for the purpose of determining the technologies that are best suited for inclusion in a nuclear reactor thermal simulator intended to test prototypical space nuclear system components. Conduction, induction, and thermoelectric electromagnetic pumps are evaluated based on their performance characteristics and the technical issues associated with incorporation into a reactor system. The thermoelectric pump is recommended for inclusion in the planned system at NASA MSFC based on its relative simplicity, low power supply mass penalty, flight heritage, and the promise of increased pump efficiency over earlier flight pump designs through the use of skutterudite thermoelectric elements.

  11. Liquid metal Flow Meter - Final Report

    SciTech Connect

    Andersen, C.; Hoogendoom, S.; Hudson, B.; Prince, J.; Teichert, K.; Wood, J.; Chase, K.

    2007-01-30

    Measuring the flow of liquid metal presents serious challenges. Current commercially-available flow meters use ultrasonic, electromagnetic, and other technologies to measure flow, but are inadequate for liquid metal flow measurement because of the high temperatures required by most liquid metals. As a result of the reactivity and high temperatures of most liquid metals, corrosion and leakage become very serious safety concerns. The purpose of this project is to develop a flow meter for Lockheed Martin that measures the flow rate of molten metal in a conduit.

  12. Technique for detecting liquid metal leaks

    DOEpatents

    Bauerle, James E.

    1979-01-01

    In a system employing flowing liquid metal as a heat transfer medium in contact with tubular members containing a working fluid, i.e., steam, liquid metal leaks through the wall of the tubular member are detected by dislodging the liquid metal compounds forming in the tubular member at the leak locations and subsequently transporting the dislodged compound in the form of an aerosol to a detector responsive to the liquid metal compound. In the application to a sodium cooled tubular member, the detector would consist of a sodium responsive device, such as a sodium ion detector.

  13. Reactor assessments of advanced bumpy torus configurations

    SciTech Connect

    Uckan, N.A.; Owen, L.W.; Spong, D.A.; Miller, R.L.; Ard, W.B.; Pipkins, J.F.; Schmitt, R.J.

    1984-02-01

    Recently, several innovative approaches were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator - snakey torus). Preliminary evaluations of reactor implications of each approach have been carried out based on magnetics (vacuum) calculations, transport and scaling relationships, and stability properties deduced from provisional configurations that implement the approach but are not necessarily optimized. Further optimization is needed in all cases to evaluate the full potential of each approach. Results of these studies indicate favorable reactor projections with a significant reduction in reactor physical size as compared to conventional EBT reactor designs carried out in the past.

  14. Safety analysis of the advanced thermionic initiative reactor

    NASA Astrophysics Data System (ADS)

    Lee, Hsing H.; Klein, Andrew C.

    1995-01-01

    Previously, detailed analysis was conducted to assess the technology developed for the Advanced Thermionic Initiative reactor. This analysis included the development of an overall system design code capability and the improvement of analytical models necessary for the assessment of the use of single cell thermionic fuel elements in a low power space nuclear reactor. The present analysis extends this effort to assess the nuclear criticality safety of the ATI reactor for various different scenarios. The analysis discusses the efficacy of different methods of reactor control such as control rods, and control drums.

  15. 78 FR 66968 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-07

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor...

  16. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-03

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor...

  17. 78 FR 46621 - Status of the Office of New Reactors' Implementation of Electronic Distribution of Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-08-01

    ... COMMISSION Status of the Office of New Reactors' Implementation of Electronic Distribution of Advanced Reactor Correspondence AGENCY: Nuclear Regulatory Commission. ACTION: Implementation of electronic distribution of advanced reactor correspondence; issuance. SUMMARY: The U.S. Nuclear Regulatory Commission...

  18. Advanced Plasma Pyrolysis Assembly (PPA) Reactor and Process Development

    NASA Technical Reports Server (NTRS)

    Wheeler, Richard R., Jr.; Hadley, Neal M.; Dahl, Roger W.; Abney, Morgan B.; Greenwood, Zachary; Miller, Lee; Medlen, Amber

    2012-01-01

    Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA's Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development.

  19. Extension of the supercritical carbon dioxide brayton cycle to low reactor power operation: investigations using the coupled anl plant dynamics code-SAS4A/SASSYS-1 liquid metal reactor code system.

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2012-05-10

    Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle enabling removal of power from an autonomous load following Sodium-Cooled Fast Reactor (SFR) down to decay heat levels such that the S-CO{sub 2} cycle can be used to cool the reactor until decay heat can be removed by the normal shutdown heat removal system or a passive decay heat removal system such as Direct Reactor Auxiliary Cooling System (DRACS) loops with DRACS in-vessel heat exchangers. This capability of the new control strategy eliminates the need for use of a separate shutdown heat removal system which might also use supercritical CO{sub 2}. It has been found that this capability can be achieved by introducing a new control mechanism involving shaft speed control for the common shaft joining the turbine and two compressors following reduction of the load demand from the electrical grid to zero. Following disconnection of the generator from the electrical grid, heat is removed from the intermediate sodium circuit through the sodium-to-CO{sub 2} heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO{sub 2}-to-water cooler. To investigate the effectiveness of shaft speed control, calculations are carried out using the coupled Plant Dynamics Code-SAS4A/SASSYS-1 code for a linear load reduction transient for a 1000 MWt metallic-fueled SFR with autonomous load following. No deliberate motion of control rods or adjustment of sodium pump speeds is assumed to take place. It is assumed that the S-CO{sub 2} turbomachinery shaft speed linearly decreases from 100 to 20% nominal following reduction of grid load to zero. The reactor power is calculated to autonomously decrease down to 3% nominal providing a lengthy window in time for the switchover to the normal shutdown heat removal system or for a passive decay heat removal system to become effective. However, the

  20. Sewage sludge dewatering using flowing liquid metals

    DOEpatents

    Carlson, Larry W.

    1986-01-01

    A method and apparatus for reducing the moisture content of a moist sewage sludge having a moisture content of about 50% to 80% and formed of small cellular micro-organism bodies having internally confined water is provided. A hot liquid metal is circulated in a circulation loop and the moist sewage sludge is injected in the circulation loop under conditions of temperature and pressure such that the confined water vaporizes and ruptures the cellular bodies. The vapor produced, the dried sludge, and the liquid metal are then separated. Preferably, the moist sewage sludge is injected into the hot liquid metal adjacent the upstream side of a venturi which serves to thoroughly mix the hot liquid metal and the moist sewage sludge. The venturi and the drying zone after the venturi are preferably vertically oriented. The dried sewage sludge recovered is available as a fuel and is preferably used for heating the hot liquid metal.

  1. Advanced Reactors Thermal Energy Transport for Process Industries

    SciTech Connect

    P. Sabharwall; S.J. Yoon; M.G. McKellar; C. Stoots; George Griffith

    2014-07-01

    The operation temperature of advanced nuclear reactors is generally higher than commercial light water reactors and thermal energy from advanced nuclear reactor can be used for various purposes such as liquid fuel production, district heating, desalination, hydrogen production, and other process heat applications, etc. Some of the major technology challenges that must be overcome before the advanced reactors could be licensed on the reactor side are qualification of next generation of nuclear fuel, materials that can withstand higher temperature, improvement in power cycle thermal efficiency by going to combined cycles, SCO2 cycles, successful demonstration of advanced compact heat exchangers in the prototypical conditions, and from the process side application the challenge is to transport the thermal energy from the reactor to the process plant with maximum efficiency (i.e., with minimum temperature drop). The main focus of this study is on doing a parametric study of efficient heat transport system, with different coolants (mainly, water, He, and molten salts) to determine maximum possible distance that can be achieved.

  2. SOAR: Space Orbiting Advanced Fusion Power Reactor.

    DTIC Science & Technology

    1987-09-01

    this are the possibility of electrostatic direct conversion and the ease of increasing magnetic fields to maintain plasma pressure . The gene- ral...106 0.2 29 290 Known Reserves 3 x 103 0.2 187 1870 MAN-MADE U.S. DOE MRC Sales 1.3/yr MRC Inventory 13.4 CANDU Reactors (by year 2000) Production 2/yr...Research Corp. currently sells up to 1.3 kg/yr and has a 13 kg inventory. The Canadian CANDU reactors produce tritium in their D20 moderators. By 1987, 5

  3. The movement of particles in liquid metals under gravity forces and the interaction of particles with advancing solid-liquid interface

    NASA Technical Reports Server (NTRS)

    Weinberg, F.

    1984-01-01

    The problems of shrinkage and gas porosity are discussed. Gravity forces enhance the removal of gas bubbles from a metal melt and contribute to the feeding of shrinkage porosity in castings. Experiments are reviewed which determine how large a density difference is required for metal particles to float or sink in a metal melt and to what extent do factors not considered in Stokes Law influence particle movement in a real system. As to the interaction of particles with an advancing solid-liquid interface, the results indicate that the metal particles are not rejected in a metal melt, and that concentrations of particles in a metal following solidification are due to other factors.

  4. Space- and time-resolved resistive measurements of liquid metal wall thickness

    NASA Astrophysics Data System (ADS)

    Mirhoseini, S. M. H.; Volpe, F. A.

    2016-11-01

    In a fusion reactor internally coated with liquid metal, it will be important to diagnose the thickness of the liquid at various locations in the vessel, as a function of time, and possibly respond to counteract undesired bulging or depletion. The electrical conductance between electrodes immersed in the liquid metal can be used as a simple proxy for the local thickness. Here a matrix of electrodes is shown to provide spatially and temporally resolved measurements of liquid metal thickness in the absence of plasma. First a theory is developed for m × n electrodes, and then it is experimentally demonstrated for 3 × 1 electrodes, as the liquid stands still or is agitated by means of a shaker. The experiments were carried out with Galinstan, but are easily extended to lithium or other liquid metals.

  5. 76 FR 34276 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-06-13

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR..., Chief, Reactor Safety Branch A, Advisory Committee on Reactor Safeguards. BILLING CODE 7590-01-P...

  6. Catalysis and surface properties of liquid metals and alloys

    SciTech Connect

    Ogino, Y.

    1987-01-01

    This book presents a historical review of the subject in order to clarify its own role in advancing the study of heterogenous catalysis. In addition, this authoritative volume; discusses the catalytic properties of liquid metals and alloys, giving a useful, schematic account of various experimental techniques; examines the mechanism of catalysis at the atomic and particle levels, defining the structures of liquid metals; covers a variety of reactions, including dehydrogenation, hydrogen transfer, coal liquefaction, and other, suggesting practical uses and additional areas for investigation; addresses basic and microscopic aspects of catalysis, exploring such advanced topics as kinetics and stereochemistry as well as optical properties and surface transition zones; and provides examples of applications, illustrating fundamental research with specific technologies that extend the range of future research possibilities.

  7. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    SciTech Connect

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-12-31

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR`s) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design.

  8. Surface chemistry of liquid metals

    NASA Technical Reports Server (NTRS)

    Mann, J. Adin, Jr.; Peebles, Henry; Peebles, Diamond; Rye, Robert; Yost, Fred

    1993-01-01

    The fundamental surface chemistry of the behavior of liquid metals spreading on a solid substrate is not at all well understood. Each of these questions involves knowing the details of the structure of interfaces and their dynamics. For example the structure of a monolayer of tin oxide on pure liquid tin is unknown. This is in contrast to the relatively large amount of data available on the structure of copper oxide monolayers on solid, pure copper. However, since liquid tin has a vapor pressure below 10(exp -10)torr for a reasonable temperature range above its melting point, it is possible to use the techniques of surface science to study the geometric, electronic and vibrational structures of these monolayers. In addition, certain techniques developed by surface chemists for the study of liquid systems can be applied to the ultra-high vacuum environment. In particular we have shown that light scattering spectroscopy can be used to study the surface tension tensor of these interfaces. The tin oxide layer in particular is very interesting in that the monolayer is rigid but admits of bending. Ellipsometric microscopy allows the visualization of monolayer thick films and show whether island formation occurs at various levels of dosing.

  9. Advanced development of immobilized enzyme reactors

    NASA Technical Reports Server (NTRS)

    Jolly, Clifford D.; Schussel, Leonard J.; Carter, Layne

    1991-01-01

    Fixed-bed reactors have been used at NASA-Marshall to purify wastewater generated by an end-use equipment facility, on the basis of a combination of multifiltration unibeds and enzyme unibeds. The enzyme beds were found to effectively remove such targeted organics as urea, alcohols, and aldehydes, down to levels lying below detection limits. The enzyme beds were also found to remove organic contaminants not specifically targeted.

  10. Development of a liquid metal slip ring

    NASA Technical Reports Server (NTRS)

    Weinberger, S. M.

    1972-01-01

    A liquid metal slip ring/solar orientation mechanism was designed and a model tested. This was a follow-up of previous efforts for the development of a gallium liquid metal slip ring in which the major problem was the formation and ejection of debris. A number of slip ring design approaches were studied. The probe design concept was fully implemented with detail drawings and a model was successfully tested for dielectric strength, shock vibration, acceleration and operation. The conclusions are that a gallium liquid metal slip ring/solar orientation mechanism is feasible and that the problem of debris formation and ejection has been successfully solved.

  11. Liquid metal ion source and alloy

    DOEpatents

    Clark, Jr., William M.; Utlaut, Mark W.; Behrens, Robert G.; Szklarz, Eugene G.; Storms, Edmund K.; Santandrea, Robert P.; Swanson, Lynwood W.

    1988-10-04

    A liquid metal ion source and alloy, wherein the species to be emitted from the ion source is contained in a congruently vaporizing alloy. In one embodiment, the liquid metal ion source acts as a source of arsenic, and in a source alloy the arsenic is combined with palladium, preferably in a liquid alloy having a range of compositions from about 24 to about 33 atomic percent arsenic. Such an alloy may be readily prepared by a combustion synthesis technique. Liquid metal ion sources thus prepared produce arsenic ions for implantation, have long lifetimes, and are highly stable in operation.

  12. Liquid Metal Pump Technologies for Nuclear Surface Power

    NASA Technical Reports Server (NTRS)

    Polzin, Kurt A.

    2007-01-01

    Multiple liquid metal pump options are reviewed for the purpose of determining the technologies that are best suited for inclusion in a nuclear reactor thermal simulator intended to rest prototypical space nuclear surface power system components. Conduction, induction and thermoelectric electromagnetic pumps are evaluated based on their performance characteristics and the technical issues associated with incorporation into a reactor system. A thermoelectric electromagnetic pump is selected as the best option for use in NASA-MSFC's Fission Surface Power-Primary Test Circuit reactor simulator based on its relative simplicity, low power supply mass penalty, flight heritage, and the promise of increased pump efficiency over those earlier pump designs through the use of skutterudite thermoelectric elements.

  13. Liquid metal cooled divertor for ARIES

    SciTech Connect

    Muraviev, E.

    1995-01-01

    A liquid metal, Ga-cooled divertor design was completed for the double null ARIES-II divertor design. The design analysis indicated a surface heat flux removal capability of up to 15 MW/m{sup 2}, and its relative easy maintenance. Design issues of configuration, thermal hydraulics, thermal stresses, liquid metal loop and safety effects were evaluated. For coolant flow control, it was found that it is necessary to use some part of the blanket cooling ducts for the draining of liquid metal from the top divertor. In order to minimize the inventory of Ga, it was recommended that the liquid metal loop equipment should be located as close to the torus as possible. More detailed analysis of transient conditions especially under accident conditions was identified as an issue that will need to be addressed.

  14. Sewage sludge dewatering using flowing liquid metals

    SciTech Connect

    Carlson, L.W.

    1986-11-04

    A method is described of reducing the moisture content of a moist sewage sludge having a moisture content of about 50-80% and formed of small cellular micro-organism bodies having internally confined water. The method comprises: circulating a hot liquid metal in a loop; forming a mixture of the moist sludge and the hot liquid metal in a portion of the loop under conditions of temperature and pressure such that the confined water vaporizes and ruptures the cellular bodies; separating the liquid metal, dried sludge, and vaporized water in a separation zone of the loop; and drawing off the dried sludge and vaporized water from the loop whereby the liquid metal is left to be recirculated in the loop.

  15. Advanced sodium fast reactor accident source terms :

    SciTech Connect

    Powers, Dana Auburn; Clement, Bernard; Denning, Richard; Ohno, Shuji; Zeyen, Roland

    2010-09-01

    An expert opinion elicitation has been used to evaluate phenomena that could affect releases of radionuclides during accidents at sodium-cooled fast reactors. The intent was to identify research needed to develop a mechanistic model of radionuclide release for licensing and risk assessment purposes. Experts from the USA, France, the European Union, and Japan identified phenomena that could affect the release of radionuclides under hypothesized accident conditions. They qualitatively evaluated the importance of these phenomena and the need for additional experimental research. The experts identified seven phenomena that are of high importance and have a high need for additional experimental research: High temperature release of radionuclides from fuel during an energetic event Energetic interactions between molten reactor fuel and sodium coolant and associated transfer of radionuclides from the fuel to the coolant Entrainment of fuel and sodium bond material during the depressurization of a fuel rod with breached cladding Rates of radionuclide leaching from fuel by liquid sodium Surface enrichment of sodium pools by dissolved and suspended radionuclides Thermal decomposition of sodium iodide in the containment atmosphere Reactions of iodine species in the containment to form volatile organic iodides. Other issues of high importance were identified that might merit further research as development of the mechanistic model of radionuclide release progressed.

  16. Tokamak with liquid metal toroidal field coil

    DOEpatents

    Ohkawa, Tihiro; Schaffer, Michael J.

    1981-01-01

    Tokamak apparatus includes a pressure vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within the pressure vessel defines a toroidal space within the liner. Liquid metal fills the reservoir outside said liner. Electric current is passed through the liquid metal over a conductive path linking the toroidal space to produce a toroidal magnetic field within the toroidal space about the major axis thereof. Toroidal plasma is developed within the toroidal space about the major axis thereof.

  17. Solar driven liquid metal MHD power generator

    NASA Technical Reports Server (NTRS)

    Lee, J. H.; Hohl, F. (Inventor)

    1983-01-01

    A solar energy collector focuses solar energy onto a solar oven which is attached to a mixer which in turn is attached to the channel of a MHD generator. Gas enters the oven and a liquid metal enters the mixer. The gas/liquid metal mixture is heated by the collected solar energy and moves through the MHD generator thereby generating electrical power. The mixture is then separated and recycled.

  18. Foundational development of an advanced nuclear reactor integrated safety code.

    SciTech Connect

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  19. Health Monitoring to Support Advanced Small Modular Reactors

    SciTech Connect

    Coble, Jamie B.; Meyer, Ryan M.; Ramuhalli, Pradeep

    2013-08-01

    Advanced small modular reactors (aSMRs) are based on advanced reactor concepts, some of which were promoted by the Generation IV International Forum, and are being considered for diverse missions including desalination of water, production of hydrogen, etc. While the existing fleet of commercial nuclear reactors provides baseload electricity, it is conceivable that aSMRs could be implemented for both baseload and load following applications. The effect of diverse operating missions and unit modularity on plant operations and maintenance (O&M) is not fully understood and limiting these costs will be essential to successful deployment of aSMRs. Integrated health monitoring concepts are proposed to support the safe and affordable operation of aSMRs over their lifetime by enabling management of significant in-vessel and in-containment active and passive components.

  20. Liquid Metal Walls and the Belt Pinch

    NASA Astrophysics Data System (ADS)

    Kotschenreuther, Michael; Dorland, W.; Manickam, J.; Menard, J.; Miller, R.

    2000-10-01

    High flux reactor designs have placed the MHD stabilizing shell well away from the first wall due to breeding and reliability issues. Liquid metal shells containing Li may allow close shells, with much higher elongation (kappa), beta and power density. MHD stability of equilibria with kappa = 2 - 6 and high bootstrap fraction are examined using JSOLVER plus PEST and TOQ plus BALOO. Compared to kappa = 2, stable beta increases by 2.5 for kappa = 3 and 5 for kappa = 6 (with little change in normalized beta). External mode stability (n = 1-10 ) with a wall is similar for kappa = 2-6; a wall at b/a 1.2-1.3 give stability. Resistive wall mode evolution is examined using the new code WALLMODE; feedback power is evaluated using dynamic Monte Carlo simulations (similar to ARIES studies). N = 0 vertical modes are acceptable with a 2-4 cm Li shell at b/a 1.05 -1.1. Resistive wall kink modes for model current profiles are stabilizable by feedback and/or liquid flow as low as 20 m/s; interfacing with PEST is in progress to examinine realistic equilibria. Modest indendation gives bootstrap island stability. Comprehensive gyrokinetic simulations with GS2 find a strong improvement with kappa for 1) ExB shearing compared to ITG/drift mode growth rates and 2) transport in nonlinear ETG simulations. Also, low edge recycling (density) boundary conditions substantially improve H mode pedestal stabilty, for additional global confinement improvement. Even with H mode scaling law confinement, kappa =6 gives ignition in a .8 m major radius device with Cu coils (12T at the coil).

  1. 78 FR 37595 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-06-21

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR.... Kathy Weaver, Acting Chief, Technical Support Branch, Advisory Committee on Reactor Safeguards....

  2. 76 FR 5218 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-28

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor... inconvenience. Dated: January 24, 2011. Antonio Dias, Chief, Reactor Safety Branch B, Advisory Committee...

  3. 77 FR 59678 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-09-28

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... Branch, Advisory Committee on Reactor Safeguards. BILLING CODE 7590-01-P...

  4. 78 FR 20959 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-04-08

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR) will hold a meeting.... Antonio Dias, Technical Advisor, Advisory Committee on Reactor Safeguards. BILLING CODE 7590-01-P...

  5. Multi-Purpose Thermal Hydraulic Loop: Advanced Reactor Technology Integral System Test (ARTIST) Facility for Support of Advanced Reactor Technologies

    SciTech Connect

    James E. O'Brien; Piyush Sabharwall; SuJong Yoon

    2001-11-01

    Effective and robust high temperature heat transfer systems are fundamental to the successful deployment of advanced reactors for both power generation and non-electric applications. Plant designs often include an intermediate heat transfer loop (IHTL) with heat exchangers at either end to deliver thermal energy to the application while providing isolation of the primary reactor system. In order to address technical feasibility concerns and challenges a new high-temperature multi-fluid, multi-loop test facility “Advanced Reactor Technology Integral System Test facility” (ARTIST) is under development at the Idaho National Laboratory. The facility will include three flow loops: high-temperature helium, molten salt, and steam/water. Details of some of the design aspects and challenges of this facility, which is currently in the conceptual design phase, are discussed

  6. 76 FR 18585 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-04-04

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water...

  7. 76 FR 11524 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-02

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactors (ABWR); Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water...

  8. Metal fire implications for advanced reactors. Part 1, literature review.

    SciTech Connect

    Nowlen, Steven Patrick; Radel, Ross F.; Hewson, John C.; Olivier, Tara Jean; Blanchat, Thomas K.

    2007-10-01

    Public safety and acceptance is extremely important for the nuclear power renaissance to get started. The Advanced Burner Reactor and other potential designs utilize liquid sodium as a primary coolant which provides distinct challenges to the nuclear power industry. Fire is a dominant contributor to total nuclear plant risk events for current generation nuclear power plants. Utilizing past experience to develop suitable safety systems and procedures will minimize the chance of sodium leaks and the associated consequences in the next generation. An advanced understanding of metal fire behavior in regards to the new designs will benefit both science and industry. This report presents an extensive literature review that captures past experiences, new advanced reactor designs, and the current state-of-knowledge related to liquid sodium combustion behavior.

  9. The Consortium for Advanced Simulation of Light Water Reactors

    SciTech Connect

    Ronaldo Szilard; Hongbin Zhang; Doug Kothe; Paul Turinsky

    2011-10-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

  10. Investigation of Liquid Metal Heat Exchanger Designs for Fission Surface Power

    NASA Technical Reports Server (NTRS)

    Dyson, Rodger W.; Penswick, Barry; Robbie, Malcolm; Geng, Steven M.

    2009-01-01

    Fission surface power is an option for future Moon and Mars surface missions. High power nuclear reactor heated Stirling convertors are an option to provide reliable power for long duration outpost operations. This report investigates various design approaches for the liquid metal to acceptor heat exchange and clarifies the details used in the analysis.

  11. Johnson Noise Thermometry for Advanced Small Modular Reactors

    SciTech Connect

    Britton, C.L.,Jr.; Roberts, M.; Bull, N.D.; Holcomb, D.E.; Wood, R.T.

    2012-09-15

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor’s physical condition. In and near the core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of the current ORNL-led project, conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced SMR Research and Development (R&D) program, is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  12. Integrated intelligent systems in advanced reactor control rooms

    SciTech Connect

    Beckmeyer, R.R.

    1989-01-01

    An intelligent, reactor control room, information system is designed to be an integral part of an advanced control room and will assist the reactor operator's decision making process by continuously monitoring the current plant state and providing recommended operator actions to improve that state. This intelligent system is an integral part of, as well as an extension to, the plant protection and control systems. This paper describes the interaction of several functional components (intelligent information data display, technical specifications monitoring, and dynamic procedures) of the overall system and the artificial intelligence laboratory environment assembled for testing the prototype. 10 refs., 5 figs.

  13. The stress analysis of a heavy liquid metal pump impeller

    NASA Astrophysics Data System (ADS)

    Ma, X. D.; Li, X. L.; Zhu, Z. Q.; Li, C. J.; Gao, S.

    2016-05-01

    Lead-based coolant reactor is a promising Generation-IV reactor. In the lead-based coolant reactor, the coolant is liquid lead or lead-bismuth eutectic. The main pump in the reactor is a very important device. It supplies force for the coolant circulation. The liquid metal has a very large density which is about ten times of the water. Also, the viscosity of the coolant is small which is about one sixth of the water. When the pump transports heavy liquid, the blade loading is heavy. The large force can cause the failure of the blade when the fatigue stress exceeds the allowable stress. The impeller fraction is a very serious accident which is strictly prohibited in the nuclear reactor. In this paper, the numerical method is used to simulate the flow field of a heavy liquid metal pump. The SST k-w turbulent model is used in the calculation to get a more precise flow structure. The hydraulic force is obtained with the one way fluid solid coupling. The maximum stress in the impeller is analyzed. The stress in the liquid metal pump is compared with that in the water pump. The calculation results show that the maximum stress of the impeller blade increases with increase of flow rate. In the design of the impeller blade thickness, the impeller strength in large operating condition should be considered. The maximum stress of the impeller blade located in the middle and near the hub of the leading edge. In this position, the blade is easy to fracture. The maximum deformation of the impeller firstly increase with increase of flow rate and then decrease with increase of flow rate. The maximum deformation exists in the middle of the leading edge when in small flow rate and in the out radius of the impeller when in large flow rate. Comparing the stress of the impeller when transporting water and LBE, the maximum stress is almost one-tenth of that in the LBE impeller which is the same ratio of the density. The static stress in different medium is proportional to the pressure

  14. Double-duct liquid metal magnetohydrodynamic engine

    DOEpatents

    Haaland, Carsten M.

    1997-01-01

    An internal combustion, liquid metal (LM) magnetohydrodynamic (MHD) engine and an alternating current (AC) magnetohydrodynamic generator, are used in combination to provide useful AC electric energy output. The engine design has four pistons and a double duct configuration, with each duct containing sodium potassium liquid metal confined between free pistons located at either end of the duct. The liquid metal is forced to flow back and forth in the duct by the movement of the pistons, which are alternatively driven by an internal combustion process. In the MHD generator, the two LM-MHD ducts pass in close proximity through a Hartmann duct with output transformer. AC power is produced by operating the engine with the liquid metal in the two generator ducts always flowing in counter directions. The amount of liquid metal maintained in the ducts may be varied. This provides a variable stroke length for the pistons. The engine/generator provides variable AC power at variable frequencies that correspond to the power demands of the vehicular propulsion. Also the engine should maintain nearly constant efficiency throughout the range of power usage. Automobiles and trucks could be powered by the invention, with no transmission or power converter devices being required.

  15. Double-duct liquid metal magnetohydrodynamic engine

    DOEpatents

    Haaland, Carsten M.

    1995-01-01

    An internal combustion, liquid metal (LM) magnetohydrodynamic (MHD) engine and an alternating current (AC) magnetohydrodynamic generator, are used in combination to provide useful AC electric energy output. The engine design has-four pistons and a double duct configuration, with each duct containing sodium potassium liquid metal confined between free pistons located at either end of the duct. The liquid metal is forced to flow back and forth in the duct by the movement of the pistons, which are alternatively driven by an internal combustion process. In the MHD generator, the two LM-MHD ducts pass in close proximity through a Hartmann duct with output transformer. AC power is produced by operating the engine with the liquid metal in the two generator ducts always flowing in counter directions. The amount of liquid metal maintained in the ducts may be varied. This provides a variable stroke length for the pistons. The engine/generator provides variable AC power at variable frequencies that correspond to the power demands of the vehicular propulsion. Also the engine should maintain nearly constant efficiency throughout the range of power usage. Automobiles and trucks could be powered by the invention, with no transmission or power converter devices being required.

  16. An advanced carbon reactor subsystem for carbon dioxide reduction

    NASA Technical Reports Server (NTRS)

    Noyes, Gary P.; Cusick, Robert J.

    1986-01-01

    An evaluation is presented of the development status of an advanced carbon-reactor subsystem (ACRS) for the production of water and dense, solid carbon from CO2 and hydrogen, as required in physiochemical air revitalization systems for long-duration manned space missions. The ACRS consists of a Sabatier Methanation Reactor (SMR) that reduces CO2 with hydrogen to form methane and water, a gas-liquid separator to remove product water from the methane, and a Carbon Formation Reactor (CFR) to pyrolize methane to carbon and hydrogen; the carbon is recycled to the SMR, while the produce carbon is periodically removed from the CFR. A preprototype ACRS under development for the NASA Space Station is described.

  17. Advances in process intensification through multifunctional reactor engineering

    SciTech Connect

    O'Hern, T. J.

    2012-03-01

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes critical to process intensification and implementation in commercial applications. Physics of the heat and mass transfer and chemical kinetics and how these processes are ultimately scaled were investigated. Specifically, we progressed the knowledge and tools required to scale a multifunctional reactor for acid-catalyzed C4 paraffin/olefin alkylation to industrial dimensions. Understanding such process intensification strategies is crucial to improving the energy efficiency and profitability of multifunctional reactors, resulting in a projected energy savings of 100 trillion BTU/yr by 2020 and a substantial reduction in the accompanying emissions.

  18. Advanced Reactor Technology -- Regulatory Technology Development Plan (RTDP)

    SciTech Connect

    Moe, Wayne Leland

    2015-05-01

    This DOE-NE Advanced Small Modular Reactor (AdvSMR) regulatory technology development plan (RTDP) will link critical DOE nuclear reactor technology development programs to important regulatory and policy-related issues likely to impact a “critical path” for establishing a viable commercial AdvSMR presence in the domestic energy market. Accordingly, the regulatory considerations that are set forth in the AdvSMR RTDP will not be limited to any one particular type or subset of advanced reactor technology(s) but rather broadly consider potential regulatory approaches and the licensing implications that accompany all DOE-sponsored research and technology development activity that deal with commercial non-light water reactors. However, it is also important to remember that certain “minimum” levels of design and safety approach knowledge concerning these technology(s) must be defined and available to an extent that supports appropriate pre-licensing regulatory analysis within the RTDP. Final resolution to advanced reactor licensing issues is most often predicated on the detailed design information and specific safety approach as documented in a facility license application and submitted for licensing review. Because the AdvSMR RTDP is focused on identifying and assessing the potential regulatory implications of DOE-sponsored reactor technology research very early in the pre-license application development phase, the information necessary to support a comprehensive regulatory analysis of a new reactor technology, and the resolution of resulting issues, will generally not be available. As such, the regulatory considerations documented in the RTDP should be considered an initial “first step” in the licensing process which will continue until a license is issued to build and operate the said nuclear facility. Because a facility license application relies heavily on the data and information generated by technology development studies, the anticipated regulatory

  19. DOE/NE robotics for advanced reactors

    SciTech Connect

    Not Available

    1991-01-01

    This document details activities during this reporting period. The Michigan group has developed, built, and tested a general purpose interface circuit for DC motors and encoders. This interface is based on an advanced microchip, the HCTL 1100 manufactured by Hewlett Packard. The HCTL 1100 can be programmed by a host computer in real-time, allowing sophisticated motion control for DC motors. At the University of Florida, work on modeling the details of the seismic isolators and the jack mechanism has been completed. A separate 3D solid view of the seismic isolator floor, with the full set of isolators shown in detail, has been constructed within IGRIP. ORNL led the robotics team at the ALMR review meeting. Discussions were held with General Electric (GE) engineers and contractors on the robotic needs for the ALMR program. The Tennessee group has completed geometric modeling of the Andros Mark VI mobile platform with two fixed tracks and for articulated tracks, the give degree-of-freedom manipulator and its end-effector, and two cameras. A graphical control of panel was developed which allow the user to operate the simulated robot. The University of Texas team visited ORNL to complete the implementation of computed-torque controller on the CESARm manipulator. This controller was previously developed and computer simulations were carried out specifically for the CESARm robot.

  20. Advanced Test Reactor - A National Scientific User Facility

    SciTech Connect

    Clifford J. Stanley

    2008-05-01

    The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected nuclear research reactor with a maximum operating power of 250 MWth. The unique serpentine configuration of the fuel elements creates five main reactor power lobes (regions) and nine flux traps. In addition to these nine flux traps there are 68 additional irradiation positions in the reactor core reflector tank. There are also 34 low-flux irradiation positions in the irradiation tanks outside the core reflector tank. The ATR is designed to provide a test environment for the evaluation of the effects of intense radiation (neutron and gamma). Due to the unique serpentine core design each of the five lobes can be operated at different powers and controlled independently. Options exist for the individual test trains and assemblies to be either cooled by the ATR coolant (i.e., exposed to ATR coolant flow rates, pressures, temperatures, and neutron flux) or to be installed in their own independent test loops where such parameters as temperature, pressure, flow rate, neutron flux, and energy can be controlled per experimenter specifications. The full-power maximum thermal neutron flux is ~1.0 x1015 n/cm2-sec with a maximum fast flux of ~5.0 x1014 n/cm2-sec. The Advanced Test Reactor, now a National Scientific User Facility, is a versatile tool in which a variety of nuclear reactor, nuclear physics, reactor fuel, and structural material irradiation experiments can be conducted. The cumulative effects of years of irradiation in a normal power reactor can be duplicated in a few weeks or months in the ATR due to its unique design, power density, and operating flexibility.

  1. Johnson Noise Thermometry for Advanced Small Modular Reactors

    SciTech Connect

    Britton Jr, Charles L; Roberts, Michael; Bull, Nora D; Holcomb, David Eugene; Wood, Richard Thomas

    2012-10-01

    Temperature is a key process variable at any nuclear power plant (NPP). The harsh reactor environment causes all sensor properties to drift over time. At the higher temperatures of advanced NPPs the drift occurs more rapidly. The allowable reactor operating temperature must be reduced by the amount of the potential measurement error to assure adequate margin to material damage. Johnson noise is a fundamental expression of temperature and as such is immune to drift in a sensor s physical condition. In and near core, only Johnson noise thermometry (JNT) and radiation pyrometry offer the possibility for long-term, high-accuracy temperature measurement due to their fundamental natures. Small, Modular Reactors (SMRs) place a higher value on long-term stability in their temperature measurements in that they produce less power per reactor core and thus cannot afford as much instrument recalibration labor as their larger brethren. The purpose of this project is to develop and demonstrate a drift free Johnson noise-based thermometer suitable for deployment near core in advanced SMR plants.

  2. Recent Advances in Pd-Based Membranes for Membrane Reactors.

    PubMed

    Arratibel Plazaola, Alba; Pacheco Tanaka, David Alfredo; Van Sint Annaland, Martin; Gallucci, Fausto

    2017-01-01

    Palladium-based membranes for hydrogen separation have been studied by several research groups during the last 40 years. Much effort has been dedicated to improving the hydrogen flux of these membranes employing different alloys, supports, deposition/production techniques, etc. High flux and cheap membranes, yet stable at different operating conditions are required for their exploitation at industrial scale. The integration of membranes in multifunctional reactors (membrane reactors) poses additional demands on the membranes as interactions at different levels between the catalyst and the membrane surface can occur. Particularly, when employing the membranes in fluidized bed reactors, the selective layer should be resistant to or protected against erosion. In this review we will also describe a novel kind of membranes, the pore-filled type membranes prepared by Pacheco Tanaka and coworkers that represent a possible solution to integrate thin selective membranes into membrane reactors while protecting the selective layer. This work is focused on recent advances on metallic supports, materials used as an intermetallic diffusion layer when metallic supports are used and the most recent advances on Pd-based composite membranes. Particular attention is paid to improvements on sulfur resistance of Pd based membranes, resistance to hydrogen embrittlement and stability at high temperature.

  3. Current Status of the Advanced High Temperature Reactor

    SciTech Connect

    Holcomb, David Eugene; Ilas, Dan; Qualls, A L; Peretz, Fred J; Varma, Venugopal Koikal; Bradley, Eric Craig; Cisneros, Anselmo T.

    2012-01-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central station type [1500 MW(e)] Fluoride salt-cooled High-temperature Reactor (FHR) that is currently under development by Oak Ridge National Laboratory for the U. S. Department of Energy, Office of Nuclear Energy's Advanced Reactor Concepts program. FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. The AHTR design option exploration is a multidisciplinary design effort that combines core neutronic and fuel configuration evaluation with structural, thermal, and hydraulic analysis to produce a reactor and vessel concept and place it within a power generation station. The AHTR design remains at the notional level of maturity, as key technologies require further development and a logically complete integrated design has not been finalized. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated.

  4. Current status of the advanced high temperature reactor

    SciTech Connect

    Holcomb, D. E.; Iias, D.; Quails, A. L.; Peretz, F. J.; Varma, V. K.; Bradley, E. C.; Cisneros, A. T.

    2012-07-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central station type [1500 MW(e)] Fluoride salt-cooled High-temperature Reactor (FHR) that is currently under development by Oak Ridge National Laboratory for the U. S. Dept. of Energy, Office of Nuclear Energy's Advanced Reactor Concepts program. FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. The AHTR design option exploration is a multidisciplinary design effort that combines core neutronic and fuel configuration evaluation with structural, thermal, and hydraulic analysis to produce a reactor and vessel concept and place it within a power generation station. The AHTR design remains at the notional level of maturity, as key technologies require further development and a logically complete integrated design has not been finalized. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. (authors)

  5. Testing of Liquid Metal Components for Nuclear Surface Power Systems

    NASA Technical Reports Server (NTRS)

    Polzin, Kurt A.; Godfroy, Thomas J.; Pearson, J. Boise

    2010-01-01

    The Early Flight Fission Test Facility (EFF-TF) was established by the Marshall Space Flight Center (MSFC) to provide a capability for performing hardware-directed activities to support multiple in-space nuclear reactor concepts by using a non-nuclear test methodology. This includes fabrication and testing at both the module/component level and near prototypic reactor configurations. The EFF-TF is currently supporting an effort to develop an affordable fission surface power (AFSP) system that could be deployed on the Lunar surface. The AFSP system is presently based on a pumped liquid metal-cooled (Sodium-Potassium eutectic, NaK-78) reactor design. This design was derived from the only fission system that the United States has deployed for space operation, the Systems for Nuclear Auxiliary Power (SNAP) 10A reactor, which was launched in 1965. Two prototypical components recently tested at MSFC were a pair of Stirling power conversion units that would be used in a reactor system to convert heat to electricity, and an annular linear induction pump (ALIP) that uses travelling electromagnetic fields to pump the liquid metal coolant through the reactor loop. First ever tests were conducted at MSFC to determine baseline performance of a pair of 1 kW Stirling convertors using NaK as the hot side working fluid. A special test rig was designed and constructed and testing was conducted inside a vacuum chamber at MSFC. This test rig delivered pumped NaK for the hot end temperature to the Stirlings and water as the working fluid on the cold end temperature. These test were conducted through a hot end temperature range between 400 to 550C in increments of 50 C and a cold end temperature range from 30 to 70 C in 20 C increments. Piston amplitudes were varied from 6 to 1 1mm in .5 mm increments. A maximum of 2240 Watts electric was produced at the design point of 550 hot end, 40 C cold end with a piston amplitude of 10.5mm. This power level was reached at a gross thermal

  6. Temperature controlled material irradiation in the advanced test reactor

    SciTech Connect

    Furstenau, R.V.; Ingrahm, F.W.

    1995-12-31

    The Advanced Test Reactor (ATR) is located at the Idaho National Engineering Laboratory (INEL) near Idaho Falls, Idaho, USA and is owned and regulated by the U.S. Department of Energy (US DOE). The ATR is operated for the US DOE by Lockheed Martin Idaho Technologies. In recent years, prime irradiation space in the ATR has been made available for use by customers having irradiation service needs in addition to the reactor`s principal user, the U.S. Naval Nuclear Propulsion Program. To enhance the reactor`s capabilities, the US DOE has initiated the development of an Irradiation Test Vehicle (ITV) capable of providing neutron spectral tailoring and temperature control for up to 28 experiments. The ATR-ITV will have the flexibility to simultaneously support a variety of experiments requiring fast, thermal or mixed spectrum neutron environments. Temperature control is accomplished by varying the thermal conductivity across a gas gap established between the experiment specimen capsule wall and the experiment `in-pile tube (IPT)` inside diameter. Thermal conductivity is adjusted by alternating the control gas mixture ratio of two gases with different thermal conductivities.

  7. Development of advanced strain diagnostic techniques for reactor environments.

    SciTech Connect

    Fleming, Darryn D.; Holschuh, Thomas Vernon,; Miller, Timothy J.; Hall, Aaron Christopher; Urrea, David Anthony,; Parma, Edward J.,

    2013-02-01

    The following research is operated as a Laboratory Directed Research and Development (LDRD) initiative at Sandia National Laboratories. The long-term goals of the program include sophisticated diagnostics of advanced fuels testing for nuclear reactors for the Department of Energy (DOE) Gen IV program, with the future capability to provide real-time measurement of strain in fuel rod cladding during operation in situ at any research or power reactor in the United States. By quantifying the stress and strain in fuel rods, it is possible to significantly improve fuel rod design, and consequently, to improve the performance and lifetime of the cladding. During the past year of this program, two sets of experiments were performed: small-scale tests to ensure reliability of the gages, and reactor pulse experiments involving the most viable samples in the Annulated Core Research Reactor (ACRR), located onsite at Sandia. Strain measurement techniques that can provide useful data in the extreme environment of a nuclear reactor core are needed to characterize nuclear fuel rods. This report documents the progression of solutions to this issue that were explored for feasibility in FY12 at Sandia National Laboratories, Albuquerque, NM.

  8. Magnetohydrodynamic effects in liquid metal batteries

    NASA Astrophysics Data System (ADS)

    Stefani, F.; Galindo, V.; Kasprzyk, C.; Landgraf, S.; Seilmayer, M.; Starace, M.; Weber, N.; Weier, T.

    2016-07-01

    Liquid metal batteries (LMBs) consist of two liquid metal electrodes and a molten salt ionic conductor sandwiched between them. The density ratios allow for a stable stratification of the three layers. LMBs were already considered as part of energy conversion systems in the 1960s and have recently received renewed interest for economical large-scale energy storage. In this paper, we concentrate on the magnetohydrodynamic aspects of this cell type with special focus on electro-vortex flows and possible effects of the Tayler instability.

  9. 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.

  10. Advanced Test Reactor National Scientific User Facility Partnerships

    SciTech Connect

    Frances M. Marshall; Todd R. Allen; Jeff B. Benson; James I. Cole; Mary Catherine Thelen

    2012-03-01

    In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

  11. The Integral Fast Reactor

    SciTech Connect

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab.

  12. Reactor applications of the Compact Fusion Advanced Rankine (CFAR) cycle for a D-T tokamak fusion reactor

    NASA Astrophysics Data System (ADS)

    Hoffman, H. A.; Logan, B. G.; Campbell, R. B.

    1988-03-01

    A preliminary design of a D-T fusion reactor blanket and MHD power conversion system is made based on the CFAR concept, and it was found that performance and costs for the reference cycle are very attractive. While much remains to be done, the potential advantage of liquid metal Rankine cycles for fusion applications are much clearer now. These include low pressures and mass flow rates, a nearly isothermal module shell which minimizes problems of thermal distortion and stresses, and an insensitivity to pressure losses in the blanket so that the two-phase MHD pressure drops in the boiling part of the blanket and the ordinary vapor pressure drops in the pebble-bed superheating zones are acceptable (the direct result of pumping a liquid rather than having to compress a gas). There are no moving parts in the high-temperature MHD power generators, no steam bottoming plant is required, only small vapor precoolers and condensers are needed because of the high heat rejection temperatures, and only a relatively small natural-draft heat exchanger is required to reject the heat to the atmosphere. The net result is a very compact fusion reactor and power conversion system which fit entirely inside an 18 meter radius reactor vault. Although a cost analysis has not yet been performed, preliminary cost estimates indicate low capital costs and a very attractive cost of electricity.

  13. Code qualification of structural materials for AFCI advanced recycling reactors.

    SciTech Connect

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L.

    2012-05-31

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the

  14. Advances in Process Intensification through Multifunctional Reactor Engineering

    SciTech Connect

    O'Hern, Timothy; Evans, Lindsay; Miller, Jim; Cooper, Marcia; Torczynski, John; Pena, Donovan; Gill, Walt

    2011-02-01

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors utilizing pulse flow. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes associated with pulse flow for implementation in commercial applications. Sandia National Laboratories (SNL) operated a pilot-scale multifunctional reactor experiment for operation with and investigation of pulse flow operation. Validation-quality data sets of the fluid dynamics, heat and mass transfer, and chemical kinetics were acquired and shared with Chemical Research and Licensing (CR&L). Experiments in a two-phase air-water system examined the effects of bead diameter in the packing, and viscosity. Pressure signals were used to detect pulsing. Three-phase experiments used immiscible organic and aqueous liquids, and air or nitrogen as the gas phase. Hydrodynamic studies of flow regimes and holdup were performed for different types of packing, and mass transfer measurements were performed for a woven packing. These studies substantiated the improvements in mass transfer anticipated for pulse flow in multifunctional reactors for the acid-catalyzed C4 paraffin/olefin alkylation process. CR&L developed packings for this alkylation process, utilizing their alkylation process pilot facilities in Pasadena, TX. These packings were evaluated in the pilot-scale multifunctional reactor experiments established by Sandia to develop a more fundamental understanding of their role in process intensification. Lummus utilized the alkylation technology developed by CR&L to design and optimize the full commercial process utilizing multifunctional reactors containing the packings developed by CR&L and evaluated by Sandia. This hydrodynamic information has been developed for multifunctional chemical reactors utilizing pulse flow, for the acid-catalyzed C4 paraffin/olefin alkylation process, and is now accessible for use in

  15. Advances in Process Intensification through Multifunctional Reactor Engineering

    SciTech Connect

    O'Hern, Timothy; Evans, Lindsay; Miller, Jim; Cooper, Marcia; Torczynski, John; Pena, Donovan; Gill, Walt; Groten, Will; Judzis, Arvids; Foley, Richard; Smith, Larry; Cross, Will; Vogt, T.

    2011-06-27

    This project was designed to advance the art of process intensification leading to a new generation of multifunctional chemical reactors utilizing pulse flow. Experimental testing was performed in order to fully characterize the hydrodynamic operating regimes associated with pulse flow for implementation in commercial applications. Sandia National Laboratories (SNL) operated a pilot-scale multifunctional reactor experiment for operation with and investigation of pulse flow operation. Validation-quality data sets of the fluid dynamics, heat and mass transfer, and chemical kinetics were acquired and shared with Chemical Research and Licensing (CR&L). Experiments in a two-phase air-water system examined the effects of bead diameter in the packing, and viscosity. Pressure signals were used to detect pulsing. Three-phase experiments used immiscible organic and aqueous liquids, and air or nitrogen as the gas phase. Hydrodynamic studies of flow regimes and holdup were performed for different types of packing, and mass transfer measurements were performed for a woven packing. These studies substantiated the improvements in mass transfer anticipated for pulse flow in multifunctional reactors for the acid-catalyzed C4 paraffin/olefin alkylation process. CR&L developed packings for this alkylation process, utilizing their alkylation process pilot facilities in Pasadena, TX. These packings were evaluated in the pilot-scale multifunctional reactor experiments established by Sandia to develop a more fundamental understanding of their role in process intensification. Lummus utilized the alkylation technology developed by CR&L to design and optimize the full commercial process utilizing multifunctional reactors containing the packings developed by CR&L and evaluated by Sandia. This hydrodynamic information has been developed for multifunctional chemical reactors utilizing pulse flow, for the acid-catalyzed C4 paraffin/olefin alkylation process, and is now accessible for use in

  16. Multiphysics analysis of liquid metal annular linear induction pumps: A project overview

    DOE PAGES

    Maidana, Carlos Omar; Nieminen, Juha E.

    2016-03-14

    Liquid metal-cooled fission reactors are both moderated and cooled by a liquid metal solution. These reactors are typically very compact and they can be used in regular electric power production, for naval and space propulsion systems or in fission surface power systems for planetary exploration. The coupling between the electromagnetics and thermo-fluid mechanical phenomena observed in liquid metal thermo-magnetic systems for nuclear and space applications gives rise to complex engineering magnetohydrodynamics and numerical problems. It is known that electromagnetic pumps have a number of advantages over rotating mechanisms: absence of moving parts, low noise and vibration level, simplicity of flowmore » rate regulation, easy maintenance and so on. However, while developing annular linear induction pumps, we are faced with a significant problem of magnetohydrodynamic instability arising in the device. The complex flow behavior in this type of devices includes a time-varying Lorentz force and pressure pulsation due to the time-varying electromagnetic fields and the induced convective currents that originates from the liquid metal flow, leading to instability problems along the device geometry. The determinations of the geometry and electrical configuration of liquid metal thermo-magnetic devices give rise to a complex inverse magnetohydrodynamic field problem were techniques for global optimization should be used, magnetohydrodynamics instabilities understood –or quantified- and multiphysics models developed and analyzed. Lastly, we present a project overview as well as a few computational models developed to study liquid metal annular linear induction pumps using first principles and the a few results of our multi-physics analysis.« less

  17. Multiphysics analysis of liquid metal annular linear induction pumps: A project overview

    SciTech Connect

    Maidana, Carlos Omar; Nieminen, Juha E.

    2016-03-14

    Liquid metal-cooled fission reactors are both moderated and cooled by a liquid metal solution. These reactors are typically very compact and they can be used in regular electric power production, for naval and space propulsion systems or in fission surface power systems for planetary exploration. The coupling between the electromagnetics and thermo-fluid mechanical phenomena observed in liquid metal thermo-magnetic systems for nuclear and space applications gives rise to complex engineering magnetohydrodynamics and numerical problems. It is known that electromagnetic pumps have a number of advantages over rotating mechanisms: absence of moving parts, low noise and vibration level, simplicity of flow rate regulation, easy maintenance and so on. However, while developing annular linear induction pumps, we are faced with a significant problem of magnetohydrodynamic instability arising in the device. The complex flow behavior in this type of devices includes a time-varying Lorentz force and pressure pulsation due to the time-varying electromagnetic fields and the induced convective currents that originates from the liquid metal flow, leading to instability problems along the device geometry. The determinations of the geometry and electrical configuration of liquid metal thermo-magnetic devices give rise to a complex inverse magnetohydrodynamic field problem were techniques for global optimization should be used, magnetohydrodynamics instabilities understood –or quantified- and multiphysics models developed and analyzed. Lastly, we present a project overview as well as a few computational models developed to study liquid metal annular linear induction pumps using first principles and the a few results of our multi-physics analysis.

  18. 10 CFR 71.97 - Advance notification of shipment of irradiated reactor fuel and nuclear waste.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... notification of shipment of irradiated reactor fuel and nuclear waste. (a) As specified in paragraphs (b), (c... advance notification of shipment of irradiated reactor fuel or nuclear waste must contain the following... irradiated reactor fuel or nuclear waste shipment; (2) A description of the irradiated reactor fuel...

  19. ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA660, INTERIOR. REACTOR INSIDE TANK. METAL ...

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

    ADVANCED REACTIVITY MEASUREMENT FACILITY, TRA-660, INTERIOR. REACTOR INSIDE TANK. METAL WORK PLATFORM ABOVE. THE REACTOR WAS IN A SMALL WATER-FILLED POOL. INL NEGATIVE NO. 66-6373. Unknown Photographer, ca. 1966 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  20. 75 FR 10840 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-03-09

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on ABWR will hold a meeting on March 18... 3, 2010. Antonio F. Dias, Chief, Reactor Safety Branch B, Advisory Committee on Reactor...

  1. 75 FR 7632 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-22

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Advanced Boiling Water Reactor (ABWR) The ACRS Subcommittee on ABWR will hold a meeting on March 2, 2010, at 11545...: February 12, 2010. Antonio F. Dias, Chief Reactor Safety Branch B, Advisory Committee on Reactor...

  2. Liquid-metal-piston MHD generator

    NASA Technical Reports Server (NTRS)

    Palmer, J. P.

    1969-01-01

    Magnetohydrodynamic generator uses a slug or piston of liquid potassium as the working fluid. An expanding vapor of the metal is allowed to reciprocate the liquid-metal-piston through a magnetic field and the expansion energy is converted directly into electrical energy.

  3. Solar-driven liquid metal magnetohydrodynamic generator

    NASA Technical Reports Server (NTRS)

    Lee, J. H.; Hohl, F.

    1981-01-01

    A solar oven heated by concentrated solar radiation as the heat source of a liquid metal magnetohydrodynamic (LMMHD) power generation system is proposed. The design allows the production of electric power in space, as well as on Earth, at high rates of efficiency. Two types of the solar oven suitable for the system are discussed.

  4. Conduction in fully ionized liquid metals

    NASA Technical Reports Server (NTRS)

    Stevenson, D. J.; Ashcroft, N. W.

    1973-01-01

    Electron transport is considered in high density fully ionized liquid metals. Ionic structure is described in terms of hard-sphere correlation functions and the scattering is determined from self-consistently screened point ions. Applications to the physical properties of the deep interior of Jupiter are briefly considered.

  5. Conduction in fully ionized liquid metals

    NASA Technical Reports Server (NTRS)

    Stevenson, D. J.; Ashcroft, N. W.

    1974-01-01

    Electron transport is considered in high-density fully ionized liquid metals. Ionic structure is described in terms of hard-sphere-correlation functions and the scattering is determined from self-consistently screened point ions. Applications to the physical properties of the deep interior of Jupiter are briefly considered.

  6. Preliminary safety evaluation of the advanced burner test reactor.

    SciTech Connect

    Dunn, F. E.; Fanning, T. H.; Cahalan, J. E.; Nuclear Engineering Division

    2006-09-15

    Results of a preliminary safety evaluation of the Advanced Burner Test Reactor (ABTR) pre-conceptual design are reported. The ABTR safety design approach is described. Traditional defense-in-depth design features are supplemented with passive safety performance characteristics that include natural circulation emergency decay heat removal and reactor power reduction by inherent reactivity feedbacks in accidents. ABTR safety performance in design-basis and beyond-design-basis accident sequences is estimated based on analyses. Modeling assumptions and input data for safety analyses are presented. Analysis results for simulation of simultaneous loss of coolant pumping power and normal heat rejection are presented and discussed, both for the case with reactor scram and the case without reactor scram. The analysis results indicate that the ABTR pre-conceptual design is capable of undergoing bounding design-basis and beyond-design-basis accidents without fuel cladding failures. The first line of defense for protection of the public against release of radioactivity in accidents remains intact with significant margin. A comparison and evaluation of general safety design criteria for the ABTR conceptual design phase are presented in an appendix. A second appendix presents SASSYS-1 computer code capabilities and modeling enhancements implemented for ABTR analyses.

  7. 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 [approximately] 20 yr in several countries. During the past decade, France and the United Kingdom have developed active programs 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 Experimental Breeder Reactor II (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. The Federal Republic of Germany (FRG) 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 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 focuses 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.

  8. Cermet-fueled reactors for advanced space applications

    SciTech Connect

    Cowan, C.L.; Palmer, R.S.; Taylor, I.N.; Vaidyanathan, S.; Bhattacharyya, S.K.; Barner, J.O.

    1987-12-01

    Cermet-fueled nuclear reactors are attractive candidates for high-performance advanced space power systems. The cermet consists of a hexagonal matrix of a refractory metal and a ceramic fuel, with multiple tubular flow channels. The high performance characteristics of the fuel matrix come from its high strength at elevated temperatures and its high thermal conductivity. The cermet fuel concept evolved in the 1960s with the objective of developing a reactor design that could be used for a wide range of mobile power generating sytems, including both Brayton and Rankine power conversion cycles. High temperature thermal cycling tests for the cermet fuel were carried out by General Electric as part of the 710 Project (General Electric 1966), and by Argonne National Laboratory in the Direct Nuclear Rocket Program (1965). Development programs for cermet fuel are currently under way at Argonne National Laboratory and Pacific Northwest Laboratory. The high temperature qualification tests from the 1960s have provided a base for the incorporation of cermet fuel in advanced space applications. The status of the cermet fuel development activities and descriptions of the key features of the cermet-fueled reactor design are summarized in this paper.

  9. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    SciTech Connect

    Joy Rempe; Darrell Knudson; Joshua Daw; Troy Unruh; Benjamin Chase; Kurt Davis; Robert Schley; Steven Taylor

    2012-08-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

  10. Enhanced In-Pile Instrumentation at the Advanced Test Reactor

    SciTech Connect

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; K. Condie

    2011-06-01

    Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility (NSUF) in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

  11. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    SciTech Connect

    Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

    2006-12-11

    This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

  12. Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems

    SciTech Connect

    D. E. Shropshire

    2009-01-01

    The Advanced Fuel Cycle Economic Analysis of Symbiotic Light-Water Reactor and Fast Burner Reactor Systems, prepared to support the U.S. Advanced Fuel Cycle Initiative (AFCI) systems analysis, provides a technology-oriented baseline system cost comparison between the open fuel cycle and closed fuel cycle systems. The intent is to understand their overall cost trends, cost sensitivities, and trade-offs. This analysis also improves the AFCI Program’s understanding of the cost drivers that will determine nuclear power’s cost competitiveness vis-a-vis other baseload generation systems. The common reactor-related costs consist of capital, operating, and decontamination and decommissioning costs. Fuel cycle costs include front-end (pre-irradiation) and back-end (post-iradiation) costs, as well as costs specifically associated with fuel recycling. This analysis reveals that there are large cost uncertainties associated with all the fuel cycle strategies, and that overall systems (reactor plus fuel cycle) using a closed fuel cycle are about 10% more expensive in terms of electricity generation cost than open cycle systems. The study concludes that further U.S. and joint international-based design studies are needed to reduce the cost uncertainties with respect to fast reactor, fuel separation and fabrication, and waste disposition. The results of this work can help provide insight to the cost-related factors and conditions needed to keep nuclear energy (including closed fuel cycles) economically competitive in the U.S. and worldwide. These results may be updated over time based on new cost information, revised assumptions, and feedback received from additional reviews.

  13. Advanced Space Nuclear Reactors from Fiction to Reality

    NASA Astrophysics Data System (ADS)

    Popa-Simil, L.

    The advanced nuclear power sources are used in a large variety of science fiction movies and novels, but their practical development is, still, in its early conceptual stages, some of the ideas being confirmed by collateral experiments. The novel reactor concept uses the direct conversion of nuclear energy into electricity, has electronic control of reactivity, being surrounded by a transmutation blanket and very thin shielding being small and light that at its very limit may be suitable to power an autonomously flying car. It also provides an improved fuel cycle producing minimal negative impact to environment. The key elements started to lose the fiction attributes, becoming viable actual concepts and goals for the developments to come, and on the possibility to achieve these objectives started to become more real because the theory shows that using the novel nano-technologies this novel reactor might be achievable in less than a century.

  14. Research reactor of the future: The advanced neutron source

    SciTech Connect

    Appleton, B.; West, C.

    1994-12-31

    Agents for cancer detection and treatment, stronger materials, better electronic gadgets, and other consumer and industrial products - these are assured benefits of a research reactor project proposed for Oak Ridge. Just as American companies have again assumed world leadership in producing semiconductor chips as well as cars and trucks, the United States is poised to retake the lead in neutron science by building and operating the $2.9 billion Advanced Neutron Source (ANS) research reactor by the start of the next century. In 1985, the neutron community, led by ORNL researchers, proposed a pioneering project, later called the ANS. Scheduled to begin operation in 2003, the ANS is seen not only as a replacement for the aging HFIR and HFBR but also as the best laboratory in the world for conducting neutron-based research.

  15. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    SciTech Connect

    Michael Swanson; Daniel Laudal

    2008-03-31

    The U.S. Department of Energy (DOE) National Energy Technology Laboratory Office of Coal and Environmental Systems has as its mission to develop advanced gasification-based technologies for affordable, efficient, zero-emission power generation. These advanced power systems, which are expected to produce near-zero pollutants, are an integral part of DOE's Vision 21 Program. DOE has also been developing advanced gasification systems that lower the capital and operating costs of producing syngas for chemical production. A transport reactor has shown potential to be a low-cost syngas producer compared to other gasification systems since its high-throughput-per-unit cross-sectional area reduces capital costs. This work directly supports the Power Systems Development Facility utilizing the KBR transport reactor located at the Southern Company Services Wilsonville, Alabama, site. Over 2800 hours of operation on 11 different coals ranging from bituminous to lignite along with a petroleum coke has been completed to date in the pilot-scale transport reactor development unit (TRDU) at the Energy & Environmental Research Center (EERC). The EERC has established an extensive database on the operation of these various fuels in both air-blown and oxygen-blown modes utilizing a pilot-scale transport reactor gasifier. This database has been useful in determining the effectiveness of design changes on an advanced transport reactor gasifier and for determining the performance of various feedstocks in a transport reactor. The effects of different fuel types on both gasifier performance and the operation of the hot-gas filter system have been determined. It has been demonstrated that corrected fuel gas heating values ranging from 90 to 130 Btu/scf have been achieved in air-blown mode, while heating values up to 230 Btu/scf on a dry basis have been achieved in oxygen-blown mode. Carbon conversions up to 95% have also been obtained and are highly dependent on the oxygen-coal ratio. Higher

  16. Nondestructive Measurements for Diagnostics of Advanced Reactor Passive Components

    SciTech Connect

    Prowant, Matthew S.; Dib, Gerges; Roy, Surajit; Luzi, Lorenzo; Ramuhalli, Pradeep

    2016-09-20

    Information on advanced reactor (AdvRx) component condition and failure probability is necessary to maintaining adequate safety margins and avoiding unplanned shutdowns, both of which have regulatory and economic consequences. Prognostic health management (PHM) technologies provide one approach to addressing these needs by providing the technical means for lifetime management of significant passive components and reactor internals. However, such systems require measurement data that are sensitive to degradation of the component. This paper describes results to date of ongoing research on nondestructive measurements of component condition for degradation mechanisms of relevance to AdvRx concepts. The focus of this paper is on in-situ ultrasonic measurements during high-temperature creep degradation. The data were analyzed to assess the sensitivity of the measurements to creep degradation, with the specific objective of assessing the suitability of the resulting correlations for remaining life prediction. The details of the measurements, results of data analysis, and ongoing research in this area are discussed.

  17. Evolutionary/advanced light water reactor data report

    SciTech Connect

    1996-02-09

    The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

  18. ADVANCED SEISMIC BASE ISOLATION METHODS FOR MODULAR REACTORS

    SciTech Connect

    E. Blanford; E. Keldrauk; M. Laufer; M. Mieler; J. Wei; B. Stojadinovic; P.F. Peterson

    2010-09-20

    Advanced technologies for structural design and construction have the potential for major impact not only on nuclear power plant construction time and cost, but also on the design process and on the safety, security and reliability of next generation of nuclear power plants. In future Generation IV (Gen IV) reactors, structural and seismic design should be much more closely integrated with the design of nuclear and industrial safety systems, physical security systems, and international safeguards systems. Overall reliability will be increased, through the use of replaceable and modular equipment, and through design to facilitate on-line monitoring, in-service inspection, maintenance, replacement, and decommissioning. Economics will also receive high design priority, through integrated engineering efforts to optimize building arrangements to minimize building heights and footprints. Finally, the licensing approach will be transformed by becoming increasingly performance based and technology neutral, using best-estimate simulation methods with uncertainty and margin quantification. In this context, two structural engineering technologies, seismic base isolation and modular steel-plate/concrete composite structural walls, are investigated. These technologies have major potential to (1) enable standardized reactor designs to be deployed across a wider range of sites, (2) reduce the impact of uncertainties related to site-specific seismic conditions, and (3) alleviate reactor equipment qualification requirements. For Gen IV reactors the potential for deliberate crashes of large aircraft must also be considered in design. This report concludes that base-isolated structures should be decoupled from the reactor external event exclusion system. As an example, a scoping analysis is performed for a rectangular, decoupled external event shell designed as a grillage. This report also reviews modular construction technology, particularly steel-plate/concrete construction using

  19. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    SciTech Connect

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  20. Advanced Test Reactor -- Testing Capabilities and Plans AND Advanced Test Reactor National Scientific User Facility -- Partnerships and Networks

    SciTech Connect

    Frances M. Marshall

    2008-07-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world’s premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner “lobes” to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. For future research, some ATR modifications and enhancements are currently planned. In 2007 the US Department of Energy designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR for material testing research by a broader user community. This paper provides more details on some of the ATR capabilities, key design features, experiments, and plans for the NSUF.

  1. Liquid Metal Machine Triggered Violin-Like Wire Oscillator.

    PubMed

    Yuan, Bin; Wang, Lei; Yang, Xiaohu; Ding, Yujie; Tan, Sicong; Yi, Liting; He, Zhizhu; Liu, Jing

    2016-10-01

    The first ever oscillation phenomenon of a copper wire embraced inside a self-powered liquid metal machine is discovered. When contacting a copper wire to liquid metal machine, it would be swallowed inside and then reciprocally moves back and forth, just like a violin bow. Such oscillation could be easily regulated by touching a steel needle on the liquid metal surface.

  2. INEL advanced test reactor plutonium-238 production feasibility assessment

    SciTech Connect

    Schnitzler, B.G. )

    1993-01-10

    Results of a preliminary neutronics assessment indicate the feasibility of [sup 238]Pu production in the Idaho National Engineering Laboratory Advanced Test Reactor (ATR). Based on the results of this assessment, an annual production of 11.3 kg [sup 238]Pu can be achieved in the ATR. An annual loading of 102 kg [sup 237]Np is required for the particular target configuration and irradiation scenario examined. The [sup 236]Pu contaminant level is approximately 6 parts per million at zero cooling time. The product quality is about 90% [sup 238]Pu. Neptunium feedstock requirements, [sup 238]Pu production rates, or product purity can be optimized depending on their relative importances.

  3. INEL advanced test reactor plutonium-238 production feasibility assessment

    NASA Astrophysics Data System (ADS)

    Schnitzler, Bruce G.

    1993-01-01

    Results of a preliminary neutronics assessment indicate the feasibility of 238Pu production in the Idaho National Engineering Laboratory Advanced Test Reactor (ATR). Based on the results of this assessment, an annual production of 11.3 kg 238Pu can be achieved in the ATR. An annual loading of 102 kg 237Np is required for the particular target configuration and irradiation scenario examined. The 236Pu contaminant level is approximately 6 parts per million at zero cooling time. The product quality is about 90% 238Pu. Neptunium feedstock requirements, 238Pu production rates, or product purity can be optimized depending on their relative importances.

  4. Advanced fuels for plutonium management in pressurized water reactors

    NASA Astrophysics Data System (ADS)

    Vasile, A.; Dufour, Ph; Golfier, H.; Grouiller, J. P.; Guillet, J. L.; Poinot, Ch; Youinou, G.; Zaetta, A.

    2003-06-01

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1. More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate.

  5. Completion of the first NGNP Advanced Gas Reactor Fuel Irradiation Experiment, AGR-1, in the Advanced Test Reactor

    SciTech Connect

    Blaine Grover; John Maki; David Petti

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and completed a very successful irradiation in early November 2009. The design of AGR-1 test train and support systems used to monitor and control the experiment during

  6. Mixing in a liquid metal electrode

    SciTech Connect

    Kelley, DH; Sadoway, DR

    2014-05-01

    Fluid mixing has first-order importance for many engineering problems in mass transport, including design and optimization of liquid-phase energy storage devices. Liquid metal batteries are currently being commercialized as a promising and economically viable technology for large-scale energy storage on worldwide electrical grids. But because these batteries are entirely liquid, fluid flow and instabilities may affect battery robustness and performance. Here we present estimates of flow magnitude and ultrasound measurements of the flow in a realistic liquid metal electrode. We find that flow does substantially affect mass transport by altering the electrode mixing time. Above a critical electrical current density, the convective flow organizes and gains speed, which promotes transport and would yield improved battery efficiency. (C) 2014 AIP Publishing LLC.

  7. Na-Zn liquid metal battery

    NASA Astrophysics Data System (ADS)

    Xu, Junli; Kjos, Ole Sigmund; Osen, Karen Sende; Martinez, Ana Maria; Kongstein, Ole Edvard; Haarberg, Geir Martin

    2016-11-01

    A new kind of membrane free liquid metal battery was developed. The battery employs liquid sodium and zinc as electrodes both in liquid state, and NaCl-CaCl2 molten salts as electrolyte. The discharge flat voltage is in the range of about 1.4 V-1.8 V, and the cycle efficiency achieved is about 90% at low discharge current densities (below 40 mA cm-2). Moreover, this battery can also be charged and discharged at high current density with good performance. The discharge flat voltage is above 1.1 V when it is discharged at 100 mA cm-2, while it is about 0.8 V with 100% cycle efficiency when it is discharged at 200 mA cm-2. Compared to other reported liquid metal battery, this battery has lower cost, which suggests broad application prospect in energy storage systems for power grid.

  8. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    NASA Astrophysics Data System (ADS)

    Wali, W. A.; Hassan, K. H.; Cullen, J. D.; Al-Shamma'a, A. I.; Shaw, A.; Wylie, S. R.

    2011-08-01

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  9. Flow excursion time scales in the advanced neutron source reactor

    SciTech Connect

    Sulfredge, C.D.

    1995-04-01

    Flow excursion transients give rise to a key thermal limit for the proposed Advanced Neutron Source (ANS) reactor because its core involves many parallel flow channels with a common pressure drop. Since one can envision certain accident scenarios in which the thermal limits set by flow excursion correlations might be exceeded for brief intervals, a key objective is to determine how long a flow excursion would take to bring about a system failure that could lead to fuel damage. The anticipated time scale for flow excursions has been examined by subdividing the process into its component phenomena: bubble nucleation and growth, deceleration of the resulting two-phase flow, and finally overcoming thermal inertia to heat up the reactor fuel plates. Models were developed to estimate the time required for each individual stage. Accident scenarios involving sudden reduction in core flow or core exit pressure have been examined, and the models compared with RELAP5 output for the ANS geometry. For a high-performance reactor like the ANS, flow excursion time scales were predicted to be in the millisecond range, so that even very brief transients might lead to fuel damage. These results should prove useful whenever one must determine the time involved in any portion of a flow excursion transient.

  10. 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.

  11. Temperature controlled material irradiation in the advanced test reactor

    NASA Astrophysics Data System (ADS)

    Ingram, F. W.; Palmer, A. J.; Stites, D. J.

    1998-10-01

    The United States Department of Energy (US DOE) has initiated the development of an Irradiation Test Vehicle (ITV) for fusion materials irradiation at the Advanced Test Reactor (ATR) in Idaho Falls, Idaho, USA. The ITV is capable of providing neutron spectral tailoring and individual temperature control for up to 15 experiment capsules simultaneously. The test vehicle consists of three In-Pile Tubes (IPTs) running the length of the reactor vessel. These IPTs are kept dry and test trains with integral instrumentation are inserted and removed through a transfer shield plate above the reactor vessel head. The test vehicle is designed to irradiate specimens as large as 2.2 cm in diameter, at temperatures of 250-800°C, achieving neutron damage rates as high as 10 displacements per atom per year. The high fast to thermal neutron flux ratio required for fusion materials testing is accomplished by using an aluminum filler to displace as much water as possible from the flux trap and surrounding the filler piece with a ring of replaceable neutron absorbing material. The gas blend temperature control system remains in place from test to test, thus hardware costs for new tests are limited to the experiment capsule train and integral instrumentation.

  12. Liquid metal switches for electromagnetic railgun systems

    SciTech Connect

    Mitcham, A.J.; Prothero, D.H.; Brooks, J.C. )

    1991-01-01

    The need for a reliable and effective commutating switch is essential to the operation of an HPG-driven railgun system. This switch must offer the lowest possible resistance during the current build up time and then must commutate the current quickly and efficiently into the railgun barrel. This paper considers the essential requirements for such a switch and, after briefly reviewing the available switch technologies, describes a new type of switch based on a liquid metal switching medium.

  13. The Liquid Metal Plasma Valve Closing Switch

    DTIC Science & Technology

    1976-11-01

    Specifications Closing HVDC LMPV Switch Converter LMPV Circuit Parameter Parameters Rating Breaker Rating Peak Voltage 50-200 kV 150 kV nominal 30...last column of the table outliner the ratings of the LMPV used in an existing circuit breaker system. The capability of the LMPV to operate at high...1976. W. 0. Eckhardt and G. A. Hofmann, ’’Ohmic Heading DC Circuit Breakers with Liquid-Metal Plasma Valves, 11 Proceedings of the Ninth Symposium on

  14. Self-fueled biomimetic liquid metal mollusk.

    PubMed

    Zhang, Jie; Yao, Youyou; Sheng, Lei; Liu, Jing

    2015-04-24

    A liquid metal motor that can "eat" aluminum food and then move spontaneously and swiftly in various solution configurations and structured channels for more than 1 h is discovered. Such a biomimetic mollusk is highly shape self-adaptive by closely conforming to the geometrical space it voyages in. The first ever self-fueled pump is illustrated as one of its typical practical utilizations.

  15. Stretchable Loudspeaker using Liquid Metal Microchannel

    PubMed Central

    Jin, Sang Woo; Park, Jeongwon; Hong, Soo Yeong; Park, Heun; Jeong, Yu Ra; Park, Junhong; Lee, Sang-Soo; Ha, Jeong Sook

    2015-01-01

    Considering the various applications of wearable and bio-implantable devices, it is desirable to realize stretchable acoustic devices for body-attached applications such as sensing biological signals, hearing aids, and notification of information via sound. In this study, we demonstrate the facile fabrication of a Stretchable Acoustic Device (SAD) using liquid metal coil of Galinstan where the SAD is operated by the electromagnetic interaction between the liquid metal coil and a Neodymium (Nd) magnet. To fabricate a liquid metal coil, Galinstan was injected into a micro-patterned elastomer channel. This fabricated SAD was operated simultaneously as a loudspeaker and a microphone. Measurements of the frequency response confirmed that the SAD was mechanically stable under both 50% uniaxial and 30% biaxial strains. Furthermore, 2000 repetitive applications of a 50% uniaxial strain did not induce any noticeable degradation of the sound pressure. Both voice and the beeping sound of an alarm clock were successfully recorded and played back through our SAD while it was attached to the wrist under repeated deformation. These results demonstrate the high potential of the fabricated SAD using Galinstan voice coil in various research fields including stretchable, wearable, and bio-implantable acoustic devices. PMID:26181209

  16. Stretchable Loudspeaker using Liquid Metal Microchannel

    NASA Astrophysics Data System (ADS)

    Jin, Sang Woo; Park, Jeongwon; Hong, Soo Yeong; Park, Heun; Jeong, Yu Ra; Park, Junhong; Lee, Sang-Soo; Ha, Jeong Sook

    2015-07-01

    Considering the various applications of wearable and bio-implantable devices, it is desirable to realize stretchable acoustic devices for body-attached applications such as sensing biological signals, hearing aids, and notification of information via sound. In this study, we demonstrate the facile fabrication of a Stretchable Acoustic Device (SAD) using liquid metal coil of Galinstan where the SAD is operated by the electromagnetic interaction between the liquid metal coil and a Neodymium (Nd) magnet. To fabricate a liquid metal coil, Galinstan was injected into a micro-patterned elastomer channel. This fabricated SAD was operated simultaneously as a loudspeaker and a microphone. Measurements of the frequency response confirmed that the SAD was mechanically stable under both 50% uniaxial and 30% biaxial strains. Furthermore, 2000 repetitive applications of a 50% uniaxial strain did not induce any noticeable degradation of the sound pressure. Both voice and the beeping sound of an alarm clock were successfully recorded and played back through our SAD while it was attached to the wrist under repeated deformation. These results demonstrate the high potential of the fabricated SAD using Galinstan voice coil in various research fields including stretchable, wearable, and bio-implantable acoustic devices.

  17. Stretchable Loudspeaker using Liquid Metal Microchannel.

    PubMed

    Jin, Sang Woo; Park, Jeongwon; Hong, Soo Yeong; Park, Heun; Jeong, Yu Ra; Park, Junhong; Lee, Sang-Soo; Ha, Jeong Sook

    2015-07-16

    Considering the various applications of wearable and bio-implantable devices, it is desirable to realize stretchable acoustic devices for body-attached applications such as sensing biological signals, hearing aids, and notification of information via sound. In this study, we demonstrate the facile fabrication of a Stretchable Acoustic Device (SAD) using liquid metal coil of Galinstan where the SAD is operated by the electromagnetic interaction between the liquid metal coil and a Neodymium (Nd) magnet. To fabricate a liquid metal coil, Galinstan was injected into a micro-patterned elastomer channel. This fabricated SAD was operated simultaneously as a loudspeaker and a microphone. Measurements of the frequency response confirmed that the SAD was mechanically stable under both 50% uniaxial and 30% biaxial strains. Furthermore, 2000 repetitive applications of a 50% uniaxial strain did not induce any noticeable degradation of the sound pressure. Both voice and the beeping sound of an alarm clock were successfully recorded and played back through our SAD while it was attached to the wrist under repeated deformation. These results demonstrate the high potential of the fabricated SAD using Galinstan voice coil in various research fields including stretchable, wearable, and bio-implantable acoustic devices.

  18. FFTF and Advanced Reactors Transition Program Resource Loaded Schedule

    SciTech Connect

    GANTT, D.A.

    2000-10-31

    This Resource Load Schedule (RLS) addresses two missions. The Advanced Reactors Transition (ART) mission, funded by DOE-EM, is to transition assigned, surplus facilities to a safe and compliant, low-cost, stable, deactivated condition (requiring minimal surveillance and maintenance) pending eventual reuse or D&D. Facilities to be transitioned include the 309 Building Plutonium Recycle Test Reactor (PRTR) and Nuclear Energy Legacy facilities. This mission is funded through the Environmental Management (EM) Project Baseline Summary (PBS) RL-TP11, ''Advanced Reactors Transition.'' The second mission, the Fast Flux Test Facility (FFTF) Project, is funded through budget requests submitted to the Office of Nuclear Energy, Science and Technology (DOE-NE). The FFTF Project mission is maintaining the FFTF, the Fuels and Materials Examination Facility (FMEF), and affiliated 400 Area buildings in a safe and compliant standby condition. This mission is to preserve the condition of the plant hardware, software, and personnel in a manner not to preclude a plant restart. This revision of the Resource Loaded Schedule (RLS) is based upon the technical scope in the latest revision of the following project and management plans: Fast Flux Test Facility Standby Plan (Reference 1); Hanford Site Sodium Management Plan (Reference 2); and 309 Building Transition Plan (Reference 4). The technical scope, cost, and schedule baseline is also in agreement with the concurrent revision to the ART Fiscal Year (FY) 2001 Multi-Year Work Plan (MYWP), which is available in an electronic version (only) on the Hanford Local Area Network, within the ''Hanford Data Integrator (HANDI)'' application.

  19. Fabrication and Pre-irradiation Characterization of a Minor Actinide and Rare Earth Containing Fast Reactor Fuel Experiment for Irradiation in the Advanced Test Reactor

    SciTech Connect

    Timothy A. Hyde

    2012-06-01

    The United States Department of Energy, seeks to develop and demonstrate the technologies needed to transmute the long-lived transuranic actinide isotopes contained in spent nuclear fuel into shorter lived fission products, thereby decreasing the volume of material requiring disposal and reducing the long-term radiotoxicity and heat load of high-level waste sent to a geologic repository. This transmutation of the long lived actinides plutonium, neptunium, americium and curium can be accomplished by first separating them from spent Light Water Reactor fuel using a pyro-metalurgical process, then reprocessing them into new fuel with fresh uranium additions, and then transmuted to short lived nuclides in a liquid metal cooled fast reactor. An important component of the technology is developing actinide-bearing fuel forms containing plutonium, neptunium, americium and curium isotopes that meet the stringent requirements of reactor fuels and materials.

  20. Natural Convection Heat Transfer in a Rectangular Liquid Metal Pool With Bottom Heating and Top Cooling

    SciTech Connect

    Lee, Il S.; Yu, Yong H.; Son, Hyoung M.; Hwang, Jin S.; Suh, Kune Y.

    2006-07-01

    An experimental study is performed to investigate the natural convection heat transfer characteristics with subcooled coolant to create engineering database for basic applications in a lead alloy cooled reactor. Tests are performed in the ALTOS (Applied Liquid-metal Thermal Operation Study) apparatus as part of MITHOS (Metal Integrated Thermo Hydrodynamic Operation System). A relationship is determined between the Nusselt number Nu and the Rayleigh number Ra in the liquid metal rectangular pool. Results are compared with correlations and experimental data in the literature. Given the similar Ra condition, the present test results for Nu of the liquid metal pool with top subcooling are found to be similar to those predicted by the existing correlations or experiments. The current test results are utilized to develop natural convection heat transfer correlations applicable to low Prandtl number Pr fluids that are heated from below and cooled by the external coolant above. Results from this study are slated to be used in designing BORIS (Battery Optimized Reactor Integral System), a small lead cooled modular fast reactor for deployment at remote sites cycled with MOBIS (Modular Optimized Brayton Integral System) for electricity generation, tied with NAVIS (Naval Application Vessel Integral System) for ship propulsion, joined with THAIS (Thermochemical Hydrogen Acquisition Integral System) for hydrogen production, and coupled with DORIS (Desalination Optimized Reactor Integral System) for seawater desalination. Tests are performed with Wood's metal (Pb-Bi-Sn-Cd) filling a rectangular pool whose lower surface is heated and upper surface cooled by forced convection of water. The test section is 20 cm long, 11.3 cm high and 15 cm wide. The simulant has a melting temperature of 78 deg. C. The constant temperature and heat flux condition was realized for the bottom heating once the steady state had been met. The test parameters include the heated bottom surface temperature

  1. Advanced Reactor Safety Research Division. Quarterly progress report, April 1-June 30, 1980

    SciTech Connect

    Romano, A.J.

    1980-01-01

    The Advanced Reactor Safety Research Programs Quarterly Progress Report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR safety evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  2. Advanced Reactor Safety Research Division. Quarterly progress report, January 1-March 31, 1980

    SciTech Connect

    Agrawal, A.K.; Cerbone, R.J.; Sastre, C.

    1980-06-01

    The Advanced Reactor Safety Research Programs quarterly progress report describes current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the USNRC Division of Reactor Safety Research. The projects reported each quarter are the following: HTGR Safety Evaluation, SSC Code Development, LMFBR Safety Experiments, and Fast Reactor Safety Code Validation.

  3. The Integral Fast Reactor: A practical approach to waste management

    SciTech Connect

    Laidler, J.J.

    1993-12-31

    This report discusses development of the method for pyroprocessing of spent fuel from the Integral Fast Reactor (or Advanced Liquid Metal Reactor). The technology demonstration phase, in which recycle will be demonstrated with irradiated fuel from the EBR-II reactor has been reached. Methods for recovering actinides from spent LWR fuel are at an earlier stage of development but appear to be technically feasible at this time, and a large-scale demonstration of this process has begun. The utilization of fully compatible processes for recycling valuable spent fuel materials promises to provide substantial economic incentives for future applications of the pyroprocessing technology.

  4. Nuclear reactor overflow line

    DOEpatents

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  5. Flow blockage analysis for the advanced neutron source reactor

    SciTech Connect

    Stovall, T.K.; Crabtree, J.A.; Felde, D.K.; Park, J.E.

    1996-01-01

    The Advanced Neutron Source (ANS) reactor was designed to provide a research tool with capabilities beyond those of any existing reactors. One portion of its state-of-the-art design required high-speed fluid flow through narrow channels between the fuel plates in the core. Experience with previous reactors has shown that fuel plate damage can occur when debris becomes lodged at the entrance to these channels. Such debris disrupts the fluid flow to the plate surfaces and can prevent adequate cooling of the fuel. Preliminary ANS designs addressed this issue by providing an unheated entrance length for each fuel plate so that any flow disruption would recover, thus providing adequate heat removal from the downstream, heated portions of the fuel plates. As part of the safety analysis, the adequacy of this unheated entrance length was assessed using both analytical models and experimental measurements. The Flow Blockage Test Facility (FBTF) was designed and built to conduct experiments in an environment closely matching the ANS channel geometry. The FBTF permitted careful measurements of both heat transfer and hydraulic parameters. In addition to these experimental efforts, a thin, rectangular channel was modeled using the Fluent computational fluid dynamics computer code. The numerical results were compared with the experimental data to benchmark the hydrodynamics of the model. After this comparison, the model was extended to include those elements of the safety analysis that were difficult to measure experimentally. These elements included the high wall heat flux pattern and variable fluid properties. The results were used to determine the relationship between potential blockage sizes and the unheated entrance length required.

  6. Advanced Test Reactor National Scientific User Facility Progress

    SciTech Connect

    Frances M. Marshall; Todd R. Allen; James I. Cole; Jeff B. Benson; Mary Catherine Thelen

    2012-10-01

    The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the world’s premier test reactors for studying the effects of intense neutron radiation on reactor materials and fuels. The ATR began operation in 1967, and has operated continuously since then, averaging approximately 250 operating days per year. The combination of high flux, large test volumes, and multiple experiment configuration options provide unique testing opportunities for nuclear fuels and material researchers. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected highly-enriched uranium fueled, reactor with a maximum operating power of 250 MWth. The ATR peak thermal flux can reach 1.0 x1015 n/cm2-sec, and the core configuration creates five main reactor power lobes (regions) that can be operated at different powers during the same operating cycle. In addition to these nine flux traps there are 68 irradiation positions in the reactor core reflector tank. The test positions range from 0.5” to 5.0” in diameter and are all 48” in length, the active length of the fuel. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material radiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. Goals of the ATR NSUF are to define the cutting edge of nuclear technology research in high temperature and radiation environments, contribute to improved industry performance of current and future light water reactors, and stimulate cooperative research between user groups conducting basic and applied research. The ATR NSUF has developed partnerships with other universities and national laboratories to enable ATR NSUF researchers to perform research at these other facilities, when the research objectives

  7. Supervisory Control System Architecture for Advanced Small Modular Reactors

    SciTech Connect

    Cetiner, Sacit M; Cole, Daniel L; Fugate, David L; Kisner, Roger A; Melin, Alexander M; Muhlheim, Michael David; Rao, Nageswara S; Wood, Richard Thomas

    2013-08-01

    This technical report was generated as a product of the Supervisory Control for Multi-Modular SMR Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (SMR) Research and Development Program of the U.S. Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular advanced SMR (AdvSMR) plants. This research activity advances the state-of-the art by incorporating decision making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides a brief history of hierarchical functional architectures and the current state-of-the-art, describes a reference AdvSMR to show the dependencies between systems, presents a hierarchical structure for supervisory control, indicates the importance of understanding trip setpoints, applies a new theoretic approach for comparing architectures, identifies cyber security controls that should be addressed early in system design, and describes ongoing work to develop system requirements and hardware/software configurations.

  8. Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Demonstration

    SciTech Connect

    Curtis Smith; Steven Prescott; Tony Koonce

    2014-04-01

    A key area of the Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) strategy is the development of methodologies and tools that will be used to predict the safety, security, safeguards, performance, and deployment viability of SMRs. The goal of the SMR PRA activity will be to develop quantitative methods and tools and the associated analysis framework for assessing a variety of risks. Development and implementation of SMR-focused safety assessment methods may require new analytic methods or adaptation of traditional methods to the advanced design and operational features of SMRs. We will need to move beyond the current limitations such as static, logic-based models in order to provide more integrated, scenario-based models based upon predictive modeling which are tied to causal factors. The development of SMR-specific safety models for margin determination will provide a safety case that describes potential accidents, design options (including postulated controls), and supports licensing activities by providing a technical basis for the safety envelope. This report documents the progress that was made to implement the PRA framework, specifically by way of demonstration of an advanced 3D approach to representing, quantifying and understanding flooding risks to a nuclear power plant.

  9. Overview of the US program of controls for advanced reactors

    SciTech Connect

    White, J.D.; Sackett, J.I.; Monson, R.; Lindsay, R.W.; Carroll, D.G.

    1989-01-01

    An automated control system can incorporate control goals and strategies, assessment of present and future plant status, diagnostic evaluation and maintenance planning, and signal and command validation. It has not been feasible to employ these capabilities in conventional hard-wired, analog, control systems. Recent advances in computer-based digital data acquisition systems, process controllers, fiber-optic signal transmission artificial intelligence tools and methods, and small inexpensive, fast, large-capacity computers---with both numeric and symbolic capabilities---have provided many of the necessary ingredients for developing large, practical automated control systems. Furthermore, recent reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. This paper presents an overall US national perspective for advanced controls research and development. The goals of high reliability, low operating cost and simple operation are described. The staged approach from conceptualization through implementation is discussed. Then the paper describes the work being done by ORNL, ANL and GE. The relationship of this work to the US commercial industry is also discussed.

  10. 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.

  11. Advanced neutron source reactor probabilistic flow blockage assessment

    SciTech Connect

    Ramsey, C.T.

    1995-08-01

    The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool.

  12. Fuel qualification plan for the Advanced Neutron Source Reactor

    SciTech Connect

    Copeland, G.L.

    1995-07-01

    This report describes the development and qualification plan for the fuel for the Advanced Neutron Source. The reference fuel is U{sub 3}Si{sub 2}, dispersed in aluminum and clad in 6061 aluminum. This report was prepared in May 1994, at which time the reference design was for a two-element core containing highly enriched uranium (93% {sup 235}U) . The reactor was in the process of being redesigned to accommodate lowered uranium enrichment and became a three-element core containing a higher volume fraction of uranium enriched to 50% {sup 235}U. Consequently, this report was not issued at that time and would have been revised to reflect the possibly different requirements of the lower-enrichment, higher-volume fraction fuel. Because the reactor is now being canceled, this unrevised report is being issued for archival purposes. The report describes the fabrication and inspection development plan, the irradiation tests and performance modeling to qualify performance, the transient testing that is part of the safety program, and the interactions and interfaces of the fuel development with other tasks.

  13. The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2009-09-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In

  14. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    SciTech Connect

    Not Available

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  15. The neutron texture diffractometer at the China Advanced Research Reactor

    NASA Astrophysics Data System (ADS)

    Li, Mei-Juan; Liu, Xiao-Long; Liu, Yun-Tao; Tian, Geng-Fang; Gao, Jian-Bo; Yu, Zhou-Xiang; Li, Yu-Qing; Wu, Li-Qi; Yang, Lin-Feng; Sun, Kai; Wang, Hong-Li; Santisteban, J. r.; Chen, Dong-Feng

    2016-03-01

    The first neutron texture diffractometer in China has been built at the China Advanced Research Reactor, due to strong demand for texture measurement with neutrons from the domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study texture in commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of zirconium alloy plate are presented. The comparison of texture measurements with the results obtained in HIPPO at LANSCE and Kowari at ANSTO illustrates the reliability of the texture diffractometer. Supported by National Nature Science Foundation of China (11105231, 11205248, 51327902) and International Atomic Energy Agency-TC program (CPR0012)

  16. Advanced nuclear reactor public opinion project. Interim report

    SciTech Connect

    Benson, B.

    1991-07-25

    This Interim Report summarizes the findings of our first twenty in-depth interviews in the Advanced Nuclear Reactor Public Opinion Project. We interviewed 6 industry trade association officials, 3 industry attorneys, 6 environmentalists/nuclear critics, 3 state officials, and 3 independent analysts. In addition, we have had numerous shorter discussions with various individuals concerned about nuclear power. The report is organized into the four categories proposed at our April, 1991, Advisory Group meeting: safety, cost-benefit analysis, science education, and communications. Within each category, some change of focus from that of the Advisory Group has been required, to reflect the findings of our interviews. This report limits itself to describing our findings. An accompanying memo draws some tentative conclusions.

  17. Thermal convection in a liquid metal battery

    NASA Astrophysics Data System (ADS)

    Shen, Yuxin; Zikanov, Oleg

    2016-08-01

    Generation of thermal convection flow in the liquid metal battery, a device recently proposed as a promising solution for the problem of the short-term energy storage, is analyzed using a numerical model. It is found that convection caused by Joule heating of electrolyte during charging or discharging is virtually unavoidable. It exists in laboratory prototypes larger than a few centimeters in size and should become much stronger in larger-scale batteries. The phenomenon needs further investigation in view of its positive (enhanced mixing of reactants) and negative (loss of efficiency and possible disruption of operation due to the flow-induced deformation of the electrolyte layer) effects.

  18. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-20

    ... COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.'' DATES... developed using this Catalog along with the Operator Licensing Examination Standards for Power...

  19. Testing of Liquid Metal Components for Nuclear Surface Power Systems

    NASA Technical Reports Server (NTRS)

    Polzin, K. A.; Pearson, J. B.; Godfroy, T. J.; Schoenfeld, M.; Webster, K.; Briggs, M. H.; Geng, S. M.; Adkins, H. E.; Werner, J. E.

    2010-01-01

    The capability to perform testing at both the module/component level and in near prototypic reactor configurations using a non-nuclear test methodology allowed for evaluation of two components critical to the development of a potential nuclear fission power system for the lunar surface. A pair of 1 kW Stirling power convertors, similar to the type that would be used in a reactor system to convert heat to electricity, were integrated into a reactor simulator system to determine their performance using pumped NaK as the hot side working fluid. The performance in the pumped-NaK system met or exceed the baseline performance measurements where the converters were electrically heated. At the maximum hot-side temperature of 550 C the maximum output power was 2375 watts. A specially-designed test apparatus was fabricated and used to quantify the performance of an annular linear induction pump that is similar to the type that could be used to circulate liquid metal through the core of a space reactor system. The errors on the measurements were generally much smaller than the magnitude of the measurements, permitting accurate performance evaluation over a wide range of operating conditions. The pump produced flow rates spanning roughly 0.16 to 5.7 l/s (2.5 to 90 GPM), and delta p levels from less than 1 kPa to 90 kPa (greater than 0.145 psi to roughly 13 psi). At the nominal FSP system operating temperature of 525 C the maximum efficiency was just over 4%.

  20. Liquid-metal flows: Magnetohydrodynamics and applications; Proceedings of the Fifth Beersheba International Seminar on Magnetohydrodynamic Flows and Turbulence, University of the Negev, Beersheba, Israel, Mar. 2-6, 1987

    NASA Astrophysics Data System (ADS)

    Branover, Herman; Mond, Michael; Unger, Yeshajahu

    The present collection of papers on MHD-related uses of liquid metal flows and their applications discusses topics in laminar MHD flows, MHD power generation, metallurgical MHD applications, and two-phase MHD flows. Attention is given to MHD flows with closed streamlines, nonlinear waves in liquid metals under a transverse magnetic field, liquid-metal MHD conversion of nuclear energy to electricity, the testing of optimized MHD conversion ('OMACON') systems, and aspects of a liquid-metal induction generator. Also discussed are MHD effects in liquid-metal breeder reactors, a plasma-driven MHD powerplant, modeling the recirculating flows in channel-induction surfaces, the hydrodynamics of aluminum reduction cells, free-surface determination in a levitation-melting process, the parametric interactions of waves in bubbly liquid metals, and the occurrence of cavitation in water jets.

  1. Natural Convection Heat Transfer Characteristics of Liquid Metal Cooled by Subcooled Water

    SciTech Connect

    Lee, Il S.; Yu, Yong H.; Son, Hyoung M.; Suh, Kune Y.

    2006-07-01

    An experimental study is performed to investigate the natural convection heat transfer characteristics with subcooled coolant to create engineering database for basic applications in a lead alloy cooled reactor. Tests are performed in the ALTOS (Applied Liquid-metal Thermal Operation Study) apparatus as part of MITHOS (Metal Integrated Thermo Hydrodynamic Operation System). The relationship between the Nusselt number (Nu) and the Rayleigh number (Ra) in the liquid metal is determined and compared with the correlations in the literature and experimental results. Given the similar Ra condition, the present test results for Nu of the liquid metal pool with subcooled coolant are found to be similar to those predicted by the existing correlations or measured from previous experiments. The current experimental results are utilized to develop new engineering solutions. The new experimental correlations for predicting the natural convection heat transfer are applicable to low Prandtl number (Pr) materials that are heated from below and cooled by the external coolant above. Results from this study are slated to be used to design BORIS (Battery Optimized Reactor Integral System), a small lead cooled modular fast reactor for deployment in remote sites. Tests are performed with air, water and Wood's metal (Pb-Bi-Sn-Cd) filling a rectangular pool while the lower surface is heated and the upper surface cooled by forced convection of water. The inner dimensions of the test section are 20 cm in length, 11.3 cm in height, and 15 cm in width. Wood's metal has a melting temperature of 78 deg. C. Constant temperature and heat flux condition is adopted for the bottom heating. The test parameters include the heated bottom surface temperature of the liquid metal pool, the input power to the bottom surface of the section, and the coolant temperature. (authors)

  2. Diverse transformations of liquid metals between different morphologies.

    PubMed

    Sheng, Lei; Zhang, Jie; Liu, Jing

    2014-09-10

    Transformation from a film into a sphere, rapid merging of separate objects, controlled self-rotation, and planar locomotion are the very unusual phenomena observed in liquid metals under application of an electric field to a liquid metal immersed in or sprayed with water. A mechanism for these effects is suggested and potential applications - for example the recovery of liquid metal previously injected into the body for therapeutic purposes - are outlined.

  3. Instability of the Liquid Metal-Pattern Interface in the Lost Foam Casting of Aluminum Alloys

    NASA Astrophysics Data System (ADS)

    Griffiths, W. D.; Ainsworth, M. J.

    2016-06-01

    The nature of the liquid metal-pattern interface during mold filling in the Lost Foam casting of aluminum alloys was investigated using real-time X-ray radiography for both normal expanded polystyrene, and brominated polystyrene foam patterns. Filling the pattern under the action of gravity from above or below had little effect on properties, both cases resulting in a large scatter of tensile strength values, (quantified by their Weibull Modulus). Countergravity filling at different velocities demonstrated that the least scatter of tensile strength values (highest Weibull Modulus) was associated with the slowest filling, when a planar liquid metal-pattern interface occurred. Real-time X-ray radiography showed that the advancing liquid metal front became unstable above a certain critical velocity, leading to the entrainment of the degrading pattern material and associated defects. It has been suggested that the transition of the advancing liquid metal-pattern interface into an unstable regime may be a result of Saffman-Taylor Instability.

  4. Toxicity of irradiated advanced heavy water reactor fuels.

    PubMed

    Priest, N D; Richardson, R B; Edwards, G W R

    2013-02-01

    The good neutron economy and online refueling capability of the CANDU® heavy water moderated reactor (HWR) enable it to use many different fuels such as low enriched uranium (LEU), plutonium, or thorium, in addition to its traditional natural uranium (NU) fuel. The toxicity and radiological protection methods for these proposed fuels, unlike those for NU, are not well established. This study uses software to compare the fuel composition and toxicity of irradiated NU fuel against those of two irradiated advanced HWR fuel bundles as a function of post-irradiation time. The first bundle investigated is a CANFLEX® low void reactor fuel (LVRF), of which only the dysprosium-poisoned central element, and not the outer 42 LEU elements, is specifically analyzed. The second bundle investigated is a heterogeneous high-burnup (LEU,Th)O(2) fuelled bundle, whose two components (LEU in the outer 35 elements and thorium in the central eight elements) are analyzed separately. The LVRF central element was estimated to have a much lower toxicity than that of NU at all times after shutdown. Both the high burnup LEU and the thorium fuel had similar toxicity to NU at shutdown, but due to the creation of such inhalation hazards as (238)Pu, (240)Pu, (242)Am, (242)Cm, and (244)Cm (in high burnup LEU), and (232)U and (228)Th (in irradiated thorium), the toxicity of these fuels was almost double that of irradiated NU after 2,700 d of cooling. New urine bioassay methods for higher actinoids and the analysis of thorium in fecal samples are recommended to assess the internal dose from these two fuels.

  5. Recovery of Information from the Fast Flux Test Facility for the Advanced Fuel Cycle Initiative

    SciTech Connect

    Nielsen, Deborah L.; Makenas, Bruce J.; Wootan, David W.; Butner, R. Scott; Omberg, Ronald P.

    2009-09-30

    The Fast Flux Test Facility is the most recent Liquid Metal Reactor to operate in the United States. Information from the design, construction, and operation of this reactor was at risk as the facilities associated with the reactor are being shut down. The Advanced Fuel Cycle Initiative is a program managed by the Office of Nuclear Energy of the U.S. Department of Energy with a mission to develop new fuel cycle technologies to support both current and advanced reactors. Securing and preserving the knowledge gained from operation and testing in the Fast Flux Test Facility is an important part of the Knowledge Preservation activity in this program.

  6. Advanced BWR stability monitoring tests with a hybrid reactor facility

    SciTech Connect

    He, Weidong; Huang, Zhengyu; Edwards, Robert M.

    2002-07-01

    A stability monitor for a boiling reactor is implemented and evaluated with the Penn State Hybrid Reactor System. The stability monitor is based on an extended Kalman filter which employs a reduced-order BWR reactor model. The filter uses measured power signal and estimates the void reactivity feedback gain and the decay ratio. The hybrid reactor system is a system combining a simulation module of BWR thermal hydraulics and the Penn State TRIGA reactor. A description of the hybrid system is also presented. (authors)

  7. 77 FR 62270 - Proposed Revision Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-12

    ... Non-Safety Systems for Passive Advanced Light Water Reactors AGENCY: Nuclear Regulatory Commission... Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The current SRP does not contain guidance on the proposed RTNSS for Passive Advance Light Water Reactors. DATES: Submit comments by November...

  8. 78 FR 41436 - Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-07-10

    ... COMMISSION Proposed Revision to Treatment of Non-Safety Systems for Passive Advanced Light Water Reactors... Treatment of Non-Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' The NRC seeks public...- Safety Systems (RTNSS) for Passive Advanced Light Water Reactors.'' This area includes a revised...

  9. Metal pad instabilities in liquid metal batteries

    NASA Astrophysics Data System (ADS)

    Zikanov, Oleg

    2016-11-01

    Strong variations between the electrical conductivities of electrolyte and metal layers in a liquid metal battery indicate the possibility of 'metal pad' instabilities. Deformations of the electrolyte-metal interfaces cause strong perturbations of electric currents, which, hypothetically, can generate Lorentz forces enhancing the deformations. We investigate this possibility using two models: a mechanical analogy and a two-dimensional linearized approximation. It is found that the battery is prone to instabilities of two types. One is similar to the sloshing-wave instability observed in the Hall-Héroult aluminum reduction cells. Another is new and related to the interactions of current perturbations with the azimuthal magnetic field induced by the base current. Financial support was provided by the U.S. National Science Foundation (Grant CBET 1435269).

  10. Use of liquid metals in nuclear and thermonuclear engineering, and in other innovative technologies

    NASA Astrophysics Data System (ADS)

    Rachkov, V. I.; Arnol'dov, M. N.; Efanov, A. D.; Kalyakin, S. G.; Kozlov, F. A.; Loginov, N. I.; Orlov, Yu. I.; Sorokin, A. P.

    2014-05-01

    By now, a good deal of experience has been gained with using liquid metals as coolants in nuclear power installations; extensive knowledge has been gained about the physical, thermophysical, and physicochemical properties of these coolants; and the scientific principles and a set of methods and means for handling liquid metals as coolants for nuclear power installations have been elaborated. Prototype and commercialgrade sodium-cooled NPP power units have been developed, including the BOR-60, BN-350, and BN-600 power units (the Soviet Union); the Rapsodie, Phenix, and Superphenix power units (France), the EBR-II power unit (the United States); and the PFR power unit (the United Kingdom). In Russia, dedicated nuclear power installations have been constructed, including those with a lead-bismuth coolant for nuclear submarines and with sodium-potassium alloy for spacecraft (the Buk and Topol installations), which have no analogs around the world. Liquid metals (primarily lithium and its alloy with lead) hold promise for use in thermonuclear power engineering, where they can serve not only as a coolant, but also as tritium-producing medium. In this article, the physicochemical properties of liquid metal coolants, as well as practical experience gained from using them in nuclear and thermonuclear power engineering and in innovative technologies are considered, and the lines of further research works are formulated. New results obtained from investigations carried out on the Pb-Bi and Pb for the SVBR and BREST fast-neutron reactors (referred to henceforth as fast reactors) and for controlled accelerator systems are described.

  11. Advanced Test Reactor probabilistic risk assessment methodology and results summary

    SciTech Connect

    Eide, S.A.; Atkinson, S.A.; Thatcher, T.A.

    1992-01-01

    The Advanced Test Reactor (ATR) probabilistic risk assessment (PRA) Level 1 report documents a comprehensive and state-of-the-art study to establish and reduce the risk associated with operation of the ATR, expressed as a mean frequency of fuel damage. The ATR Level 1 PRA effort is unique and outstanding because of its consistent and state-of-the-art treatment of all facets of the risk study, its comprehensive and cost-effective risk reduction effort while the risk baseline was being established, and its thorough and comprehensive documentation. The PRA includes many improvements to the state-of-the-art, including the following: establishment of a comprehensive generic data base for component failures, treatment of initiating event frequencies given significant plant improvements in recent years, performance of efficient identification and screening of fire and flood events using code-assisted vital area analysis, identification and treatment of significant seismic-fire-flood-wind interactions, and modeling of large loss-of-coolant accidents (LOCAs) and experiment loop ruptures leading to direct damage of the ATR core. 18 refs.

  12. Liquid metal embrittlement. [crack propagation in metals with liquid metal in crack space

    NASA Technical Reports Server (NTRS)

    Tiller, W. A.

    1973-01-01

    Crack propagation is discussed for metals with liquid metal in the crack space. The change in electrochemical potential of an electron in a metal due to changes in stress level along the crack surface was investigated along with the change in local chemistry, and interfacial energy due to atomic redistribution in the liquid. Coupled elastic-elastrostatic equations, stress effects on electron energy states, and crack propagation via surface roughening are discussed.

  13. Measurement of the differential pressure of liquid metals

    DOEpatents

    Metz, H.J.

    1975-09-01

    This patent relates to an improved means for measuring the differential pressure between any two points in a process liquid metal coolant loop, wherein the flow of liquid metal in a pipe is opposed by a permanent magnet liquid metal pump until there is almost zero flow shown by a magnetic type flowmeter. The pressure producing the liquid metal flow is inferred from the rate of rotation of the permanent magnet pump. In an alternate embodiment, a differential pressure transducer is coupled to a process pipeline by means of high-temperature bellows or diaphragm seals, and a permanent magnet liquid metal pump in the high-pressure transmission line to the pressure transducer can be utilized either for calibration of the transducer or for determining the process differential pressure as a function of the magnet pump speed. (auth)

  14. Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

    SciTech Connect

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma; Al Rashdan, Ahmad; Tsvetkov, Pavel Valeryevich; Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

    2011-05-01

    This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

  15. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    DTIC Science & Technology

    2006-12-01

    calculate the generation of Polonium - 210 in reactors cooled by lead and lead- bismuth eutectic. The motivation for this is to address a noted lack of...calculate the generation of Polonium - 210 in reactors cooled by lead and lead-bismuth eutectic. The motivation for this is to address a noted lack of...coolants. The objectives of thesis are two fold. The first objective is to independently calculate the generation of Polonium - 210 in reactors

  16. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    NASA Technical Reports Server (NTRS)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  17. Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

    SciTech Connect

    John Jackson; Todd Allen; Frances Marshall; Jim Cole

    2013-03-01

    The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials

  18. Numerical Modeling of Inclusion Behavior in Liquid Metal Processing

    NASA Astrophysics Data System (ADS)

    Bellot, Jean-Pierre; Descotes, Vincent; Jardy, Alain

    2013-09-01

    Thermomechanical performance of metallic alloys is directly related to the metal cleanliness that has always been a challenge for metallurgists. During liquid metal processing, particles can grow or decrease in size either by mass transfer with the liquid phase or by agglomeration/fragmentation mechanisms. As a function of numerical density of inclusions and of the hydrodynamics of the reactor, different numerical modeling approaches are proposed; in the case of an isolated particle, the Lagrangian technique coupled with a dissolution model is applied, whereas in the opposite case of large inclusion phase concentration, the population balance equation must be solved. Three examples of numerical modeling studies achieved at Institut Jean Lamour are discussed. They illustrate the application of the Lagrangian technique (for isolated exogenous inclusion in titanium bath) and the Eulerian technique without or with the aggregation process: for precipitation and growing of inclusions at the solidification front of a Maraging steel, and for endogenous inclusions in the molten steel bath of a gas-stirred ladle, respectively.

  19. Proceedings of a Symposium on Advanced Compact Reactor Systems

    NASA Technical Reports Server (NTRS)

    1983-01-01

    Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

  20. Design of GA thermochemical water-splitting process for the Mirror Advanced Reactor System

    SciTech Connect

    Brown, L.C.

    1983-04-01

    GA interfaced the sulfur-iodine thermochemical water-splitting cycle to the Mirror Advanced Reactor System (MARS). The results of this effort follow as one section and part of a second section to be included in the MARS final report. This section describes the process and its interface to the reactor. The capital and operating costs for the hydrogen plant are described.

  1. 75 FR 66803 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-29

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR); Notice of Meeting The ACRS Subcommittee on ABWR will hold a meeting...

  2. 3-D THERMAL EVALUATIONS FOR a FUELED EXPERIMENT in the ADVANCED TEST REACTOR

    SciTech Connect

    Ambrosek, R.G.; Chang, G.S.; Utterbeck, D.J.

    2004-10-06

    The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large ''B'' experimental facility. A test configuration has been identified for demonstrating fuel types typical of gas cooled reactors or fast reactors that may play a role in closing the fuel cycle or increasing efficiency via high temperature operation Plans are to have 6 capsules, each containing 12 compacts, for the test configuration. Each capsule will have its own temperature control system. Passing a helium-neon gas through the void regions between the fuel compacts and the graphite carrier and between the graphite carrier and the capsule wall will control temperature. This design with three compacts per axial level was evaluated for thermal performance to ascertain the temperature distributions in the capsule and test specimens with heating rates that encompass the range of initial heat generation rates.

  3. 3-D Thermal Evaluations for a Fueled Experiment in the Advanced Test Reactor

    SciTech Connect

    Richard Ambrosek; Gray Chang; Debra Utterbeck

    2004-10-01

    The DOE Advanced Fuel Cycle Initiative and Generation IV reactor programs are developing new fuel types for use in the current Light Water Reactors and future advanced reactor concepts. The Advanced Gas Reactor program is planning to test fuel to be used in the Next Generation Nuclear Plant (NGNP) nuclear reactor. Preliminary information for assessing performance of the fuel will be obtained from irradiations performed in the Advanced Test Reactor large “B” experimental facility. A test configurations has been identified for demonstrating fuel types typical of gas cooled reactors or fast reactors that may play a role in closing the fuel cycle or increasing efficiency via high temperature operation Plans are to have 6 capsules, each containing 12 compacts, for the test configuration. Each capsule will have its own temperature control system. Passing a helium-neon gas through the void regions between the fuel compacts and the graphite carrier and between the graphite carrier and the capsule wall will control temperature. This design with three compacts per axial level was evaluated for thermal performance to ascertain the temperature distributions in the capsule and test specimens with heating rates that encompass the range of initial heat generation rates.

  4. The Advanced Test Reactor National Scientific User Facility Advancing Nuclear Technology

    SciTech Connect

    T. R. Allen; J. B. Benson; J. A. Foster; F. M. Marshall; M. K. Meyer; M. C. Thelen

    2009-05-01

    To help ensure the long-term viability of nuclear energy through a robust and sustained research and development effort, the U.S. Department of Energy (DOE) designated the Advanced Test Reactor and associated post-irradiation examination facilities a National Scientific User Facility (ATR NSUF), allowing broader access to nuclear energy researchers. The mission of the ATR NSUF is to provide access to world-class nuclear research facilities, thereby facilitating the advancement of nuclear science and technology. The ATR NSUF seeks to create an engaged academic and industrial user community that routinely conducts reactor-based research. Cost free access to the ATR and PIE facilities is granted based on technical merit to U.S. university-led experiment teams conducting non-proprietary research. Proposals are selected via independent technical peer review and relevance to DOE mission. Extensive publication of research results is expected as a condition for access. During FY 2008, the first full year of ATR NSUF operation, five university-led experiments were awarded access to the ATR and associated post-irradiation examination facilities. The ATR NSUF has awarded four new experiments in early FY 2009, and anticipates awarding additional experiments in the fall of 2009 as the results of the second 2009 proposal call. As the ATR NSUF program mature over the next two years, the capability to perform irradiation research of increasing complexity will become available. These capabilities include instrumented irradiation experiments and post-irradiation examinations on materials previously irradiated in U.S. reactor material test programs. The ATR critical facility will also be made available to researchers. An important component of the ATR NSUF an education program focused on the reactor-based tools available for resolving nuclear science and technology issues. The ATR NSUF provides education programs including a summer short course, internships, faculty-student team

  5. Mirror Advanced Reactor Study (MARS): executive summary and overview

    SciTech Connect

    Logan, B.G.; Perkins, L.J.; Gordon, J.D.

    1984-07-01

    Two self-consistent MARS configurations are discussed - a 1200-MWe commercial electricity-generating plant and a synguels-generating plant that produces hydrogen with an energy equivalent to 26,000 barrels of oil per day. The MARS machine emphasizes the attractive features of the tandem mirror concept, including steady-state operation, a small-diameter high-beta plasma, a linear central cell with simple low-maintenance blankets, low first-wall heat fluxes (<10 W/cm/sup 2/), no driven plasma currents or associated disruptions, natural halo impurity diversion, and direct conversion of end-loss charged-particle power. The MARS electric plant produces 2600 MW of fusion power in a 130-m-long central cell. Advanced tandem-mirror plasma-engineering concepts, a high-efficiency liquid lithium-lead (Li/sub 17/Pb/sub 83/) blanket, and efficient direct electrical conversion of end loss power combine to produce a high net plant efficiency of 36%. With a total capital cost of $2.9 billion (constant 1983 dollars), the MARS electric plant produces busbar electricity at approx. 7 cents/kW-hour. The MARS synfuels plant produces 3500 MW of fusion power in a 150-m-long central cell. A helium-gas-cooled silicon carbide pebble-bed blanket provides high-temperature (1000/sup 0/C) heat to a thermochemical water-splitting cycle and the resulting hydrogen is catalytically converted to methanol for distribution. With a total capital cost of $3.6 billion (constant 1983 dollars), the synfuels plant produces methanol fuel at about $1.7/gal. The major features of the MARS reactor include sloshing-ion thermal barrier plugs for efficient plasma confinement, a high efficiency blanket, high-field (24-T) choke cells, drift pumping for trapped plasma species, quasi-optical electron-cyclotron resonant heating (ECRH) systems, and a component gridless direct converter.

  6. Advanced reactor development: The LMR integral fast reactor program at Argonne

    SciTech Connect

    Till, C.E.

    1990-01-01

    Reactor technology for the 21st Century must develop with characteristics that can now be seen to be important for the future, quite different from the things when the fundamental materials and design choices for present reactors were made in the 1950s. Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 3 figs.

  7. Worldwide advanced nuclear power reactors with passive and inherent safety: What, why, how, and who

    SciTech Connect

    Forsberg, C.W.; Reich, W.J.

    1991-09-01

    The political controversy over nuclear power, the accidents at Three Mile Island (TMI) and Chernobyl, international competition, concerns about the carbon dioxide greenhouse effect and technical breakthroughs have resulted in a segment of the nuclear industry examining power reactor concepts with PRIME safety characteristics. PRIME is an acronym for Passive safety, Resilience, Inherent safety, Malevolence resistance, and Extended time after initiation of an accident for external help. The basic ideal of PRIME is to develop power reactors in which operator error, internal sabotage, or external assault do not cause a significant release of radioactivity to the environment. Several PRIME reactor concepts are being considered. In each case, an existing, proven power reactor technology is combined with radical innovations in selected plant components and in the safety philosophy. The Process Inherent Ultimate Safety (PIUS) reactor is a modified pressurized-water reactor, the Modular High Temperature Gas-Cooled Reactor (MHTGR) is a modified gas-cooled reactor, and the Advanced CANDU Project is a modified heavy-water reactor. In addition to the reactor concepts, there is parallel work on super containments. The objective is the development of a passive box'' that can contain radioactivity in the event of any type of accident. This report briefly examines: why a segment of the nuclear power community is taking this new direction, how it differs from earlier directions, and what technical options are being considered. A more detailed description of which countries and reactor vendors have undertaken activities follows. 41 refs.

  8. Process for preparing liquid metal electrical contact device

    NASA Technical Reports Server (NTRS)

    Lovell, R. R.; Berkopec, F. D.; Culp, D. H. (Inventor)

    1977-01-01

    The parts of an electrical contact device are treated by sputter etching to remove the parent metal oxide. Prior to exposure of the electrodes to any oxygen, a sacrificial metal is sputter deposited on the parts. Preferably this sacrificial metal is one that oxidizes slowly and is readily dissolved by the liquid metal. The sacrificial metal may then be removed from unwanted areas. The remainder of the ring and the probe to be wet by the liquid metal are submerged in the liquid metal or the liquid metal is flushed over these areas, preferably while they are being slightly abraded, unitl all the sacrificial material on these portions is wet by the liquid metal. In doing so the liquid metal dissolves the sacrificial metal and permanently wets the parent metal. Preferred materials used in the process and for the electrodes of electrical contact devices are high purity (99.0%) nickel or AISI type 304 stainless steel for the electrical contact devices, gallium as the liquid metal, and gold as the sacrificial material.

  9. Recent applications of liquid metals featuring nanoscale surface oxides

    NASA Astrophysics Data System (ADS)

    Neumann, Taylor V.; Dickey, Michael D.

    2016-05-01

    This proceeding describes recent efforts from our group to control the shape and actuation of liquid metal. The liquid metal is an alloy of gallium and indium which is non-toxic, has negligible vapor pressure, and develops a thin, passivating surface oxide layer. The surface oxide allows the liquid metal to be patterned and shaped into structures that do not minimize interfacial energy. The surface oxide can be selectively removed by changes in pH or by applying a voltage. The surface oxide allows the liquid metal to be 3D printed to form free-standing structures. It also allows for the liquid metal to be injected into microfluidic channels and to maintain its shape within the channels. The selective removal of the oxide results in drastic changes in surface tension that can be used to control the flow behavior of the liquid metal. The metal can also wet thin, solid films of metal that accelerates droplets of the liquid along the metal traces .Here we discuss the properties and applications of liquid metal to make soft, reconfigurable electronics.

  10. Heavy liquid metals: Research programs at PSI

    SciTech Connect

    Takeda, Y.

    1996-06-01

    The author describes work at PSI on thermohydraulics, thermal shock, and material tests for mechnical properties. In the presentation, the focus is on two main programs. (1) SINQ LBE target: The phase II study program for SINQ is planned. A new LBE loop is being constructed. The study has the following three objectives: (a) Pump study - design work on an electromagnetic pump to be integrated into the target. (b) Heat pipe performance test - the use of heat pipes as an additional component of the target cooling system is being considered, and it may be a way to futher decouple the liquid metal and water coolant loops. (c) Mixed convection experiment - in order to find an optimal configuration of the additional flow guide for window cooling, mixed convection around the window is to be studied. The experiment will be started using water and then with LBE. (2) ESS Mercury target: For ESS target study, the following experimental studies are planned, some of which are exampled by trial experiments. (a) Flow around the window: Flow mapping around the hemi-cylindrical window will be made for optimising the flow channels and structures, (b) Geometry optimisation for minimizing a recirculation zone behind the edge of the flow separator, (c) Flow induced vibration and buckling problem for a optimised structure of the flow separator and (d) Gas-liquid two-phase flow will be studied by starting to establish the new experimental method of measuring various kinds of two-phase flow characteristics.

  11. Modern Aspects of Liquid Metal Engineering

    NASA Astrophysics Data System (ADS)

    Czerwinski, Frank

    2017-02-01

    Liquid metal engineering (LME) refers to a variety of physical and/or chemical treatments of molten metals aimed at influencing their solidification characteristics. Although the fundamentals have been known for decades, only recent progress in understanding solidification mechanisms has renewed an interest in opportunities this technique creates for an improvement of castings. This review covers conventional and novel concepts of LME with their application to modern manufacturing techniques based not only on liquid but also on semisolid routes. The role of external forces applied to the melt combined with grain nucleation control is explained along with laboratory- and commercial-scale equipment designed for implementation of various concepts exploring mechanical, electromagnetic, and ultrasound principles. An influence of melt treatments on quality of the final product is considered through distinguishing between internal integrity of net shape components and the alloy microstructure. Recent global developments indicate that exploring the synergy of melt chemistry and physical treatments achieved through LME allows creating the optimum conditions for nucleation and growth during solidification, positively affecting quality of castings.

  12. Tokamak with liquid metal for inducing toroidal electrical field

    DOEpatents

    Ohkawa, Tihiro

    1981-01-01

    A tokamak apparatus includes a vessel for defining a reservoir and confining liquid therein. A toroidal liner disposed within said vessel defines a toroidal space within the liner confines gas therein. Liquid metal fills the reservoir outside the liner. A magnetic field is established in the liquid metal to develop magnetic flux linking the toroidal space. The gas is ionized. The liquid metal and the toroidal space are moved relative to one another transversely of the space to generate electric current in the ionized gas in the toroidal space about its major axis and thereby heat plasma developed in the toroidal space.

  13. Contactless Inductive Bubble Detection in a Liquid Metal Flow

    PubMed Central

    Gundrum, Thomas; Büttner, Philipp; Dekdouk, Bachir; Peyton, Anthony; Wondrak, Thomas; Galindo, Vladimir; Eckert, Sven

    2016-01-01

    The detection of bubbles in liquid metals is important for many technical applications. The opaqueness and the high temperature of liquid metals set high demands on the measurement system. The high electrical conductivity of the liquid metal can be exploited for contactless methods based on electromagnetic induction. We will present a measurement system which consists of one excitation coil and a pickup coil system on the opposite sides of the pipe. With this sensor we were able to detect bubbles in a sodium flow inside a stainless steel pipe and bubbles in a column filled with a liquid Gallium alloy. PMID:26751444

  14. Ultrafine Pitch Stencil Printing of Liquid Metal Alloys.

    PubMed

    Lazarus, Nathan; Bedair, Sarah S; Kierzewski, Iain M

    2017-01-18

    With high conductivity and stretchable for large cross-sections, liquid metals such as galinstan are promising for creating stretchable devices and interconnects. Creating high resolution features in parallel is challenging, with most techniques limited to a hundred micrometers or more. In this work, multilevel electroplated stencils are investigated for printing liquid metals, with galinstan features as small as ten micrometers printed on soft elastomers, a factor of 10 reduction over past liquid metal stencil printing. Capacitors and resistive strain sensors are also demonstrated, showing the potential for creating stretchable conductors and devices.

  15. Dynamic interactions of Leidenfrost droplets on liquid metal surface

    NASA Astrophysics Data System (ADS)

    Ding, Yujie; Liu, Jing

    2016-09-01

    Leidenfrost dynamic interaction effects of the isopentane droplets on the surface of heated liquid metal were disclosed. Unlike conventional rigid metal, such conductive and deformable liquid metal surface enables the levitating droplets to demonstrate rather abundant and complex dynamics. The Leidenfrost droplets at different diameters present diverse morphologies and behaviors like rotation and oscillation. Depending on the distance between the evaporating droplets, they attract and repulse each other through the curved surfaces beneath them and their vapor flows. With high boiling point up to 2000 °C, liquid metal offers a unique platform for testing the evaporating properties of a wide variety of liquid even solid.

  16. Reconfigurable liquid metal circuits by Laplace pressure shaping

    NASA Astrophysics Data System (ADS)

    Cumby, Brad L.; Hayes, Gerard J.; Dickey, Michael D.; Justice, Ryan S.; Tabor, Christopher E.; Heikenfeld, Jason C.

    2012-10-01

    We report reconfigurable circuits formed by liquid metal shaping with <10 pounds per square inch (psi) Laplace and vacuum pressures. Laplace pressure drives liquid metals into microreplicated trenches, and upon release of vacuum, the liquid metal dewets into droplets that are compacted to 10-100× less area than when in the channel. Experimental validation includes measurements of actuation speeds exceeding 30 cm/s, simple erasable resistive networks, and switchable 4.5 GHz antennas. Such capability may be of value for next generation of simple electronic switches, tunable antennas, adaptive reflectors, and switchable metamaterials.

  17. Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel - Final Technical Report

    SciTech Connect

    William Anderson; James Tulenko; Bradley Rearden; Gary Harms

    2008-09-11

    The nuclear industry interest in advanced fuel and reactor design often drives towards fuel with uranium enrichments greater than 5 wt% 235U. Unfortunately, little data exists, in the form of reactor physics and criticality benchmarks, for uranium enrichments ranging between 5 and 10 wt% 235U. The primary purpose of this project is to provide benchmarks for fuel similar to what may be required for advanced light water reactors (LWRs). These experiments will ultimately provide additional information for application to the criticality-safety bases for commercial fuel facilities handling greater than 5 wt% 235U fuel.

  18. Development of an advanced antineutrino detector for reactor monitoring

    NASA Astrophysics Data System (ADS)

    Classen, T.; Bernstein, A.; Bowden, N. S.; Cabrera-Palmer, B.; Ho, A.; Jonkmans, G.; Kogler, L.; Reyna, D.; Sur, B.

    2015-01-01

    Here we present the development of a compact antineutrino detector for the purpose of nuclear reactor monitoring, improving upon a previously successful design. This paper will describe the design improvements of the detector which increases the antineutrino detection efficiency threefold over the previous effort. There are two main design improvements over previous generations of detectors for nuclear reactor monitoring: dual-ended optical readout and single volume detection mass. The dual-ended optical readout eliminates the need for fiducialization and increases the uniformity of the detector's optical response. The containment of the detection mass in a single active volume provides more target mass per detector footprint, a key design criteria for operating within a nuclear power plant. This technology could allow for real-time monitoring of the evolution of a nuclear reactor core, independent of reactor operator declarations of fuel inventories, and may be of interest to the safeguards community.

  19. Advances in Process Intensification through Multifunctional Reactor Engineering

    SciTech Connect

    2004-07-01

    This factsheet describes a research project whose goal is to develop the knowledge and tools required to develop and scale a novel multiphase pulse-flow, catalytic reactor for acid catalyzed C4 paraffin/olefin alkylation, to industrial dimensions.

  20. Liquid metal embrittlement susceptibility of ferritic martensitic steel in liquid lead alloys

    NASA Astrophysics Data System (ADS)

    Van den Bosch, J.; Bosch, R. W.; Sapundjiev, D.; Almazouzi, A.

    2008-06-01

    The susceptibility of the ferritic-martensitic steels T91 and EUROFER97 to liquid metal embrittlement (LME) in lead alloys has been examined under various conditions. T91, which is currently the most promising candidate material for the high temperature components of the future accelerator driven system (ADS) was tested in liquid lead bismuth eutectic (LBE), whereas the reduced activation steel, EUROFER97 which is under consideration to be the structural steel for fusion reactors was tested in liquid lead lithium eutectic. These steels, similar in microstructure and mechanical properties in the unirradiated condition were tested for their susceptibility to LME as function of temperature (150-450 °C) and strain rate (1 × 10 -3-1 × 10 -6 s -1). Also, the influence of pre-exposure and surface stress concentrators was evaluated for both steels in, respectively, liquid PbBi and PbLi environment. To assess the LME effect, results of the tests in liquid metal environment are compared with tests in air or inert gas environment. Although both unirradiated and irradiated smooth ferritic-martensitic steels do not show any or little deterioration of mechanical properties in liquid lead alloy environment compared to their mechanical properties in gas as function of temperature and strain rate, pre-exposure or the presence of surface stress concentrators does lead to a significant decrease in total elongation for certain test conditions depending on the type of liquid metal environment. The results are discussed in terms of wetting enhanced by liquid metal corrosion or crack initiation processes.

  1. Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports

    SciTech Connect

    1993-09-15

    This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

  2. Advanced-power-reactor design concepts and performance characteristics

    NASA Technical Reports Server (NTRS)

    Davison, H. W.; Kirchgessner, T. A.; Springborn, R. H.; Yacobucci, H. G.

    1974-01-01

    Five reactor cooling concepts which allow continued reactor operation following a single rupture of the coolant system are presented for application with the APR. These concepts incorporate convective cooling, double containment, or heat pipes to ensure operation after a coolant line rupture. Based on an evaluation of several control system concepts, a molybdenum clad, beryllium oxide sliding reflector located outside the pressure vessel is recommended.

  3. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    SciTech Connect

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  4. NUCLEAR MAGNETIC RELAXATION IN LIQUID METALS, ALLOYS, AND SALTS.

    DTIC Science & Technology

    NUCLEAR MAGNETIC RESONANCE, *ALKALI METAL ALLOYS, *LIQUID METALS, * SALTS , NUCLEAR MAGNETIC RESONANCE, NUCLEAR MAGNETIC RESONANCE, RELAXATION TIME... SODIUM , GALLIUM, SODIUM ALLOYS, THALLIUM, THALLIUM COMPOUNDS, MELTING, NUCLEAR SPINS, QUANTUM THEORY, OPERATORS(MATHEMATICS), BIBLIOGRAPHIES, INTEGRAL EQUATIONS, TEST EQUIPMENT, MATHEMATICAL ANALYSIS.

  5. Liquid metal heat sink for high-power laser diodes

    NASA Astrophysics Data System (ADS)

    Vetrovec, John; Litt, Amardeep S.; Copeland, Drew A.; Junghans, Jeremy; Durkee, Roger

    2013-02-01

    We report on the development of a novel, ultra-low thermal resistance active heat sink (AHS) for thermal management of high-power laser diodes (HPLD) and other electronic and photonic components. AHS uses a liquid metal coolant flowing at high speed in a miniature closed and sealed loop. The liquid metal coolant receives waste heat from an HPLD at high flux and transfers it at much reduced flux to environment, primary coolant fluid, heat pipe, or structure. Liquid metal flow is maintained electromagnetically without any moving parts. Velocity of liquid metal flow can be controlled electronically, thus allowing for temperature control of HPLD wavelength. This feature also enables operation at a stable wavelength over a broad range of ambient conditions. Results from testing an HPLD cooled by AHS are presented.

  6. Impact dynamics of oxidized liquid metal drops

    NASA Astrophysics Data System (ADS)

    Xu, Qin; Brown, Eric; Jaeger, Heinrich M.

    2013-04-01

    With exposure to air, many liquid metals spontaneously generate an oxide layer on their surface. In oscillatory rheological tests, this skin is found to introduce a yield stress that typically dominates the elastic response but can be tuned by exposing the metal to hydrochloric acid solutions of different concentration. We systematically studied the normal impact of eutectic gallium-indium (eGaIn) drops under different oxidation conditions and show how this leads to two different dynamical regimes. At low impact velocity (or low Weber number), eGaIn droplets display strong recoil and rebound from the impacted surface when the oxide layer is removed. In addition, the degree of drop deformation or spreading during impact is controlled by the oxide skin. We show that the scaling law known from ordinary liquids for the maximum spreading radius as a function of impact velocity can still be applied to the case of oxidized eGaIn if an effective Weber number We is employed that uses an effective surface tension factoring in the yield stress. In contrast, no influence on spreading from different oxidations conditions is observed for high impact velocity. This suggests that the initial kinetic energy is mostly damped by bulk viscous dissipation. Results from both regimes can be collapsed in an impact phase diagram controlled by two variables, the maximum spreading factor Pm=R0/Rm, given by the ratio of initial to maximum drop radius, and the impact number K=We/Re4/5, which scales with the effective Weber number We as well as the Reynolds number Re. The data exhibit a transition from capillary to viscous behavior at a critical impact number Kc≈0.1.

  7. Investigation of a liquid-metal magnetohydrodynamic power system.

    NASA Technical Reports Server (NTRS)

    Elliott, D. G.; Hays, L. G.; Cerini, D. J.; Bogdanoff, D. W.

    1972-01-01

    Liquid-metal magnetohydrodynamic power conversion is being investigated for nuclear-electric propulsion. A liquid-metal MHD converter has no moving mechanical parts and requires a heat source temperature of only 1300 K. Cycle efficiencies of 5% to 8% for single-stage converters and 10% for multistage converters appear attainable. The specific weight of a 240 kWe MHD power plant has been estimated as 30 kg/kWe with shielding for unmanned science missions.

  8. Solar-Driven Liquid-Metal MHD Generator

    NASA Technical Reports Server (NTRS)

    Hohl, F.; Lee, J. H.

    1982-01-01

    Liquid-metal magnetohydrodynamic (MHD) power generator with solar oven as its heat source has potential to produce electric power in space and on Earth at high efficiency. Generator focuses radiation from Sun to heat driving gas that pushes liquid metal past magnetic coil. Power is extracted directly from electric currents set up in conducting liquid. Using solar energy as fuel can save considerable costs and payload weight, compared to previous systems.

  9. Review of the proposed materials of construction for the SBWR and AP600 advanced reactors

    SciTech Connect

    Diercks, D.R.; Shack, W.J.; Chung, H.M.; Kassner, T.F.

    1994-06-01

    Two advanced light water reactor (LWR) concepts, namely the General Electric Simplified Boiling Water Reactor (SBWR) and the Westinghouse Advanced Passive 600 MWe Reactor (AP600), were reviewed in detail by Argonne National Laboratory. The objectives of these reviews were to (a) evaluate proposed advanced-reactor designs and the materials of construction for the safety systems, (b) identify all aging and environmentally related degradation mechanisms for the materials of construction, and (c) evaluate from the safety viewpoint the suitability of the proposed materials for the design application. Safety-related systems selected for review for these two LWRs included (a) reactor pressure vessel, (b) control rod drive system and reactor internals, (c) coolant pressure boundary, (d) engineered safety systems, (e) steam generators (AP600 only), (f) turbines, and (g) fuel storage and handling system. In addition, the use of cobalt-based alloys in these plants was reviewed. The selected materials for both reactors were generally sound, and no major selection errors were found. It was apparent that considerable thought had been given to the materials selection process, making use of lessons learned from previous LWR experience. The review resulted in the suggestion of alternate an possibly better materials choices in a number of cases, and several potential problem areas have been cited.

  10. Overview of the Consortium for the Advanced Simulation of Light Water Reactors (CASL)

    NASA Astrophysics Data System (ADS)

    Kulesza, Joel A.; Franceschini, Fausto; Evans, Thomas M.; Gehin, Jess C.

    2016-02-01

    The Consortium for Advanced Simulation of Light Water Reactors (CASL) was established in July 2010 for the purpose of providing advanced modeling and simulation solutions for commercial nuclear reactors. The primary goal is to provide coupled, higher-fidelity, usable modeling and simulation capabilities than are currently available. These are needed to address light water reactor (LWR) operational and safety performance-defining phenomena that are not yet able to be fully modeled taking a first-principles approach. In order to pursue these goals, CASL has participation from laboratory, academic, and industry partners. These partners are pursuing the solution of ten major "Challenge Problems" in order to advance the state-of-the-art in reactor design and analysis to permit power uprates, higher burnup, life extension, and increased safety. At present, the problems being addressed by CASL are primarily reactor physics-oriented; however, this paper is intended to introduce CASL to the reactor dosimetry community because of the importance of reactor physics modelling and nuclear data to define the source term for that community and the applicability and extensibility of the transport methods being developed.

  11. Liquid metal actuation by electrical control of interfacial tension

    NASA Astrophysics Data System (ADS)

    Eaker, Collin B.; Dickey, Michael D.

    2016-09-01

    By combining metallic electrical conductivity with low viscosity, liquid metals and liquid metal alloys offer new and exciting opportunities to serve as reconfigurable components of electronic, microfluidic, and electromagnetic devices. Here, we review the physics and applications of techniques that utilize voltage to manipulate the interfacial tension of liquid metals; such techniques include electrocapillarity, continuous electrowetting, electrowetting-on-dielectric, and electrochemistry. These techniques lower the interfacial tension between liquid metals and a surrounding electrolyte by driving charged species (or in the case of electrochemistry, chemical species) to the interface. The techniques are useful for manipulating and actuating liquid metals at sub-mm length scales where interfacial forces dominate. We focus on metals and alloys that are liquid near or below room temperature (mercury, gallium, and gallium-based alloys). The review includes discussion of mercury—despite its toxicity—because it has been utilized in numerous applications and it offers a way of introducing several phenomena without the complications associated with the oxide layer that forms on gallium and its alloys. The review focuses on the advantages, applications, opportunities, challenges, and limitations of utilizing voltage to control interfacial tension as a method to manipulate liquid metals.

  12. Sensitivity Studies of Advanced Reactors Coupled to High Temperature Electrolysis (HTE) Hydrogen Production Processes

    SciTech Connect

    Edwin A. Harvego; Michael G. McKellar; James E. O'Brien; J. Stephen Herring

    2007-04-01

    High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 °C to 950 °C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the steam or air sweep loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycle producing the highest efficiencies varied depending on the temperature range considered.

  13. Advances in process intensification through multifunctional reactor engineering.

    SciTech Connect

    Cooper, Marcia A.; Miller, James Edward; O'Hern, Timothy John; Gill, Walter; Evans, Lindsey R.

    2011-02-01

    A multifunctional reactor is a chemical engineering device that exploits enhanced heat and mass transfer to promote production of a desired chemical, combining more than one unit operation in a single system. The main component of the reactor system under study here is a vertical column containing packing material through which liquid(s) and gas flow cocurrently downward. Under certain conditions, a range of hydrodynamic regimes can be achieved within the column that can either enhance or inhibit a desired chemical reaction. To study such reactors in a controlled laboratory environment, two experimental facilities were constructed at Sandia National Laboratories. One experiment, referred to as the Two-Phase Experiment, operates with two phases (air and water). The second experiment, referred to as the Three-Phase Experiment, operates with three phases (immiscible organic liquid and aqueous liquid, and nitrogen). This report describes the motivation, design, construction, operational hazards, and operation of the both of these experiments. Data and conclusions are included.

  14. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    SciTech Connect

    Holbrook, Mark; Kinsey, Jim

    2015-03-01

    In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

  15. MODELING ASSUMPTIONS FOR THE ADVANCED TEST REACTOR FRESH FUEL SHIPPING CONTAINER

    SciTech Connect

    Rick J. Migliore

    2009-09-01

    The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is currently licensed per 10 CFR 71 to transport a fresh fuel element for either the Advanced Test Reactor, the University of Missouri Research Reactor (MURR), or the Massachusetts Institute of Technology Research Reactor (MITR-II). During the licensing process, the Nuclear Regulatory Commission (NRC) raised a number of issues relating to the criticality analysis, namely (1) lack of a tolerance study on the fuel and packaging, (2) moderation conditions during normal conditions of transport (NCT), (3) treatment of minor hydrogenous packaging materials, and (4) treatment of potential fuel damage under hypothetical accident conditions (HAC). These concerns were adequately addressed by modifying the criticality analysis. A tolerance study was added for both the packaging and fuel elements, full-moderation was included in the NCT models, minor hydrogenous packaging materials were included, and fuel element damage was considered for the MURR and MITR-II fuel types.

  16. Advanced reactors and associated fuel cycle facilities: safety and environmental impacts.

    PubMed

    Hill, R N; Nutt, W M; Laidler, J J

    2011-01-01

    The safety and environmental impacts of new technology and fuel cycle approaches being considered in current U.S. nuclear research programs are contrasted to conventional technology options in this paper. Two advanced reactor technologies, the sodium-cooled fast reactor (SFR) and the very high temperature gas-cooled reactor (VHTR), are being developed. In general, the new reactor technologies exploit inherent features for enhanced safety performance. A key distinction of advanced fuel cycles is spent fuel recycle facilities and new waste forms. In this paper, the performance of existing fuel cycle facilities and applicable regulatory limits are reviewed. Technology options to improve recycle efficiency, restrict emissions, and/or improve safety are identified. For a closed fuel cycle, potential benefits in waste management are significant, and key waste form technology alternatives are described.

  17. Advanced Fast Reactor - 100 (AFR-100) Report for the Technical Review Panel

    SciTech Connect

    Grandy, Christopher; Sienicki, James J.; Moisseytsev, Anton; Krajtl, Lubomir; Farmer, Mitchell T.; Kim, Taek K.; Middleton, B.

    2014-06-04

    This report is written to provide an overview of the Advanced Fast Reactor-100 in the requested format for a DOE technical review panel. This report was prepared with information that is responsive to the DOE Request for Information, DE-SOL-0003674 Advanced Reactor Concepts, dated February 27, 2012 from DOE’s Office of Nuclear Energy, Office of Nuclear Reactor Technologies. The document consists of two main sections. The first section is a summary of the AFR-100 design including the innovations that are incorporated into the design. The second section contains a series of tables that respond to the various questions requested of the reactor design team from the subject DOE RFI.

  18. 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…

  19. Advances in the development of wire mesh reactor for coal gasification studies.

    PubMed

    Zeng, Cai; Chen, Lei; Liu, Gang; Li, Wenhua; Huang, Baoming; Zhu, Hongdong; Zhang, Bing; Zamansky, Vladimir

    2008-08-01

    In an effort to further understand the coal gasification behavior in entrained-flow gasifiers, a high pressure and high temperature wire mesh reactor with new features was recently built. An advanced LABVIEW-based temperature measurement and control system were adapted. Molybdenum wire mesh with aperture smaller than 70 mum and type D thermocouple were used to enable high carbon conversion (>90%) at temperatures >1000 degrees C. Gaseous species from wire mesh reactor were quantified using a high sensitivity gas chromatography. The material balance of coal pyrolysis in wire mesh reactor was demonstrated for the first time by improving the volatile's quantification techniques.

  20. Advances in the development of wire mesh reactor for coal gasification studies - article no. 084102

    SciTech Connect

    Zeng, C.; Chen, L.; Liu, G.; Li, W.H.; Huang, B.M.; Zhu, H.D.; Zhang, B.; Zamansky, V.

    2008-08-15

    In an effort to further understand the coal gasification behavior in entrained-flow gasifiers, a high pressure and high temperature wire mesh reactor with new features was recently built. An advanced LABVIEW-based temperature measurement and control system were adapted. Molybdenum wire mesh with aperture smaller than 70 {mu} m and type D thermocouple were used to enable high carbon conversion ({gt}90%) at temperatures {gt}1000 {sup o}C. Gaseous species from wire mesh reactor were quantified using a high sensitivity gas chromatography. The material balance of coal pyrolysis in wire mesh reactor was demonstrated for the first time by improving the volatile's quantification techniques.

  1. Compatibility of materials with liquid metal targets for SNS

    SciTech Connect

    DiStefano, J.R.; Pawel, S.J.; DeVan, J.H.

    1996-06-01

    Several heavy liquid metals are candidates as the target in a spallation neutron source: Hg, Pb, Bi, and Pb-Bi eutectic. Systems with these liquid metals have been used in the past and a data-base on compatibility already exists. Two major compatibility issues have been identified when selecting a container material for these liquid metals: temperature gradient mass transfer and liquid metal embrittlement or LME. Temperature gradient mass transfer refers to dissolution of material from the high temperature portions of a system and its deposition in the lower temperature areas. Solution and deposition rate constants along with temperature, {Delta}T, and velocity are usually the most important parameters. For most candidate materials mass transfer corrosion has been found to be proportionately worse in Bi compared with Hg and Pb. For temperatures to {approx}550{degrees}C, ferritic/martensitic steels have been satisfactory in Pb or Hg systems and the maximum temperature can be extended to {approx}650{degrees}C with additions of inhibitors to the liquid metal, e.g. Mg, Ti, Zr. Above {approx}600{degrees}C, austenitic stainless steels have been reported to be unsatisfactory, largely because of the mass transfer of nickel. Blockage of flow from deposition of material is usually the life-limiting effect of this type of corrosion. However, mass transfer corrosion at lower temperatures has not been studied. At low temperatures (usually < 150{degrees}C), LME has been reported for some liquid metal/container alloy combinations. Liquid metal embrittlement, like hydrogen embrittlement, results in brittle fracture of a normally ductile material.

  2. Theory of the spin-1 bosonic liquid metal - Equilibrium properties of liquid metallic deuterium

    NASA Technical Reports Server (NTRS)

    Oliva, J.; Ashcroft, N. W.

    1984-01-01

    The theory of a two-component quantum fluid comprised of spin-1/2 fermions and nonzero spin bosons is examined. This system is of interest because it embodies a possible quantum liquid metallic phase of highly compressed deuterium. Bose condensation is assumed present and the two cases of nuclear-spin-polarized and -unpolarized systems are considered. A significant feature in the unpolarized case is the presence of a nonmagnetic mode with quadratic dispersion owing its existence to nonzero boson spin. The physical character of this mode is examined in detail within a Bogoliubov approach. The specific heat, bulk modulus, spin susceptibility, and thermal expansion are all determined. Striking contrasts in the specific heats and thermal-expansion coefficients of the liquid and corresponding normal solid metallic phase are predicted.

  3. Two-dimensional ultrasound Doppler velocimeter for flow mapping of unsteady liquid metal flows.

    PubMed

    Franke, S; Lieske, H; Fischer, A; Büttner, L; Czarske, J; Räbiger, D; Eckert, S

    2013-03-01

    We present a novel pulsed-wave ultrasound Doppler system for fluid flow investigations being able to determine two-dimensional vector fields of flow velocities. Electromagnetically-driven liquid metal flows appear as an attractive application field for such a measurement system. Two linear ultrasound transducer arrays each equipped with 25 transducer elements are used to measure the flow field in a square plane of 67×67 mm(2). The application of advanced processing methods as a multi-beam operation, an interlaced echo signal acquisition and a segmental array technique enable high data acquisition rates and concurrently a high spatial resolution, which have not been obtained so far for flow measurements in liquid metals. The extended pulsing strategy and essential operation principles such as the multiplexing electronic concept will be presented within this paper. The capabilities of the measuring system make it suitable for investigations of non-transparent, turbulent flows. Here, we present measurements of liquid metal flows driven by a rotating magnetic field for demonstration purposes. The measuring setup realized here reveals details of the swirling fluid motion in a horizontal section of a cube. Frame acquisition rates up to 30 fps were achieved for a complete two-dimensional flow mapping.

  4. Investigation on heavy liquid metal cooling of ADS fuel pin assemblies

    NASA Astrophysics Data System (ADS)

    Litfin, K.; Batta, A.; Class, A. G.; Wetzel, Th.; Stieglitz, R.

    2011-08-01

    In the framework of accelerator driven sub-critical reactor systems heavy liquid metals are considered as coolant for the reactor core and the spallation target. In particular lead or lead bismuth eutectic (LBE) exhibit efficient heat removal properties and high production rate of neutrons. However, the excellent heat conductivity of LBE-flows expressed by a low molecular Prandtl number of the order 10 -2 requires improved modeling of the turbulent heat transfer. Although various models for thermal hydraulics of LBE flows are existing, validated heat transfer correlations for ADS-relevant conditions are still missing. In order to validate the sub-channel codes and computational fluid dynamics codes used to design fuel assemblies, the comparison with experimental data is inevitable. Therefore, an experimental program composed of three major experiments, a single electrically heated rod, a 19-pin hexagonal water rod bundle and a LBE rod bundle, has been initiated at the Karlsruhe Liquid metal Laboratory (KALLA) of the Karlsruhe Institute of Technology, in order to quantify and separate the individual phenomena occurring in the momentum and energy transfer of a fuel assembly.

  5. Mixed convection in liquid metal flow in a horizontal duct with strong axial magnetic field

    NASA Astrophysics Data System (ADS)

    Zhang, Xuan; Zikanov, Oleg

    2016-11-01

    The work is motivated by design of self-cooled liquid-metal breeder blankets for Tokamak fusion reactors. Thermal convection caused by non-uniform internal heating in a liquid metal flow in a horizontal duct with strong axial magnetic field is analyzed numerically. Axial magnetic field is considered strong enough (the Hartmann number up to 104 corresponding to typical reactor condition) to suppress the streamwise variation of the flow, so a two-dimensional fully developed flow is studied. Duct walls are assumed to be thermally and electrically insulated. The non-uniform internal heat deposited by captured neutrons is fully diverted by the mean flow. Realistically high Grashof (up to 1011) and Reynolds (up to 106) numbers are considered. It is found that the state of the flow is strongly affected by the vertical stable stratification developing in response to the streamwise growth of mean temperature. Two flow regimes are identified: the regime with developed transverse convection at moderate Grashof numbers, and the regime, in which convection is suppressed at high Grashof numbers. Financial support was provided by the U.S. National Science Foundation (Grant CBET 1435269) and by the University of Michigan - Dearborn.

  6. Assessment of Startup Fuel Options for the GNEP Advanced Burner Reactor (ABR)

    SciTech Connect

    Jon Carmack; Kemal O. Pasamehmetoglu; David Alberstein

    2008-02-01

    The Global Nuclear Energy Program (GNEP) includes a program element for the development and construction of an advanced sodium cooled fast reactor to demonstrate the burning (transmutation) of significant quantities of minor actinides obtained from a separations process and fabricated into a transuranic bearing fuel assembly. To demonstrate and qualify transuranic (TRU) fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype is needed. The ABR would necessarily be started up using conventional metal alloy or oxide (U or U, Pu) fuel. Startup fuel is needed for the ABR for the first 2 to 4 core loads of fuel in the ABR. Following start up, a series of advanced TRU bearing fuel assemblies will be irradiated in qualification lead test assemblies in the ABR. There are multiple options for this startup fuel. This report provides a description of the possible startup fuel options as well as possible fabrication alternatives available to the program in the current domestic and international facilities and infrastructure.

  7. 10 CFR 830 Major Modification Determination for the Advanced Test Reactor Remote Monitoring and Management Capability

    SciTech Connect

    Bohachek, Randolph Charles

    2015-09-01

    The Advanced Test Reactor (ATR; TRA-670), which is located in the ATR Complex at Idaho National Laboratory, was constructed in the 1960s for the purpose of irradiating reactor fuels and materials. Other irradiation services, such as radioisotope production, are also performed at ATR. While ATR is safely fulfilling current mission requirements, assessments are continuing. These assessments intend to identify areas to provide defense–in-depth and improve safety for ATR. One of the assessments performed by an independent group of nuclear industry experts recommended that a remote accident management capability be provided. The report stated that: “contemporary practice in commercial power reactors is to provide a remote shutdown station or stations to allow shutdown of the reactor and management of long-term cooling of the reactor (i.e., management of reactivity, inventory, and cooling) should the main control room be disabled (e.g., due to a fire in the control room or affecting the control room).” This project will install remote reactor monitoring and management capabilities for ATR. Remote capabilities will allow for post scram reactor management and monitoring in the event the main Reactor Control Room (RCR) must be evacuated.

  8. SMAHTR - A Concept for a Small, Modular Advanced High Temperaure Reactor

    SciTech Connect

    Gehin, Jess C; Greene, Sherrell R; Holcomb, David Eugene; Carbajo, Juan J; Cisneros, Anselmo T; Corwin, William R; Ilas, Dan; Wilson, Dane F; Varma, Venugopal Koikal; Bradley, Eric Craig; Yoder, III, Graydon L

    2010-01-01

    Several new high temperature reactor concepts, referred to as Fluoride Salt Cooled High Temperature Reactors (FHRs), have been developed over the past decade. These FHRs use a liquid salt coolant combined with high temperature gas-cooled reactor fuels (TRISO) and graphite structural materials to provide a reactor that operates at very high temperatures and is scalable to large sizes perhaps exceeding 2400 MWt. This paper presents a new small FHR the Small Modular Advanced High Temperature Reactor or SmAHTR . SmAHTR is targeted at applications that require compact, high temperature heat sources either for high efficiency electricity production or process heat applications. A preliminary SmAHTR concept has been developed that delivers 125 MWt of energy in an integral primary system design that places all primary and decay heat removal heat exchangers inside the reactor vessel. The current reactor baseline concept utilizes a prismatic fuel block core, but multiple removable fuel assembly concepts are under evaluation as well. The reactor vessel size is such that it can be transported on a standard tractor-trailer to support simplified deployment. This paper will provide a summary of the current SmAHTR system concept and on-going technology and system architecture trades studies.

  9. Compiled reports on the applicability of selected codes and standards to advanced reactors

    SciTech Connect

    Benjamin, E.L.; Hoopingarner, K.R.; Markowski, F.J.; Mitts, T.M.; Nickolaus, J.R.; Vo, T.V.

    1994-08-01

    The following papers were prepared for the Office of Nuclear Regulatory Research of the U.S. Nuclear Regulatory Commission under contract DE-AC06-76RLO-1830 NRC FIN L2207. This project, Applicability of Codes and Standards to Advance Reactors, reviewed selected mechanical and electrical codes and standards to determine their applicability to the construction, qualification, and testing of advanced reactors and to develop recommendations as to where it might be useful and practical to revise them to suit the (design certification) needs of the NRC.

  10. Piping benchmark problems for the General Electric Advanced Boiling Water Reactor

    SciTech Connect

    Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

    1993-08-01

    To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boiling water reactor standard design, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard design. It will be required that the combined license holders demonstrate that their solutions to these problems are in agreement with the benchmark problem set.

  11. Advanced Reactor Safety Program – Stakeholder Interaction and Feedback

    SciTech Connect

    Szilard, Ronaldo H.; Smith, Curtis L.

    2014-08-01

    In the Spring of 2013, the Idaho National Laboratory (INL) began discussions with industry stakeholders on how to upgrade our safety analysis capabilities. The focus of these improvements would primarily be on advanced safety analysis capabilities that could help the nuclear industry analyze, understand, and better predict complex safety problems. The current environment in the DOE complex is such that recent successes in high performance computer modeling and simulation could lead the nuclear industry to benefit from these advances, as long as an effort to translate these advances into realistic applications is made. Upgrading the nuclear industry modeling analysis capabilities is a significant effort that would require participation and coordination from all industry segments: research, engineering, vendors, and operations. We focus here on interactions with industry stakeholders to develop sound advanced safety analysis applications propositions that could have a positive impact on industry long term operation, hence advancing the state of nuclear safety.

  12. Insights from the WGRISK workshop on the PSA of advanced and new reactors

    SciTech Connect

    Georgescu, G.; Ahn, K. I.; Amri, A.

    2012-07-01

    Probabilistic Safety Assessment /Probabilistic Risk Assessment for new and advanced reactors is recognized as an essential complement of the deterministic approaches to achieve improved safety and performances of new nuclear power plants, comparing to the operating plants. However, the development of PSA to these reactors is encountered to concurrent challenges, mainly due to the limited available design information, as well as due to potentially new initiating events, accident sequences and phenomena. The use of PSA in the decision making process is also challenging since the resulting PSA may not sufficiently reflect the future as-built, as-operated plant information. In order to address these aspects, the OECD/NEA/WGRISK initiated two coordinated tasks on 'PSA for Advanced Reactors' and 'PSA in the frame of Design and Commissioning of New NPPs'. In this context, a joint workshop was organized by OECD, during which related subjects were presented and discussed, including PSA for generation IV reactors, PSA for evolutionary reactors, PSA for small modular reactors, severe accidents and Level 2 PSA, Level 3 PSA and consequences analysis, digital I and C modeling, passive systems reliability, safety-security interface, as well as the results of the surveys performed in the frame of theses WGRISK tasks. (authors)

  13. Accelerated development of Zr-containing new generation ferritic steels for advanced nuclear reactors

    SciTech Connect

    Tan, Lizhen; Yang, Ying; Sridharan, K.

    2015-12-01

    The mission of the Nuclear Energy Enabling Technologies (NEET) program is to develop crosscutting technologies for nuclear energy applications. Advanced structural materials with superior performance at elevated temperatures are always desired for nuclear reactors, which can improve reactor economics, safety margins, and design flexibility. They benefit not only new reactors, including advanced light water reactors (LWRs) and fast reactors such as the sodium-cooled fast reactor (SFR) that is primarily designed for management of high-level wastes, but also life extension of the existing fleet when component exchange is needed. Developing and utilizing the modern materials science tools (experimental, theoretical, and computational tools) is an important path to more efficient alloy development and process optimization. The ultimate goal of this project is, with the aid of computational modeling tools, to accelerate the development of Zr-bearing ferritic alloys that can be fabricated using conventional steelmaking methods. The new alloys are expected to have superior high-temperature creep performance and excellent radiation resistance as compared to Grade 91. The designed alloys were fabricated using arc-melting and drop-casting, followed by hot rolling and conventional heat treatments. Comprehensive experimental studies have been conducted on the developed alloys to evaluate their hardness, tensile properties, creep resistance, Charpy impact toughness, and aging resistance, as well as resistance to proton and heavy ion (Fe2+) irradiation.

  14. Review of liquid metal heat pipe work at Los Alamos

    SciTech Connect

    Reid, R.S.; Merrigan, M.A.; Sena, J.T.

    1990-01-01

    A survey of space-power related liquid metal heat pipe work at Los Alamos National Laboratory is presented. Heat pipe development at Los Alamos has been on-going since 1963. Heat pipes were initially developed for thermionic nuclear-electrical power production in space. Since then Los Alamos has developed liquid metal heat pipes for numerous applications related to high temperature systems in both the space and terrestrial environments. Some of these applications include thermionic electrical generators, thermoelectric energy conversion (both in-core and direct radiation), thermal energy storage, hypersonic vehicle leading edge cooling, and heat pipe vapor laser cells. Some of the work performed at Los Alamos has been documented in internal reports that are often little-known. A representative description and summary of progress in space-related liquid metal heat pipe technology is provided followed by a reference section citing sources where these works may be found. 53 refs.

  15. Stretchable Metamaterial Absorber Using Liquid Metal-Filled Polydimethylsiloxane (PDMS)

    PubMed Central

    Kim, Kyeongseob; Lee, Dongju; Eom, Seunghyun; Lim, Sungjoon

    2016-01-01

    A stretchable metamaterial absorber is proposed in this study. The stretchability was achieved by liquid metal and polydimethylsiloxane (PDMS). To inject liquid metal, microfluidic channels were fabricated using PDMS powers and microfluidic-channel frames, which were built using a three-dimensional printer. A top conductive pattern and ground plane were designed after considering the easy injection of liquid metal. The proposed metamaterial absorber comprises three layers of PDMS substrate. The top layer is for the top conductive pattern, and the bottom layer is for the meandered ground plane. Flat PDMS layers were inserted between the top and bottom PDMS layers. The measured absorptivity of the fabricated absorber was 97.8% at 18.5 GHz, and the absorption frequency increased from 18.5 to 18.65 GHz as the absorber was stretched from its original length (5.2 cm) to 6.4 cm. PMID:27077861

  16. Review of liquid metal heat pipe work at Los Alamos

    SciTech Connect

    Reid, R.S.; Merrigan, M.A.; Sena, J.T. )

    1991-01-10

    A survey of space-power related liquid metal heat pipe work at Los Alamos National Laboratory is presented. Heat pipe development at Los Alamos has been on-going since 1963. Heat pipes were initially developed for thermionic nuclear-electrical power production in space. Since then Los Alamos has developed liquid metal heat pipes for numerous applications related to high temperature systems in both the space and terrestrial environments. Some of these applications include thermionic electrical generators, thermoelectric energy conversion (both in-core and direct radiation), thermal energy storage, hypersonic vehicle leading edge cooling, and heat pipe vapor laser cells. Some of the work performed at Los Alamos has been documented in internal reports that are often little-known. A representative description and summary of progress in space-related liquid metal heat pipe technology is provided followed by a reference section citing sources where these works may be found.

  17. Stretchable Metamaterial Absorber Using Liquid Metal-Filled Polydimethylsiloxane (PDMS).

    PubMed

    Kim, Kyeongseob; Lee, Dongju; Eom, Seunghyun; Lim, Sungjoon

    2016-04-11

    A stretchable metamaterial absorber is proposed in this study. The stretchability was achieved by liquid metal and polydimethylsiloxane (PDMS). To inject liquid metal, microfluidic channels were fabricated using PDMS powers and microfluidic-channel frames, which were built using a three-dimensional printer. A top conductive pattern and ground plane were designed after considering the easy injection of liquid metal. The proposed metamaterial absorber comprises three layers of PDMS substrate. The top layer is for the top conductive pattern, and the bottom layer is for the meandered ground plane. Flat PDMS layers were inserted between the top and bottom PDMS layers. The measured absorptivity of the fabricated absorber was 97.8% at 18.5 GHz, and the absorption frequency increased from 18.5 to 18.65 GHz as the absorber was stretched from its original length (5.2 cm) to 6.4 cm.

  18. Temperature-dependent liquid metal flowrate control device

    DOEpatents

    Carlson, Roger D.

    1978-01-01

    A temperature-dependent liquid metal flowrate control device includes a magnet and a ferromagnetic member defining therebetween a flow path for liquid metal, the ferromagnetic member being formed of a material having a curie temperature at which a change in the flow rate of the liquid metal is desired. According to the preferred embodiment the magnet is a cylindrical rod magnet axially disposed within a cylindrical member formed of a curie material and having iron pole pieces at the ends. A cylindrical iron shunt and a thin wall stainless steel barrier are disposed in the annulus between magnet and curie material. Below the curie temperature flow between steel barrier and curie material is impeded and above the curie temperature flow impedance is reduced.

  19. Emerging Applications of Liquid Metals Featuring Surface Oxides

    PubMed Central

    2014-01-01

    Gallium and several of its alloys are liquid metals at or near room temperature. Gallium has low toxicity, essentially no vapor pressure, and a low viscosity. Despite these desirable properties, applications calling for liquid metal often use toxic mercury because gallium forms a thin oxide layer on its surface. The oxide interferes with electrochemical measurements, alters the physicochemical properties of the surface, and changes the fluid dynamic behavior of the metal in a way that has, until recently, been considered a nuisance. Here, we show that this solid oxide “skin” enables many new applications for liquid metals including soft electrodes and sensors, functional microcomponents for microfluidic devices, self-healing circuits, shape-reconfigurable conductors, and stretchable antennas, wires, and interconnects. PMID:25283244

  20. Electrokinetic actuation of liquid metal for reconfigurable radio frequency devices

    NASA Astrophysics Data System (ADS)

    Gough, Ryan C.

    Liquid metals are an attractive material choice for designers wishing to combine the advantages of metals, such as high electrical conductivity, thermal conductivity, and reflectivity, with the inherently dynamic nature of fluids. Liquid metals have been utilized for a wide variety of applications, but their high electrical conductivity, surface smoothness, and linear response makes them especially attractive as tuning elements within reconfigurable radio frequency (RF) devices. The recent introduction of non-toxic liquid metal alloys onto the commercial market has further fueled interest in this versatile material. Early experiments with liquid metal as an RF tuning element have yielded promising results, but have largely depended on externally applied pressure to actuate the liquid metal. For commercial implementation this would necessitate the use of clunky and inefficient micro-pumps, which can require both high voltages and high power consumption. This reliance on hydraulic pumping has been a significant barrier to the incorporation of liquid metal as an RF tuning element in applications outside of a laboratory setting. Here, several electrical actuation techniques are demonstrated that allow for the rapid and repeatable actuation of non-toxic gallium alloys as tuning elements within reconfigurable RF devices. These techniques leverage the naturally high surface tension of liquid metals, as well as the unique electrochemistry of gallium-based alloys, to exercise wide-ranging and high fidelity control over both the metal's shape and position. Furthermore, this control is exercised with voltage and power levels that are each better than an order of magnitude below that achievable with conventional micro-pumps. This control does not require the constant application of actuation signals in order to maintain an actuated state, and can even be 'self-actuated', with the liquid metal supplying its own kinetic energy via the electrochemical conversion of its native

  1. 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.

  2. Precise Nuclear Data Measurements Possible with the NIFFTE fissionTPC for Advanced Reactor Designs

    NASA Astrophysics Data System (ADS)

    Towell, Rusty; Niffte Collaboration

    2015-10-01

    The Neutron Induced Fission Fragment Tracking Experiment (NIFFTE) Collaboration has applied the proven technology of Time Projection Chambers (TPC) to the task of precisely measuring fission cross sections. With the NIFFTE fission TPC, precise measurements have been made during the last year at the Los Alamos Neutron Science Center from both U-235 and Pu-239 targets. The exquisite tracking capabilities of this device allow the full reconstruction of charged particles produced by neutron beam induced fissions from a thin central target. The wealth of information gained from this approach will allow systematics to be controlled at the level of 1%. The fissionTPC performance will be presented. These results are critical to the development of advanced uranium-fueled reactors. However, there are clear advantages to developing thorium-fueled reactors such as Liquid Fluoride Thorium Reactors over uranium-fueled reactors. These advantages include improved reactor safety, minimizing radioactive waste, improved reactor efficiency, and enhanced proliferation resistance. The potential for using the fissionTPC to measure needed cross sections important to the development of thorium-fueled reactors will also be discussed.

  3. Thermal Control Using Liquid-Metal Bridge Switches

    NASA Technical Reports Server (NTRS)

    Hirsa, Amir H.; Olles, Joseph; Tilger, Christopher

    2013-01-01

    A short term effort (3-months) was undertaken to demonstrate the feasibility of a novel method to locally control the heat transfer rate and demonstrate the potential to achieve a turndown ratio of approximately 10:1. The technology had to be demonstrated to be at a TRL of 2-3, with a plan to advance it to a TRL 5-6. Here, we show that the concept recently developed in our laboratory, namely the pinned-contact, double droplet switch made by overfilling a hole drilled in a suitable substrate can be implemented with a low-melting temperature metal. When toggled near a second substrate, a liquid bridge can be reversibly connected or disconnected, on demand. We have shown experimentally that liquid-metal bridge switches can be made from gallium with a suitable choice of substrate materials, activation strategies, and control techniques. Individual as well as arrays of gallium bridge switches were shown to be feasible and can be robustly controlled. The very short response time of the bridge connection and disconnection (on the order of 1 millisecond) provides for utility in a wide range of applications. The liquid bridge switches may be controlled actively or passively. We have shown through computations and analysis that liquid bridge switches provide locally large turndown ratios (on the order of 103:1), so a relatively sparse packing of them would be needed to obtain the desired turndown ratio of 10:1. For the laboratory demonstrations, pressure activation was utilized. Simple designs for a passive control strategy are presented which are highly attractive for several reasons, including i) large turndown ratio, ii) no solid-moving parts, and iii) stable operation. Finally, we note that passive systems do not require any electronics for their control. This along with the relatively small molecular weight of candidate materials for the system, makes for a robust design outside of Earth?s magnetic field, where spacecraft are subject to significant radiation bombardment.

  4. Meeting Summary Advanced Light Water Reactor Fuels Industry Meeting Washington DC October 27 - 28, 2011

    SciTech Connect

    Not Listed

    2011-11-01

    The Advanced LWR Fuel Working Group first met in November of 2010 with the objective of looking 20 years ahead to the role that advanced fuels could play in improving light water reactor technology, such as waste reduction and economics. When the group met again in March 2011, the Fukushima incident was still unfolding. After the March meeting, the focus of the program changed to determining what we could do in the near term to improve fuel accident tolerance. Any discussion of fuels with enhanced accident tolerance will likely need to consider an advanced light water reactor with enhanced accident tolerance, along with the fuel. The Advanced LWR Fuel Working Group met in Washington D.C. on October 72-18, 2011 to continue discussions on this important topic.

  5. Method and system to directly produce electrical power within the lithium blanket region of a magnetically confined, deuterium-tritium (DT) fueled, thermonuclear fusion reactor

    DOEpatents

    Woolley, Robert D.

    1999-01-01

    A method for integrating liquid metal magnetohydrodynamic power generation with fusion blanket technology to produce electrical power from a thermonuclear fusion reactor located within a confining magnetic field and within a toroidal structure. A hot liquid metal flows from a liquid metal blanket region into a pump duct of an electromagnetic pump which moves the liquid metal to a mixer where a gas of predetermined pressure is mixed with the pressurized liquid metal to form a Froth mixture. Electrical power is generated by flowing the Froth mixture between electrodes in a generator duct. When the Froth mixture exits the generator the gas is separated from the liquid metal and both are recycled.

  6. Method and System to Directly Produce Electrical Power within the Lithium Blanket Region of a Magnetically Confined, Deuterium-Tritium (DT) Fueled, Thermonuclear Fusion Reactor

    SciTech Connect

    Woolley, Robert D.

    1998-09-22

    A method for integrating liquid metal magnetohydrodynamic power generation with fusion blanket technology to produce electrical power from a thermonuclear fusion reactor located within a confining magnetic field and within a toroidal structure. A hot liquid metal flows from a liquid metal blanket region into a pump duct of an electromagnetic pump which moves the liquid metal to a mixer where a gas of predetermined pressure is mixed with the pressurized liquid metal to form a Froth mixture. Electrical power is generated by flowing the Froth mixture between electrodes in a generator duct. When the Froth mixture exits the generator the gas is separated from the liquid metal and both are recycled.

  7. Tokamaks with high-performance resistive magnets: advanced test reactors and prospects for commercial applications

    SciTech Connect

    Bromberg, L.; Cohn, D.R.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

    1981-10-01

    Scoping studies have been made of tokamak reactors with high performance resistive magnets which maximize advantages gained from high field operation and reduced shielding requirements, and minimize resistive power requirements. High field operation can provide very high values of fusion power density and n tau/sub e/ while the resistive power losses can be kept relatively small. Relatively high values of Q' = Fusion Power/Magnet Resistive Power can be obtained. The use of high field also facilitates operation in the DD-DT advanced fuel mode. The general engineering and operational features of machines with high performance magnets are discussed. Illustrative parameters are given for advanced test reactors and for possible commercial reactors. Commercial applications that are discussed are the production of fissile fuel, electricity generation with and without fissioning blankets and synthetic fuel production.

  8. 2015 Groundwater Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Ponds

    SciTech Connect

    Lewis, Michael George

    2016-02-01

    This report summarizes radiological monitoring results from groundwater wells associated with the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste Ponds Reuse Permit (I-161-02). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  9. 2013 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect

    Mike Lewis

    2014-02-01

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  10. 2014 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect

    Lewis, Mike

    2015-02-01

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  11. 2012 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect

    Mike Lewis

    2013-02-01

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  12. 2011 Radiological Monitoring Results Associated with the Advanced Test Reactor Complex Cold Waste Pond

    SciTech Connect

    Mike Lewis

    2012-02-01

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site's Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  13. 2010 Radiological Monitoring Results Associated with the Advance Test Reactor Complex Cold Waste Pond

    SciTech Connect

    mike lewis

    2011-02-01

    This report summarizes radiological monitoring performed of the Idaho National Laboratory Site’s Advanced Test Reactor Complex Cold Waste wastewater prior to discharge into the Cold Waste Pond and of specific groundwater monitoring wells associated with the Industrial Wastewater Reuse Permit (#LA-000161-01, Modification B). All radiological monitoring is performed to fulfill Department of Energy requirements under the Atomic Energy Act.

  14. Advanced Computational Thermal Studies and their Assessment for Supercritical-Pressure Reactors (SCRs)

    SciTech Connect

    D. M. McEligot; J. Y. Yoo; J. S. Lee; S. T. Ro; E. Lurien; S. O. Park; R. H. Pletcher; B. L. Smith; P. Vukoslavcevic; J. M. Wallace

    2009-04-01

    The goal of this laboratory / university collaboration of coupled computational and experimental studies is the improvement of predictive methods for supercritical-pressure reactors. The general objective is to develop supporting knowledge needed of advanced computational techniques for the technology development of the concepts and their safety systems.

  15. Advanced propulsion engine assessment based on a cermet reactor

    NASA Technical Reports Server (NTRS)

    Parsley, Randy C.

    1993-01-01

    A preferred Pratt & Whitney conceptual Nuclear Thermal Rocket Engine (NTRE) has been designed based on the fundamental NASA priorities of safety, reliability, cost, and performance. The basic philosophy underlying the design of the XNR2000 is the utilization of the most reliable form of ultrahigh temperature nuclear fuel and development of a core configuration which is optimized for uniform power distribution, operational flexibility, power maneuverability, weight, and robustness. The P&W NTRE system employs a fast spectrum, cermet fueled reactor configured in an expander cycle to ensure maximum operational safety. The cermet fuel form provides retention of fuel and fission products as well as high strength. A high level of confidence is provided by benchmark analysis and independent evaluations.

  16. Liquid metal magnetohydrodynamics (LMMHD) technology transfer feasibility study. Volume 1: Summary

    NASA Technical Reports Server (NTRS)

    Phen, R. L.; Hays, L. G.; Alper, M. E.

    1973-01-01

    The potential application of liquid metal magnetohydrodynamics (LMMHD) to central station utility power generation through the period to 1990 is examined. Included are: (1) a description of LMMHD and a review of its development status, (2) LMMHD preliminary design for application to central station utility power generation, (3) evaluation of LMMHD in comparison with conventional and other advanced power generation systems and (4) a technology development plan. One of the major conclusions found is that the most economic and technically feasible application of LMMHD is a topping cycle to a steam plant, taking advantage of high temperatures available but not usable by the steam cycle.

  17. Liquid Metal Walls, Lithium, And Low Recycling Boundary Conditions In Tokamaks

    SciTech Connect

    R. Majeski

    2010-01-15

    At present, the only solid material believed to be a viable option for plasma-facing components (PFCs) in a fusion reactor is tungsten. Operated at the lower temperatures typical of present-day fusion experiments, tungsten is known to suffer from surface degradation during long-term exposure to helium-containing plasmas, leading to reduced thermal conduction to the bulk, and enhanced erosion. Existing alloys are also quite brittle at temperatures under 700oC. However, at a sufficiently high operating temperature (700 - 1000 oC), tungsten is selfannealing and it is expected that surface damage will be reduced to the point where tungsten PFCs will have an acceptable lifetime in a reactor environment. The existence of only one potentially viable option for solid PFCs, though, constitutes one of the most significant restrictions on design space for DEMO and follow-on fusion reactors. In contrast, there are several candidates for liquid metal-based PFCs, including gallium, tin, lithium, and tin-lithium eutectics. We will discuss options for liquid metal walls in tokamaks, looking at both high and low recycling materials. We will then focus in particular on one of the candidate liquids, lithium. Lithium is known to have a high chemical affinity for hydrogen, and has been shown in test stands1 and fusion experiments2,3 to produce a low recycling surface, especially when liquid. Because it is also low-Z and is usable in a tokamak over a reasonable temperature range (200 - 400 oC), it has been now been used as a PFC in several confinement experiments (TFTR, T11- M, CDX-U, NSTX, FTU, and TJ-II), with favorable results. The consequences of substituting low recycling walls for the traditional high recycling variety on tokamak equilibria are very extensive. We will discuss some of the expected modifications, briefly reviewing experimental results, and comparing the results to expectations.

  18. Final Assembly and Initial Irradiation of the First Advanced Gas Reactor Fuel Development and Qualification Experiment in the Advanced Test Reactor

    SciTech Connect

    S. B. Grover

    2007-05-01

    The United States Department of Energy’s Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating eight separate low enriched uranium (LEU) oxycarbide (UCO) tri-isotropic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The AGR fuel experiments will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing.1,2 The experiments, which will each consist of six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The final design phase for the first experiment was completed in 2005, and the fabrication and assembly of the first experiment test train (designated AGR-1) as well as the support systems and fission product monitoring system that will monitor and control the experiment

  19. Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

    SciTech Connect

    Bucknor, Matthew D.; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin

    2016-01-01

    Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended due to deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Centering on an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive reactor cavity cooling system following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. While this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability for the reactor cavity cooling system (and the reactor system in general) to the postulated transient event.

  20. Requirements for Prognostic Health Management of Passive Components in Advanced Small Modular Reactors

    SciTech Connect

    Meyer, Ryan M.; Coble, Jamie B.; Ramuhalli, Pradeep

    2013-08-01

    Advanced small modular reactors (aSMRs), which are based on modularization of advanced reactor concepts, may provide a longer-term alternative to traditional light-water reactors and near term small modular reactors (SMRs), which are based on integral pressurized water reactor (iPWR) concepts. aSMRs are conceived for applications in remote locations and for diverse missions that include providing process or district heating, water desalination, and hydrogen production. Several challenges exist with respect to cost-effective operations and maintenance (O&M) of aSMRs, including the impacts of aggressive operating environments and modularity, and limiting these costs and staffing needs will be essential to ensuring the economic feasibility of aSMR deployment. In this regard, prognostic health management (PHM) systems have the potential to play a vital role in supporting the deployment of aSMR systems. This paper identifies requirements and technical gaps associated with implementation of PHM systems for passive aSMR components.

  1. A superconductor to superfluid phase transition in liquid metallic hydrogen.

    PubMed

    Babaev, Egor; Sudbø, Asle; Ashcroft, N W

    2004-10-07

    Although hydrogen is the simplest of atoms, it does not form the simplest of solids or liquids. Quantum effects in these phases are considerable (a consequence of the light proton mass) and they have a demonstrable and often puzzling influence on many physical properties, including spatial order. To date, the structure of dense hydrogen remains experimentally elusive. Recent studies of the melting curve of hydrogen indicate that at high (but experimentally accessible) pressures, compressed hydrogen will adopt a liquid state, even at low temperatures. In reaching this phase, hydrogen is also projected to pass through an insulator-to-metal transition. This raises the possibility of new state of matter: a near ground-state liquid metal, and its ordered states in the quantum domain. Ordered quantum fluids are traditionally categorized as superconductors or superfluids; these respective systems feature dissipationless electrical currents or mass flow. Here we report a topological analysis of the projected phase of liquid metallic hydrogen, finding that it may represent a new type of ordered quantum fluid. Specifically, we show that liquid metallic hydrogen cannot be categorized exclusively as a superconductor or superfluid. We predict that, in the presence of a magnetic field, liquid metallic hydrogen will exhibit several phase transitions to ordered states, ranging from superconductors to superfluids.

  2. Turbulent convection in liquid metal with and without rotation.

    PubMed

    King, Eric M; Aurnou, Jonathan M

    2013-04-23

    The magnetic fields of Earth and other planets are generated by turbulent, rotating convection in liquid metal. Liquid metals are peculiar in that they diffuse heat more readily than momentum, quantified by their small Prandtl numbers, Pr < 1. Most analog models of planetary dynamos, however, use moderate Pr fluids, and the systematic influence of reducing Pr is not well understood. We perform rotating Rayleigh-Bénard convection experiments in the liquid metal gallium (Pr = 0.025) over a range of nondimensional buoyancy forcing (Ra) and rotation periods (E). Our primary diagnostic is the efficiency of convective heat transfer (Nu). In general, we find that the convective behavior of liquid metal differs substantially from that of moderate Pr fluids, such as water. In particular, a transition between rotationally constrained and weakly rotating turbulent states is identified, and this transition differs substantially from that observed in moderate Pr fluids. This difference, we hypothesize, may explain the different classes of magnetic fields observed on the Gas and Ice Giant planets, whose dynamo regions consist of Pr < 1 and Pr > 1 fluids, respectively.

  3. Wideband-Switchable Metamaterial Absorber Using Injected Liquid Metal

    NASA Astrophysics Data System (ADS)

    Kim, Hyung Ki; Lee, Dongju; Lim, Sungjoon

    2016-08-01

    Metamaterial absorbers can provide good solutions for radar-cross-section (RCS) reduction. In spite of their attractive features of thinness, lightness, and low cost, resonant metamaterial absorbers have a drawback of narrow bandwidth. For practical radar applications, wideband absorbers are necessary. In this paper, we propose a wideband-switchable metamaterial absorber using liquid metal. In order to reduce RCS both for X-band and C-band, the switchable Jerusalem cross (JC) resonator is introduced. The JC resonator consists of slotted circular rings, chip resistors, and microfluidic channels. The JC resonator is etched on a flexible printed circuit board (FPCB), and the microfluidic channels are laser-etched on a polydimethylsiloxane (PDMS) material. The proposed absorber can switch the absorption frequency band by injecting a liquid metal alloy into the channels. The performance of the absorber was demonstrated through full-wave simulation and through measurements employing prototypes. The experimental results showed absorption ratios of over 90% from 7.43 GHz to 14.34 GHz, and from 5.62 GHz to 7.3 GHz, with empty channels and liquid metal-filled channels, respectively. Therefore, the absorption band was successfully switched between the C-band (4–8 GHz) and the X-band (8–12 GHz) by injecting liquid metal eutectic gallium indium alloy (EGaIn) into the channels.

  4. Wideband-Switchable Metamaterial Absorber Using Injected Liquid Metal.

    PubMed

    Kim, Hyung Ki; Lee, Dongju; Lim, Sungjoon

    2016-08-22

    Metamaterial absorbers can provide good solutions for radar-cross-section (RCS) reduction. In spite of their attractive features of thinness, lightness, and low cost, resonant metamaterial absorbers have a drawback of narrow bandwidth. For practical radar applications, wideband absorbers are necessary. In this paper, we propose a wideband-switchable metamaterial absorber using liquid metal. In order to reduce RCS both for X-band and C-band, the switchable Jerusalem cross (JC) resonator is introduced. The JC resonator consists of slotted circular rings, chip resistors, and microfluidic channels. The JC resonator is etched on a flexible printed circuit board (FPCB), and the microfluidic channels are laser-etched on a polydimethylsiloxane (PDMS) material. The proposed absorber can switch the absorption frequency band by injecting a liquid metal alloy into the channels. The performance of the absorber was demonstrated through full-wave simulation and through measurements employing prototypes. The experimental results showed absorption ratios of over 90% from 7.43 GHz to 14.34 GHz, and from 5.62 GHz to 7.3 GHz, with empty channels and liquid metal-filled channels, respectively. Therefore, the absorption band was successfully switched between the C-band (4-8 GHz) and the X-band (8-12 GHz) by injecting liquid metal eutectic gallium indium alloy (EGaIn) into the channels.

  5. Turbulent convection in liquid metal with and without rotation

    PubMed Central

    King, Eric M.; Aurnou, Jonathan M.

    2013-01-01

    The magnetic fields of Earth and other planets are generated by turbulent, rotating convection in liquid metal. Liquid metals are peculiar in that they diffuse heat more readily than momentum, quantified by their small Prandtl numbers, . Most analog models of planetary dynamos, however, use moderate fluids, and the systematic influence of reducing is not well understood. We perform rotating Rayleigh–Bénard convection experiments in the liquid metal gallium over a range of nondimensional buoyancy forcing and rotation periods (E). Our primary diagnostic is the efficiency of convective heat transfer . In general, we find that the convective behavior of liquid metal differs substantially from that of moderate fluids, such as water. In particular, a transition between rotationally constrained and weakly rotating turbulent states is identified, and this transition differs substantially from that observed in moderate fluids. This difference, we hypothesize, may explain the different classes of magnetic fields observed on the Gas and Ice Giant planets, whose dynamo regions consist of and fluids, respectively. PMID:23569262

  6. Wideband-Switchable Metamaterial Absorber Using Injected Liquid Metal

    PubMed Central

    Kim, Hyung Ki; Lee, Dongju; Lim, Sungjoon

    2016-01-01

    Metamaterial absorbers can provide good solutions for radar-cross-section (RCS) reduction. In spite of their attractive features of thinness, lightness, and low cost, resonant metamaterial absorbers have a drawback of narrow bandwidth. For practical radar applications, wideband absorbers are necessary. In this paper, we propose a wideband-switchable metamaterial absorber using liquid metal. In order to reduce RCS both for X-band and C-band, the switchable Jerusalem cross (JC) resonator is introduced. The JC resonator consists of slotted circular rings, chip resistors, and microfluidic channels. The JC resonator is etched on a flexible printed circuit board (FPCB), and the microfluidic channels are laser-etched on a polydimethylsiloxane (PDMS) material. The proposed absorber can switch the absorption frequency band by injecting a liquid metal alloy into the channels. The performance of the absorber was demonstrated through full-wave simulation and through measurements employing prototypes. The experimental results showed absorption ratios of over 90% from 7.43 GHz to 14.34 GHz, and from 5.62 GHz to 7.3 GHz, with empty channels and liquid metal-filled channels, respectively. Therefore, the absorption band was successfully switched between the C-band (4–8 GHz) and the X-band (8–12 GHz) by injecting liquid metal eutectic gallium indium alloy (EGaIn) into the channels. PMID:27546310

  7. Performance of solar thermal systems with liquid metal MHD conversion

    NASA Astrophysics Data System (ADS)

    Pierson, E. S.; Jackson, W. D.; Berry, G.; Petrick, M.; Dennis, C.

    1984-06-01

    Liquid metal magnetohydrodynamic conversion (LMMHD) is found to be compatible with concentrating solar receivers employing a liquid metal as a heat transfer medium and offers significant increases in the system thermal efficiency over the 33% considered attainable with conventional turbo-machinery. There are two candidate liquid metals - sodium and lithium. With sodium at a temperature of 1150 F (922 K), the maximum calculated efficiency is 39.5% while with lithium at 1400 F (1033 K) a peak efficiency for 46.5% is predicted. Up to two percentage points may be added by temperature increase and/or parameter limit relaxation in the sodium case. The sodium steam heat exchanger is eliminated in liquid metal systems. Where LMMHD systems employ the same working fluid as the solar receiver, no recirculating pump is required as pumping power is provided directly by the cycle. For sodium, coupling with either a gas turbine or a steam turbine is beneficial and provides similar performance. With lithium, the gas turbine cycle is clearly superior.

  8. Testing of an advanced thermochemical conversion reactor system

    SciTech Connect

    Not Available

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective of this project was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions. 6 refs., 59 figs., 29 tabs.

  9. Testing of an advanced thermochemical conversion reactor system

    NASA Astrophysics Data System (ADS)

    1990-01-01

    This report presents the results of work conducted by MTCI to verify and confirm experimentally the ability of the MTCI gasification process to effectively generate a high-quality, medium-Btu gas from a wider variety of feedstock and waste than that attainable in air-blown, direct gasification systems. The system's overall simplicity, due to the compact nature of the pulse combustor, and the high heat transfer rates attainable within the pulsating flow resonance tubes, provide a decided and near-term potential economic advantage for the MTCI indirect gasification system. The primary objective was the design, construction, and testing of a Process Design Verification System for an indirectly heated, thermochemical fluid-bed reactor and a pulse combustor an an integrated system that can process alternative renewable sources of energy such as biomass, black liquor, municipal solid waste and waste hydrocarbons, including heavy oils into a useful product gas. The test objectives for the biomass portion of this program were to establish definitive performance data on biomass feedstocks covering a wide range of feedstock qualities and characteristics. The test objectives for the black liquor portion of this program were to verify the operation of the indirect gasifier on commercial black liquor containing 65 percent solids at several temperature levels and to characterize the bed carbon content, bed solids particle size and sulfur distribution as a function of gasification conditions.

  10. Examination of loop-operator-initiated events for the advanced test reactor

    SciTech Connect

    Durney, J.L.; Majumdar, D.

    1989-01-01

    The Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory is a unique high-flux test reactor having nine major test positions for irradiation of reactor materials. These test positions contain inpile tubes (IPT) that are connected to external piping and equipment (loops) to provide the high-temperature, high-pressure environment for the testing. The design of the core has intimately integrated the IPTs into the fuel region by means of a serpentine fuel arrangement resulting in a close reactivity coupling between the loop thermal hydraulics and the core. Consequently, operator actions potentially have an impact on the reactor power transients resulting from off-normal conditions in these facilities. This paper examines these operator-initiated events and their consequences. The analysis of loop-operator-initiated events indicates there is no damage to the reactor core even when assuming no operator intervention for mitigation. However, analysis does assume a scram occurs when required by the reactor protection systems.

  11. A new advanced fixed in-core instrumentation for a PWR reactor

    NASA Astrophysics Data System (ADS)

    Barbet, M.; Guillery, M.

    1981-06-01

    Gamma thermometer studies have been done at E.D.F. for four years. These studies started in France with a feasibility study in 1975. E.D.F.'s scope was to develop a new fixed "in-core" instrumentation for PWR based on the gamma heat measurements. The advanced gamma thermometer design has been done in such a way to be able to manufacture strings of 6 to 9 detectors each. The results of gamma thermometer make up in 1976 were encouraging and E.D.F. went on to develop a gamma thermometer assembly for a reactor application. Before being mounted on the reactor vessel, the gamma thermometer strings are calibrated in a loop test by means of an electrical current giving the ΔT versus the specific power ( W/ g). The loop test simulates the thermohydraulic conditions in the reactor tube guide. Two gamma thermometer strings have been installed in the BUGEY 5 reactor since June 1979. Four gamma thermometer strings are provided for insertion in the TRICASTIN 2 reactor and four more gamma thermometer strings are manufactured to be ready for the start up of the TRICASTIN 3 reactor in 1980.

  12. Experiments with Liquid Metal Walls: Status of the Lithium Tokamak Experiment

    SciTech Connect

    Kaita, Robert; Boyle, Dennis; Gray, Timothy; Granstedt, Erik; Hammett, Gregory; Jacobson, Craig M; Jones, Andrew; Kozub, Thomas; Kugel, Henry; Leblanc, Benoit; Logan, Nicholas; Lucia, Matthew; Lundberg, Daniel; Majeski, Richard; Mansfield, Dennis; Menard, Jonathan; Spaleta, Jeffrey; Strickler, Trevor; Timberlak, John

    2010-02-16

    Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The Lithium Tokamak Experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the Current Drive Experiment-Upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in Ohmically-heated tokamak plasmas was achieved with a toroidal liquid lithium limiter. The LTX extends this liquid lithium PFC by using a conducting conformal shell that almost completely surrounds the plasma. By heating the shell, a lithium coating on the plasma-facing side can be kept liquefied. A consequence of the low-recycling conditions from liquid lithium walls is the need for efficient plasma fueling. For this purpose, a molecular cluster injector is being developed. Future plans include the installation of a neutral beam for core plasma fueling, and also ion temperature measurements using charge-exchange recombination spectroscopy. Low edge recycling is also predicted to reduce temperature gradients that drive drift wave turbulence. Gyrokinetic simulations are in progress to calculate fluctuation levels and transport for LTX plasmas, and new fluctuation diagnostics are under development to test these predictions. __________________________________________________

  13. Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

    SciTech Connect

    Tsai, H.; Gomes, I.C.; Smith, D.L.; Palmer, A.J.; Ingram, F.W.; Wiffen, F.W.

    1998-09-01

    The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

  14. Advanced Reactors Transition program fiscal year 1998 multi-year work plan

    SciTech Connect

    Gantt, D.A.

    1997-09-25

    The mission of the Advanced Reactors Transition program is two-fold. First, the program is to maintain the Fast Flux Test Facility (FFTF) and the Fuels and Materials Examination Facility (FMEF) in Standby to support a possible future role in the tritium production strategy. Secondly, the program is to continue deactivation activities which do not conflict with the Standby directive. On-going deactivation activities include the processing of non-usable, irradiated, FFTF components for storage or disposal; deactivation of Nuclear Energy legacy test facilities; and deactivation of the Plutonium Recycle Test Reactor (PRTR) facility, 309 Building.

  15. DEVELOPMENT OF HUMAN FACTORS ENGINEERING GUIDANCE FOR SAFETY EVALUATIONS OF ADVANCED REACTORS.

    SciTech Connect

    O'HARA, J.; PERSENSKY, J.; SZABO, A.

    2006-10-01

    Advanced reactors are expected to be based on a concept of operations that is different from what is currently used in today's reactors. Therefore, regulatory staff may need new tools, developed from the best available technical bases, to support licensing evaluations. The areas in which new review guidance may be needed and the efforts underway to address the needs will be discussed. Our preliminary results focus on some of the technical issues to be addressed in three areas for which new guidance may be developed: automation and control, operations under degraded conditions, and new human factors engineering methods and tools.

  16. Contribution to the validation of MCNP neutronics design of the advanced neutron source reactor

    SciTech Connect

    Rubio, G.A.; Ougouag, A.M.; Wemple, C.A.; Ryskamp, J.M. )

    1993-01-01

    In the research and development plan of the advanced neutron source (ANS) reactor, one of the steps planned for neutronics design methods validation is to compare computational model results and experimental results for a critical facility exhibiting similarities with the ANS reactor. One such facility is the FOEHN experiment. In this paper, an MCNP model of the FOEHN experiment is developed, and its results are compared with experimental data from the literature. The MCNP models reproduces measured quantities of interest with a high level of agreement.

  17. Advanced Test Reactor In-Canal Ultrasonic Scanner: Experiment Design and Initial Results on Irradiated Plates

    SciTech Connect

    D. M. Wachs; J. M. Wight; D. T. Clark; J. M. Williams; S. C. Taylor; D. J. Utterbeck; G. L. Hawkes; G. S. Chang; R. G. Ambrosek; N. C. Craft

    2008-09-01

    An irradiation test device has been developed to support testing of prototypic scale plate type fuels in the Advanced Test Reactor. The experiment hardware and operating conditions were optimized to provide the irradiation conditions necessary to conduct performance and qualification tests on research reactor type fuels for the RERTR program. The device was designed to allow disassembly and reassembly in the ATR spent fuel canal so that interim inspections could be performed on the fuel plates. An ultrasonic scanner was developed to perform dimensional and transmission inspections during these interim investigations. Example results from the AFIP-2 experiment are presented.

  18. Research reactor usage at the Idaho National Engineering Laboratory in support of university research and education

    SciTech Connect

    Woodall, D.M.; Dolan, T.J.; Stephens, A.G. )

    1990-01-01

    The Idaho National Engineering Laboratory is a US Department of Energy laboratory which has a substantial history of research and development in nuclear reactor technologies. There are a number of available nuclear reactor facilities which have been incorporated into the research and training needs of university nuclear engineering programs. This paper addresses the utilization of the Advanced Reactivity Measurement Facility (ARMF) and the Coupled Fast Reactivity Measurement Facility (CFRMF) for thesis and dissertation research in the PhD program in Nuclear Science and Engineering by the University of Idaho and Idaho State University. Other reactors at the INEL are also being used by various members of the academic community for thesis and dissertation research, as well as for research to advance the state of knowledge in innovative nuclear technologies, with the EBR-II facility playing an essential role in liquid metal breeder reactor research. 3 refs.

  19. Reactor physics studies for the Advanced Fuel Cycle Initiative (AFCI) Reactor-Accelerator Coupling Experiments (RACE) Project

    NASA Astrophysics Data System (ADS)

    Stankovskiy, Evgeny Yuryevich

    In the recently completed RACE Project of the AFCI, accelerator-driven subcritical systems (ADS) experiments were conducted to develop technology of coupling accelerators to nuclear reactors. In these experiments electron accelerators induced photon-neutron reactions in heavy-metal targets to initiate fission reactions in ADS. Although the Idaho State University (ISU) RACE ADS was constructed only to develop measurement techniques for advanced experiments, many reactor kinetics experiments were conducted there. In the research reported in this dissertation, a method was developed to calculate kinetics parameters for measurement and calculation of the reactivity of ADS, a safety parameter that is necessary for control and monitoring of power production. Reactivity is measured in units of fraction of delayed versus prompt neutron from fission, a quantity that cannot be directly measured in far-subcritical reactors such as the ISU RACE configuration. A new technique is reported herein to calculate it accurately and to predict kinetic behavior of a far-subcritical ADS. Experiments conducted at ISU are first described and experimental data are presented before development of the kinetic theory used in the new computational method. Because of the complexity of the ISU ADS, the Monte-Carlo method as applied in the MCNP code is most suitable for modeling reactor kinetics. However, the standard method of calculating the delayed neutron fraction produces inaccurate values. A new method was developed and used herein to evaluate actual experiments. An advantage of this method is that its efficiency is independent of the fission yield of delayed neutrons, which makes it suitable for fuel with a minor actinide component (e.g. transmutation fuels). The implementation of this method is based on a correlated sampling technique which allows the accurate evaluation of delayed and prompt neutrons. The validity of the obtained results is indicated by good agreement between experimental

  20. Advanced Fusion Reactors for Space Propulsion and Power Systems

    NASA Technical Reports Server (NTRS)

    Chapman, John J.

    2011-01-01

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles "exhaust" momentum can be used directly to produce high ISP thrust and also offer possibility of power conversion into electricity. p- 11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  1. Advanced Fusion Reactors for Space Propulsion and Power Systems

    SciTech Connect

    Chapman, John J.

    2011-06-15

    In recent years the methodology proposed for conversion of light elements into energy via fusion has made steady progress. Scientific studies and engineering efforts in advanced fusion systems designs have introduced some new concepts with unique aspects including consideration of Aneutronic fuels. The plant parameters for harnessing aneutronic fusion appear more exigent than those required for the conventional fusion fuel cycle. However aneutronic fusion propulsion plants for Space deployment will ultimately offer the possibility of enhanced performance from nuclear gain as compared to existing ionic engines as well as providing a clean solution to Planetary Protection considerations and requirements. Proton triggered 11Boron fuel (p- 11B) will produce abundant ion kinetic energy for In-Space vectored thrust. Thus energetic alpha particles' exhaust momentum can be used directly to produce high Isp thrust and also offer possibility of power conversion into electricity. p-11B is an advanced fusion plant fuel with well understood reaction kinematics but will require some new conceptual thinking as to the most effective implementation.

  2. Integral Fast Reactor concept

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative LMR concept, being developed at Argonne National Laboratory, that fully exploits the inherent properties of liquid metal cooling and metallic fuel to achieve breakthroughs in economics and inherent safety. This paper describes key features and potential advantages of the IFR concept, technology development status, fuel cycle economics potential, and future development path.

  3. Liquid Metals: Stretchable, High-k Dielectric Elastomers through Liquid-Metal Inclusions (Adv. Mater. 19/2016).

    PubMed

    Bartlett, Michael D; Fassler, Andrew; Kazem, Navid; Markvicka, Eric J; Mandal, Pratiti; Majidi, Carmel

    2016-05-01

    An all-soft-matter composite consisting of liquid metal microdroplets embedded in a soft elastomer matrix is presented by C. Majidi and co-workers on page 3726. This composite exhibits a high dielectric constant while maintaining exceptional elasticity and compliance. The image shows the composite's microstructure captured by 3D X-ray imaging using a nano-computed tomographic scanner.

  4. The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels

    SciTech Connect

    Forsberg, C.W.; Peterson, P.F.; Ott, L.

    2004-10-06

    Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases

  5. NEW INSTRUMENTS AND METHODS OF MEASUREMENTS: Liquid-metal ion emitters

    NASA Astrophysics Data System (ADS)

    Gabovich, M. D.

    1983-05-01

    This article describes and discusses the fundamental laws of ion emission from liquid-metal tips in a strong electric field. The widespread views of a liquid-metal emitter as being the smoothed tip of a Taylor cone are examined critically. The instability of a liquid metal in an electric field is discussed, and in line with this, an alternative concept is given of a sharp-tipped electrohydrodynamic emitter. The prospects for applying liquid-metal ion emitters are noted.

  6. In-Situ Creep Testing Capability for the Advanced Test Reactor

    SciTech Connect

    B. G. Kim; J. L. Rempe; D. L. Knudson; K. G. Condie; B. H. Sencer

    2012-09-01

    An instrumented creep testing capability is being developed for specimens irradiated in Pressurized Water Reactor (PWR) coolant conditions at the Advanced Test Reactor (ATR). The test rig has been developed such that samples will be subjected to stresses ranging from 92 to 350 MPa at temperatures between 290 and 370 °C up to at least 2 dpa (displacement per atom). The status of Idaho National Laboratory (INL) efforts to develop the test rig in-situ creep testing capability for the ATR is described. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper reports efforts by INL to evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory (HTTL). Initial data from autoclave tests with 304 stainless steel (304 SS) specimens are reported.

  7. A thermodynamic approach for advanced fuels of gas-cooled reactors

    NASA Astrophysics Data System (ADS)

    Guéneau, C.; Chatain, S.; Gossé, S.; Rado, C.; Rapaud, O.; Lechelle, J.; Dumas, J. C.; Chatillon, C.

    2005-09-01

    For both high temperature reactor (HTR) and gas cooled fast reactor (GFR) systems, the high operating temperature in normal and accidental conditions necessitates the assessment of the thermodynamic data and associated phase diagrams for the complex system constituted of the fuel kernel, the inert materials and the fission products. A classical CALPHAD approach, coupling experiments and thermodynamic calculations, is proposed. Some examples of studies are presented leading with the CO and CO 2 gas formation during the chemical interaction of [UO 2± x/C] in the HTR particle, and the chemical compatibility of the couples [UN/SiC], [(U, Pu)N/SiC], [(U, Pu)N/TiN] for the GFR system. A project of constitution of a thermodynamic database for advanced fuels of gas-cooled reactors is proposed.

  8. Monitoring and Analysis of In-Pile Phenomena in Advanced Test Reactor using Acoustic Telemetry

    SciTech Connect

    Agarwal, Vivek; Smith, James A.; Jewell, James Keith

    2015-02-01

    The interior of a nuclear reactor presents a particularly harsh and challenging environment for both sensors and telemetry due to high temperatures and high fluxes of energetic and ionizing particles among the radioactive decay products. A number of research programs are developing acoustic-based sensing approach to take advantage of the acoustic transmission properties of reactor cores. Idaho National Laboratory has installed vibroacoustic receivers on and around the Advanced Test Reactor (ATR) containment vessel to take advantage of acoustically telemetered sensors such as thermoacoustic (TAC) transducers. The installation represents the first step in developing an acoustic telemetry infrastructure. This paper presents the theory of TAC, application of installed vibroacoustic receivers in monitoring the in-pile phenomena inside the ATR, and preliminary data processing results.

  9. Design and Status of RERTR Irradiation Tests in the Advanced Test Reactor

    SciTech Connect

    Daniel M. Wachs; Richard G. Ambrosek; Gray Chang; Mitchell K. Meyer

    2006-10-01

    Irradiation testing of U-Mo based fuels is the central component of the Reduced Enrichment for Research and Test Reactors (RERTR) program fuel qualification plan. Several RERTR tests have recently been completed or are planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory in Idaho Falls, ID. Four mini-plate experiments in various stages of completion are described in detail, including the irradiation test design, objectives, and irradiation conditions. Observations made during and after the in-reactor RERTR-7A experiment breach are summarized. The irradiation experiment design and planned irradiation conditions for full-size plate test are described. Progress toward element testing will be reviewed.

  10. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    SciTech Connect

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided.

  11. Thermodynamics of wetting by liquid metals

    NASA Astrophysics Data System (ADS)

    Yost, F. G.; Romig, A. D., Jr.

    The wetting of metal surfaces by molten solder is usually considered to be driven solely by an interfacial energy imbalance. The effect of chemical reactions on the wetting process is neglected, although the growth of an intermetallic layer in the wetted interface is commonly observed. In this work, the energy release during the incremental advance of a spreading solder droplet due to the interfacial energy imbalance and the formation of the intermetallic layer is calculated. The free energy of formation, (DELTA)G, of the intermetallic layer is shown to be an important driving force for solder wetting. This approach to setting has been applied to three systems, Cu-Sn, Cu-Sb and Cu-Cd. Liquid Sn, Sb and Cd react with solid Cu to form Cu6Sn5(eta), Cu2Sb(GAMMA) and CuCd3(epsilon), respectively.

  12. Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

    SciTech Connect

    O'Brien, James E.; Sabharwall, Piyush; Yoon, Su -Jong; Housley, Gregory K.

    2014-09-01

    This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

  13. 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.

  14. Radio-toxicity of spent fuel of the advanced heavy water reactor.

    PubMed

    Anand, S; Singh, K D S; Sharma, V K

    2010-01-01

    The Advanced Heavy Water Reactor (AHWR) is a new power reactor concept being developed at Bhabha Atomic Research Centre, Mumbai. The reactor retains many desirable features of the existing Pressurised Heavy Water Reactor (PHWR), while incorporating new, advanced safety features. The reactor aims to utilise the vast thorium resources available in India. The reactor core will use plutonium as the make-up fuel, while breeding (233)U in situ. On account of this unique combination of fuel materials, the operational characteristics of the fuel as determined by its radioactivity, decay heat and radio-toxicity are being viewed with great interest. Radio-toxicity of the spent fuel is a measure of potential radiological hazard to the members of the public and also important from the ecological point of view. The radio-toxicity of the AHWR fuel is extremely high to start with, being approximately 10(4) times that of the fresh natural U fuel used in a PHWR, and continues to remain relatively high during operation and subsequent cooling. A unique feature of this fuel is the peak observed in its radio-toxicity at approximately 10(5) y of decay cooling. The delayed increase in fuel toxicity has been traced primarily to a build-up of (229)Th, (230)Th and (226)Ra. This phenomenon has been observed earlier for thorium-based fuels and is confirmed for the AHWR fuel. This paper presents radio-toxicity data for AHWR spent fuel up to a period of 10(6) y and the results are compared with the radio-toxicity of PHWR.

  15. Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

    SciTech Connect

    Forsberg, C.W.

    1989-01-01

    Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' (Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety) is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document.

  16. Characteristics of the boat inductor for keeping liquid metal in the suspended state

    NASA Technical Reports Server (NTRS)

    Fogel, A. A.; Siforova, T. A.; Mezdrogina, M. M.

    1985-01-01

    Characteristics of the boat inductor for keeping liquid metal in the suspended state are examined. Behavioral features of the liquid metal, and the suspension boundary of liquid metal in the lower position are discussed. It is concluded that the inductor can be used to crystallize metals in the suspended state.

  17. Advanced High-Temperature, High-Pressure Transport Reactor Gasification

    SciTech Connect

    Michael L. Swanson

    2005-08-30

    The transport reactor development unit (TRDU) was modified to accommodate oxygen-blown operation in support of a Vision 21-type energy plex that could produce power, chemicals, and fuel. These modifications consisted of changing the loop seal design from a J-leg to an L-valve configuration, thereby increasing the mixing zone length and residence time. In addition, the standpipe, dipleg, and L-valve diameters were increased to reduce slugging caused by bubble formation in the lightly fluidized sections of the solid return legs. A seal pot was added to the bottom of the dipleg so that the level of solids in the standpipe could be operated independently of the dipleg return leg. A separate coal feed nozzle was added that could inject the coal upward into the outlet of the mixing zone, thereby precluding any chance of the fresh coal feed back-mixing into the oxidizing zone of the mixing zone; however, difficulties with this coal feed configuration led to a switch back to the original downward configuration. Instrumentation to measure and control the flow of oxygen and steam to the burner and mix zone ports was added to allow the TRDU to be operated under full oxygen-blown conditions. In total, ten test campaigns have been conducted under enriched-air or full oxygen-blown conditions. During these tests, 1515 hours of coal feed with 660 hours of air-blown gasification and 720 hours of enriched-air or oxygen-blown coal gasification were completed under this particular contract. During these tests, approximately 366 hours of operation with Wyodak, 123 hours with Navajo sub-bituminous coal, 143 hours with Illinois No. 6, 106 hours with SUFCo, 110 hours with Prater Creek, 48 hours with Calumet, and 134 hours with a Pittsburgh No. 8 bituminous coal were completed. In addition, 331 hours of operation on low-rank coals such as North Dakota lignite, Australian brown coal, and a 90:10 wt% mixture of lignite and wood waste were completed. Also included in these test campaigns was

  18. Numerical analysis of experiments with gas injection into liquid metal coolant

    NASA Astrophysics Data System (ADS)

    Usov, E. V.; Lobanov, P. D.; Pribaturin, N. A.; Mosunova, N. A.; Chuhno, V. I.; Kutlimetov, A. E.

    2016-10-01

    Presented paper contains results of a numerical analysis of experiments with gas injection in water and liquid metal which have been performed at the Institute of Thermophysics Russian Academy of Science (IT RAS). Obtained experimental data are very important to predict processes that take place in the BREST-type reactor during the hypothetical accident with damage of the steam generator tubes, and may be used as a benchmark to validate thermo-hydraulic codes. Detailed description of models to simulate transport of gas phase in a vertical liquid column is presented in a current paper. Two-fluid model with closing relation for wall friction and interface friction coefficients was used to simulate processes which take place in a liquid during injection of gaseous phase. It has being shown that proposed models allow obtaining a good agreement between experimental data and calculation results.

  19. Degradation of HT9 under simultaneous ion beam irradiation and liquid metal corrosion

    NASA Astrophysics Data System (ADS)

    Frazer, D.; Qvist, S.; Parker, S.; Krumwiede, D. L.; Caro, M.; Tesmer, J.; Maloy, S. A.; Wang, Y. Q.; Hosemann, P.

    2016-10-01

    A potentially promising coolant/structural material pair for a liquid-metal-cooled fast reactors is lead bismuth eutectic (LBE) coolant with the ferritic/martensitic steel HT9. The challenge of deploying LBE, however, is the corrosive environment it creates for structural materials. This corrosion can be mitigated with precise oxygen content control in the LBE to allow for the growth of passive protective oxide layers on the surface of the steel. In this paper, results are reported from the Irradiation Corrosion Experiment II (ICE-II), which allowed the simultaneous irradiation of a sample while in contact with LBE. It was found that a characteristic multilayer structure with an outer Fe3O4 oxide and inner FeCr2O4 spinel was grown and the oxidation was significantly larger in the irradiated region when compared to the region that was only exposed to LBE corrosion. Possible mechanisms are discussed to help understand this irradiation enhanced corrosion behavior.

  20. Roadmap for development of an advanced head-end reactor

    SciTech Connect

    Del Cul, G.D.; Johnson, J.A.; Spencer, B.B.; Collins, E.D.

    2013-07-01

    A novel dry treatment process for used nuclear fuel (UNF) using nitrogen dioxide is being developed to remove volatile and semi-volatile fission products and convert the monolithic fuel material to a fine powder suitable as a feed to many different separations processes. The process may be considered an advanced form of voloxidation, which was envisioned to remove tritium from the fuel prior to introduction of the fuel into the aqueous separations systems, where subsequent separation of tritium from the water would be difficult and expensive. The product from NO{sub 2} reaction can be selectively chosen to be U{sub 3}O{sub 8}, UO{sub 3}, or a nitrate by adjusting the processing conditions; all products are generated at temperatures lower than those used in standard voloxidation. All the fundamental tenants of the process have been successfully demonstrated as a proof of principle, and many aspects have been corroborated multiple times at laboratory scale. The goal of this roadmap is to define the activities required to develop the process to a technology-readiness level sufficient to an engineering-scale implementation. (authors)

  1. Application of the LBB regulatory approach to the steamlines of advanced WWER 1000 reactor

    SciTech Connect

    Kiselyov, V.A.; Sokov, L.M.

    1997-04-01

    The LBB regulatory approach adopted in Russia in 1993 as an extra safety barrier is described for advanced WWER 1000 reactor steamline. The application of LBB concept requires the following additional protections. First, the steamline should be a highly qualified piping, performed in accordance with the applicable regulations and guidelines, carefully screened to verify that it is not subjected to any disqualifying failure mechanism. Second, a deterministic fracture mechanics analysis and leak rate evaluation have been performed to demonstrate that postulated through-wall crack that yields 95 1/min at normal operation conditions is stable even under seismic loads. Finally, it has been verified that the leak detection systems are sufficiently reliable, diverse and sensitive, and that adequate margins exist to detect a through wall crack smaller than the critical size. The obtained results are encouraging and show the possibility of the application of the LBB case to the steamline of advanced WWER 1000 reactor.

  2. Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

    SciTech Connect

    Wood, RT

    2004-09-27

    This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

  3. After Action Report: Advanced Test Reactor Complex 2015 Evaluated Drill October 6, 2015

    SciTech Connect

    Holmes, Forest Howard

    2015-11-01

    The Advanced Test Reactor (ATR) Complex, operated by Battelle Energy Alliance, LLC, at the Idaho National Laboratory (INL) conducted an evaluated drill on October 6, 2015, to allow the ATR Complex emergency response organization (ERO) to demonstrate the ability to respond to and mitigate an emergency by implementing the requirements of DOE O 151.1C, “Comprehensive Emergency Management System.”

  4. Loss-of-coolant accident analyses of the Advanced Neutron Source Reactor

    SciTech Connect

    Chen, N.C.J.; Yoder, G.L. ); Wendel, M.W. )

    1991-01-01

    Currently in the conceptual design stage, the Advanced Neutron Source Reactor (ANSR) will operate at a high heat flux, a high mass flux, an a high degree of coolant subcooling. Loss-of-coolant accident (LOCA) analyses using RELAP5 have been performed as part of an early evaluation of ANSR safety issues. This paper discusses the RELAP5 ANSR conceptual design system model and preliminary LOCA simulation results. Some previous studies were conducted for the preconceptual design. 12 refs., 7 figs.

  5. Advanced reactors transition FY 1997 multi-year work plan WBS 7.3

    SciTech Connect

    Hulvey, R.K.

    1996-09-27

    This document describes in detail the work to be accomplised in FY 1997 and the out-years for the Advanced Reactors Transition (WBS 7.3) under the management of the Babcock & Wilcox Hanford Company. This document also includes specific milestones and funding profiles. Based upon the Fiscal Year 1997 Multi-Year Work Plan, the Department of Energy will provide authorization to perform the work described.

  6. Advanced Light Water Reactor Program: Program management and staff review methodology

    SciTech Connect

    Moran, D.H.

    1986-12-01

    This report summarizes the NRC/EPRI coordinated effort to develop design requirements for a standardized advanced light water reactor (ALWR) and the procedures for screening and applying new generic safety issues to this program. The end-product will be an NRC-approved ALWR Requirements Document for use by the nuclear industry in generating designs of LWRs to be constructed for operation in the 1990s and beyond.

  7. Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors

    SciTech Connect

    William Richins; Stephen Novascone; Cheryl O'Brien

    2009-08-01

    Summary of SMIRT20 Preconference Topical Workshop – Identifying Structural Issues in Advanced Reactors William Richins1, Stephen Novascone1, and Cheryl O’Brien1 1Idaho National Laboratory, US Dept. of Energy, Idaho Falls, Idaho, USA, e-mail: William.Richins@inl.gov The Idaho National Laboratory (INL, USA) and IASMiRT sponsored an international forum Nov 5-6, 2008 in Porvoo, Finland for nuclear industry, academic, and regulatory representatives to identify structural issues in current and future advanced reactor design, especially for extreme conditions and external threats. The purpose of this Topical Workshop was to articulate research, engineering, and regulatory Code development needs. The topics addressed by the Workshop were selected to address critical industry needs specific to advanced reactor structures that have long lead times and can be the subject of future SMiRT technical sessions. The topics were; 1) structural/materials needs for extreme conditions and external threats in contemporary (Gen. III) and future (Gen. IV and NGNP) advanced reactors and 2) calibrating simulation software and methods that address topic 1 The workshop discussions and research needs identified are presented. The Workshop successfully produced interactive discussion on the two topics resulting in a list of research and technology needs. It is recommended that IASMiRT communicate the results of the discussion to industry and researchers to encourage new ideas and projects. In addition, opportunities exist to retrieve research reports and information that currently exists, and encourage more international cooperation and collaboration. It is recommended that IASMiRT continue with an off-year workshop series on select topics.

  8. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    SciTech Connect

    Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  9. Compact power reactor

    DOEpatents

    Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

  10. Light Water Reactor Sustainability Program Advanced Seismic Soil Structure Modeling

    SciTech Connect

    Bolisetti, Chandrakanth; Coleman, Justin Leigh

    2015-06-01

    of interest. The specific nonlinear soil behavior included in the NLSSI calculation presented in this report is gapping and sliding. Other NLSSI effects are not included in the calculation. The results presented in this report document initial model runs in the linear and nonlinear analysis process. Final comparisons between traditional and advanced SPRA will be presented in the September 30th deliverable.

  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. Nucleation and the spall strength of liquid metals

    NASA Astrophysics Data System (ADS)

    Lopanitsyna, N. Yu; Kuksin, A. Yu

    2016-11-01

    This article presents calculation of the nucleation rate for liquid metals (Al, Fe, Mo) based on molecular dynamic simulation for embedded atom method (EAM) potentials. The dependence of nucleation rate on pressure and temperature could be approximated accurately in the form of classical nucleation theory taking into account surface tension dependency on pore radius σ = σ0/(1 + 2δ/r), where σ—surface tension, δ—the Tolman length. Basing on the results of the calculations, we have developed a model allowing calculating the spall strength of liquid metals under tension using such parameters as surface tension, viscosity, which could be measured experimentally. The obtained results for Mo and Al are consistent with experimental data and direct MD calculations at strain rates approx. 1010-1011 s-1.

  13. A Liquid Metal Flume for Free Surface Magnetohydrodynamic Experiments

    SciTech Connect

    Nornberg, M.D.; Ji, H.; Peterson, J.L.; Rhoads, J.R.

    2008-08-27

    We present an experiment designed to study magnetohydrodynamic effects in free-surface channel flow. The wide aspect ratio channel (the width to height ratio is about 15) is completely enclosed in an inert atmosphere to prevent oxidization of the liquid metal. A custom-designed pump reduces entrainment of oxygen, which was found to be a problem with standard centrifugal and gear pumps. Laser Doppler Velocimetry experiments characterize velocity profiles of the flow. Various flow constraints mitigate secondary circulation and end effects on the flow. Measurements of the wave propagation characteristics in the liquid metal demonstrate the surfactant effect of surface oxides and the damping of fluctuations by a cross-channel magnetic field.

  14. Topology-generating interfacial pattern formation during liquid metal dealloying

    PubMed Central

    Geslin, Pierre-Antoine; McCue, Ian; Gaskey, Bernard; Erlebacher, Jonah; Karma, Alain

    2015-01-01

    Liquid metal dealloying has emerged as a novel technique to produce topologically complex nanoporous and nanocomposite structures with ultra-high interfacial area and other unique properties relevant for diverse material applications. This process is empirically known to require the selective dissolution of one element of a multicomponent solid alloy into a liquid metal to obtain desirable structures. However, how structures form is not known. Here we demonstrate, using mesoscale phase-field modelling and experiments, that nano/microstructural pattern formation during dealloying results from the interplay of (i) interfacial spinodal decomposition, forming compositional domain structures enriched in the immiscible element, and (ii) diffusion-coupled growth of the enriched solid phase and the liquid phase into the alloy. We highlight how those two basic mechanisms interact to yield a rich variety of topologically disconnected and connected structures. Moreover, we deduce scaling laws governing microstructural length scales and dealloying kinetics. PMID:26582248

  15. Topology-generating interfacial pattern formation during liquid metal dealloying

    SciTech Connect

    Geslin, Pierre -Antoine; McCue, Ian; Gaskey, Bernard; Erlebacher, Jonah; Karma, Alain

    2015-11-19

    Liquid metal dealloying has emerged as a novel technique to produce topologically complex nanoporous and nanocomposite structures with ultra-high interfacial area and other unique properties relevant for diverse material applications. This process is empirically known to require the selective dissolution of one element of a multicomponent solid alloy into a liquid metal to obtain desirable structures. However, how structures form is not known. Here we demonstrate, using mesoscale phase-field modelling and experiments, that nano/microstructural pattern formation during dealloying results from the interplay of (i) interfacial spinodal decomposition, forming compositional domain structures enriched in the immiscible element, and (ii) diffusion-coupled growth of the enriched solid phase and the liquid phase into the alloy. We highlight how those two basic mechanisms interact to yield a rich variety of topologically disconnected and connected structures. Furthermore, we deduce scaling laws governing microstructural length scales and dealloying kinetics.

  16. Liquid Metal Propellant Feed System for Plasma Propulsion

    NASA Technical Reports Server (NTRS)

    Markusic, T. E.

    2004-01-01

    High-power plasma thrusters that utilize molten metallic propellants (e.g., the Lithium Lorentz Force Accelerator) are currently being investigated as a primary propulsion option for in-space nuclear-electric systems. A critical component of the thruster is the propellant feed system, which must reliably and accurately pump liquid metal into the thruster discharge chamber. We present design details and calibration results for a compact liquid metal propellant feed system that contains no moving parts, for use in laboratory testing of plasma thrusters. Feed line pressure is maintained using an MHD flow coupler, and the flow rate is monitored using a simple voltage divider, which is submerged in the propellant reservoir. Results for lithium and gallium propellants show capability to meter propellant at flow rates up to 10 +/- 0.1 mg/s.

  17. Topology-generating interfacial pattern formation during liquid metal dealloying

    DOE PAGES

    Geslin, Pierre -Antoine; McCue, Ian; Gaskey, Bernard; ...

    2015-11-19

    Liquid metal dealloying has emerged as a novel technique to produce topologically complex nanoporous and nanocomposite structures with ultra-high interfacial area and other unique properties relevant for diverse material applications. This process is empirically known to require the selective dissolution of one element of a multicomponent solid alloy into a liquid metal to obtain desirable structures. However, how structures form is not known. Here we demonstrate, using mesoscale phase-field modelling and experiments, that nano/microstructural pattern formation during dealloying results from the interplay of (i) interfacial spinodal decomposition, forming compositional domain structures enriched in the immiscible element, and (ii) diffusion-coupled growthmore » of the enriched solid phase and the liquid phase into the alloy. We highlight how those two basic mechanisms interact to yield a rich variety of topologically disconnected and connected structures. Furthermore, we deduce scaling laws governing microstructural length scales and dealloying kinetics.« less

  18. Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR)

    SciTech Connect

    A. Joseph Palmer; Gerry L. McCormick; Shannon J. Corrigan

    2010-06-01

    2010 International Congress on Advances in Nuclear Power Plants (ICAPP’10) ANS Annual Meeting Imbedded Topical San Diego, CA June 13 – 17, 2010 Hydraulic Shuttle Irradiation System (HSIS) Recently Installed in the Advanced Test Reactor (ATR) Author: A. Joseph Palmer, Mechanical Engineer, Irradiation Test Programs, 208-526-8700, Joe.Palmer@INL.gov Affiliation: Idaho National Laboratory P.O. Box 1625, MS-3840 Idaho Falls, ID 83415 INL/CON-10-17680 ABSTRACT Most test reactors are equipped with shuttle facilities (sometimes called rabbit tubes) whereby small capsules can be inserted into the reactor and retrieved during power operations. With the installation of Hydraulic Shuttle Irradiation System (HSIS) this capability has been restored to the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The general design and operating principles of this system were patterned after the hydraulic rabbit at Oak Ridge National Laboratory’s (ORNL) High Flux Isotope Reactor (HFIR), which has operated successfully for many years. Using primary coolant as the motive medium the HSIS system is designed to simultaneously transport fourteen shuttle capsules, each 16 mm OD x 57 mm long, to and from the B-7 position of the reactor. The B-7 position is one of the higher flux positions in the reactor with typical thermal and fast (>1 Mev) fluxes of 2.8E+14 n/cm2/sec and 1.9E+14 n/cm2/sec respectively. The available space inside each shuttle is approximately 14 mm diameter x 50 mm long. The shuttle containers are made from titanium which was selected for its low neutron activation properties and durability. Shuttles can be irradiated for time periods ranging from a few minutes to several months. The Send and Receive Station (SRS) for the HSIS is located 2.5 m deep in the ATR canal which allows irradiated shuttles to be easily moved from the SRS to a wet loaded cask, or transport pig. The HSIS system first irradiated (empty) shuttles in September 2009 and has since completed

  19. Radiogenic lead as coolant, reflector and moderator in advanced fast reactors

    NASA Astrophysics Data System (ADS)

    Kulikov, E. G.

    2017-01-01

    Main purpose of the study is assessing reasonability for recovery, production and application of radiogenic lead as a coolant, neutron moderator and neutron reflector in advanced fast reactors. When performing the study, thermal, physical and neutron-physical properties of natural and radiogenic lead were analyzed. The following results were obtained: 1. Radiogenic lead with high content of isotope 208Pb can be extracted from thorium or mixed thorium-uranium ores because 208Pb is a final product of 232Th natural decay chain. 2. The use of radiogenic lead with high 208Pb content in advanced fast reactors and accelerator-driven systems (ADS) makes it possible to improve significantly their neutron-physical and thermal-hydraulic parameters. 3. The use of radiogenic lead with high 208Pb content in advanced fast reactors as a coolant opens the possibilities for more intense fuel breeding and for application of well-known oxide fuel instead of the promising but not tested enough nitride fuel under the same safety parameters. 4. The use of radiogenic lead with high 208Pb content in ADS as a coolant can upgrade substantially the level of neutron flux in the ADS blanket, which enables effective transmutation of radioactive wastes with low cross-sections of radiative neutron capture.

  20. Design of a Gas Test Loop Facility for the Advanced Test Reactor

    SciTech Connect

    C. A. Wemple

    2005-09-01

    The Office of Nuclear Energy within the U.S. Department of Energy (DOE-NE) has identified the need for irradiation testing of nuclear fuels and materials, primarily in support of the Generation IV (Gen-IV) and Advanced Fuel Cycle Initiative (AFCI) programs. These fuel development programs require a unique environment to test and qualify potential reactor fuel forms. This environment should combine a high fast neutron flux with a hard neutron spectrum and high irradiation temperature. An effort is presently underway at the Idaho National Laboratory (INL) to modify a large flux trap in the Advanced Test Reactor (ATR) to accommodate such a test facility [1,2]. The Gas Test Loop (GTL) Project Conceptual Design was initiated to determine basic feasibility of designing, constructing, and installing in a host irradiation facility, an experimental vehicle that can replicate with reasonable fidelity the fast-flux test environment needed for fuels and materials irradiation testing for advanced reactor concepts. Such a capability will be needed if programs such as the AFCI, Gen-IV, the Next Generation Nuclear Plant (NGNP), and space nuclear propulsion are to meet development objectives and schedules. These programs are beginning some irradiations now, but many call for fast flux testing within this decade.

  1. Advanced Reactors-Intermediate Heat Exchanger (IHX) Coupling: Theoretical Modeling and Experimental Validation

    SciTech Connect

    Utgikar, Vivek; Sun, Xiaodong; Christensen, Richard; Sabharwall, Piyush

    2016-12-29

    The overall goal of the research project was to model the behavior of the advanced reactorintermediate heat exchange system and to develop advanced control techniques for off-normal conditions. The specific objectives defined for the project were: 1. To develop the steady-state thermal hydraulic design of the intermediate heat exchanger (IHX); 2. To develop mathematical models to describe the advanced nuclear reactor-IHX-chemical process/power generation coupling during normal and off-normal operations, and to simulate models using multiphysics software; 3. To develop control strategies using genetic algorithm or neural network techniques and couple these techniques with the multiphysics software; 4. To validate the models experimentally The project objectives were accomplished by defining and executing four different tasks corresponding to these specific objectives. The first task involved selection of IHX candidates and developing steady state designs for those. The second task involved modeling of the transient and offnormal operation of the reactor-IHX system. The subsequent task dealt with the development of control strategies and involved algorithm development and simulation. The last task involved experimental validation of the thermal hydraulic performances of the two prototype heat exchangers designed and fabricated for the project at steady state and transient conditions to simulate the coupling of the reactor- IHX-process plant system. The experimental work utilized the two test facilities at The Ohio State University (OSU) including one existing High-Temperature Helium Test Facility (HTHF) and the newly developed high-temperature molten salt facility.

  2. Development of Advanced High Uranium Density Fuels for Light Water Reactors

    SciTech Connect

    Blanchard, James; Butt, Darryl; Meyer, Mitchell; Xu, Peng

    2016-02-15

    This work conducts basic materials research (fabrication, radiation resistance, thermal conductivity, and corrosion response) on U3Si2 and UN, two high uranium density fuel forms that have a high potential for success as advanced light water reactor (LWR) fuels. The outcome of this proposed work will serve as the basis for the development of advance LWR fuels, and utilization of such fuel forms can lead to the optimization of the fuel performance related plant operating limits such as power density, power ramp rate and cycle length.

  3. Liquid Metal Thermal Electric Converter bench test module

    SciTech Connect

    Lukens, L.L.; Andraka, C.E.; Moreno, J.B.

    1988-04-01

    This report describes the design, fabrication, and test of a Liquid Metal Thermal Electric Converter Bench Test Module. The work presented in this document was conducted as a part of Heat Engine Task of the US Department of Energy's (DOE) Solar Thermal Technology Program. The objective of this task is the development and evaluation of heat engine technologies applicable to distributed receiver systems, in particular, dish electric systems.

  4. Lessons learned from the tokamak Advanced Reactor Innovation and Evaluation Study (ARIES)

    SciTech Connect

    Krakowski, R.A.; Bathke, C.G.; Miller, R.L.; Werley, K.A.

    1994-07-01

    Lessons from the four-year ARIES (Advanced Reactor Innovation and Evaluation Study) investigation of a number of commercial magnetic-fusion-energy (MFE) power-plant embodiments of the tokamak are summarized. These lessons apply to physics, engineering and technology, and environmental, safety, and health (ES&H) characteristics of projected tokamak power plants. Summarized herein are the composite conclusions and lessons developed in the course of four conceptual tokamak power-plant designs. A general conclusion from this extensive investigation of the commercial potential of tokamak power plants is the need for combined, symbiotic advances in both physics, engineering, and materials before economic competitiveness with developing advanced energy sources can be realized. Advances in materials are also needed for the exploitation of environmental advantages otherwise inherent in fusion power.

  5. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    SciTech Connect

    Qualls, A L; Cetiner, M S; Wilson, Jr, T L

    2012-04-30

    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

  6. Ultrasound Velocity Measurement in a Liquid Metal Electrode.

    PubMed

    Perez, Adalberto; Kelley, Douglas H

    2015-08-05

    A growing number of electrochemical technologies depend on fluid flow, and often that fluid is opaque. Measuring the flow of an opaque fluid is inherently more difficult than measuring the flow of a transparent fluid, since optical methods are not applicable. Ultrasound can be used to measure the velocity of an opaque fluid, not only at isolated points, but at hundreds or thousands of points arrayed along lines, with good temporal resolution. When applied to a liquid metal electrode, ultrasound velocimetry involves additional challenges: high temperature, chemical activity, and electrical conductivity. Here we describe the experimental apparatus and methods that overcome these challenges and allow the measurement of flow in a liquid metal electrode, as it conducts current, at operating temperature. Temperature is regulated within ±2 °C using a Proportional-Integral-Derivative (PID) controller that powers a custom-built furnace. Chemical activity is managed by choosing vessel materials carefully and enclosing the experimental setup in an argon-filled glovebox. Finally, unintended electrical paths are carefully prevented. An automated system logs control settings and experimental measurements, using hardware trigger signals to synchronize devices. This apparatus and these methods can produce measurements that are impossible with other techniques, and allow optimization and control of electrochemical technologies like liquid metal batteries.

  7. First-principles calculation of entropy for liquid metals.

    PubMed

    Desjarlais, Michael P

    2013-12-01

    We demonstrate the accurate calculation of entropies and free energies for a variety of liquid metals using an extension of the two-phase thermodynamic (2PT) model based on a decomposition of the velocity autocorrelation function into gas-like (hard sphere) and solid-like (harmonic) subsystems. The hard sphere model for the gas-like component is shown to give systematically high entropies for liquid metals as a direct result of the unphysical Lorentzian high-frequency tail. Using a memory function framework we derive a generally applicable velocity autocorrelation and frequency spectrum for the diffusive component which recovers the low-frequency (long-time) behavior of the hard sphere model while providing for realistic short-time coherence and high-frequency tails to the spectrum. This approach provides a significant increase in the accuracy of the calculated entropies for liquid metals and is compared to ambient pressure data for liquid sodium, aluminum, gallium, tin, and iron. The use of this method for the determination of melt boundaries is demonstrated with a calculation of the high-pressure bcc melt boundary for sodium. With the significantly improved accuracy available with the memory function treatment for softer interatomic potentials, the 2PT model for entropy calculations should find broader application in high energy density science, warm dense matter, planetary science, geophysics, and material science.

  8. Impregnated-electrode-type liquid metal ion source

    NASA Astrophysics Data System (ADS)

    Ishikawa, J.; Gotoh, Y.; Tsuji, H.; Takagi, T.

    We have developed an impregnated-electrode-type liquid metal ion source whose tip is a sintered-porous structure made of a refractory metal such as tungsten. By this structure the ratio of the liquid metal surface area facing the vacuum to the volume is low, which decreases useless metal evaporation from the surface. The maximum vapour pressure of the metal in operation for this ion source is 10 -1-10 0 Torr, which is 2-3 orders of magnitude higher than that for the needle type. Therefore, useful metal ions such as Ga +, Au +, Ag +, In +, Si 2+, Ge 2+, and Sb 2+ can be extracted from single element metals or alloys. The porous structure of the tip has also an effect on the positive control of the liquid metal flow rate to the tip head. Thus, a stable operation with a high current of a few hundreds of μA can be obtained together with a low current high brightness ion beam. Therefore, this ion source is suitable not only for microfocusing but also for a general use as a metal ion source.

  9. Development of magnetic liquid metal suspensions for magnetohydrodynamics

    NASA Astrophysics Data System (ADS)

    Carle, Florian; Bai, Kunlun; Casara, Joshua; Vanderlick, Kyle; Brown, Eric

    2017-01-01

    We demonstrate how to suspend various magnetic and nonmagnetic particles in liquid metals and characterize their properties relevant to magnetohydrodynamics (MHD). The suspending method uses an acid as a flux to eliminate oxidation from both metal particles and liquid, which allows the particles to be wetted and suspended into the liquid if the particles have higher conductivity than the liquid. With this process we were able to suspend a wide range of particle materials and sizes from 40 nm to 500 μ m into three different liquid metal bases and volume fractions ϕ up to the liquid-solid transition ϕc. By controlling the volume fraction of iron particles in liquid eGaIn, we increased the magnetic permeability by a factor of 5.0 and the electrical conductivity by 13% over that of the pure liquid metal, which gives these materials the potential to exhibit strong MHD effects on the laboratory scale that are usually only observable in the cores of planets and stars. By adding nonmagnetic zinc particles, we increased the viscosity by a factor of 160 while keeping the magnetic and electrical properties nearly constant, which would allow independent control of MHD effects from turbulence. We show that the suspensions flow like Newtonian fluids up to the volume fraction of the liquid-solid transition ϕc.

  10. 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.

  11. Advanced Mesh-Enabled Monte carlo capability for Multi-Physics Reactor Analysis

    SciTech Connect

    Wilson, Paul; Evans, Thomas; Tautges, Tim

    2012-12-24

    This project will accumulate high-precision fluxes throughout reactor geometry on a non- orthogonal grid of cells to support multi-physics coupling, in order to more accurately calculate parameters such as reactivity coefficients and to generate multi-group cross sections. This work will be based upon recent developments to incorporate advanced geometry and mesh capability in a modular Monte Carlo toolkit with computational science technology that is in use in related reactor simulation software development. Coupling this capability with production-scale Monte Carlo radiation transport codes can provide advanced and extensible test-beds for these developments. Continuous energy Monte Carlo methods are generally considered to be the most accurate computational tool for simulating radiation transport in complex geometries, particularly neutron transport in reactors. Nevertheless, there are several limitations for their use in reactor analysis. Most significantly, there is a trade-off between the fidelity of results in phase space, statistical accuracy, and the amount of computer time required for simulation. Consequently, to achieve an acceptable level of statistical convergence in high-fidelity results required for modern coupled multi-physics analysis, the required computer time makes Monte Carlo methods prohibitive for design iterations and detailed whole-core analysis. More subtly, the statistical uncertainty is typically not uniform throughout the domain, and the simulation quality is limited by the regions with the largest statistical uncertainty. In addition, the formulation of neutron scattering laws in continuous energy Monte Carlo methods makes it difficult to calculate adjoint neutron fluxes required to properly determine important reactivity parameters. Finally, most Monte Carlo codes available for reactor analysis have relied on orthogonal hexahedral grids for tallies that do not conform to the geometric boundaries and are thus generally not well

  12. PRELIMINARY DATA CALL REPORT ADVANCED BURNER REACTOR START UP FUEL FABRICATION FACILITY

    SciTech Connect

    S. T. Khericha

    2007-04-01

    The purpose of this report is to provide data for preparation of a NEPA Environmental Impact Statement in support the U. S. Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP). One of the GNEP objectives is to reduce the inventory of long lived actinide from the light water reactor (LWR) spent fuel. The LWR spent fuel contains Plutonium (Pu) -239 and other transuranics (TRU) such as Americium-241. One of the options is to transmute or burn these actinides in fast neutron spectra as well as generate the electricity. A sodium-cooled Advanced Recycling Reactor (ARR) concept has been proposed to achieve this goal. However, fuel with relatively high TRU content has not been used in the fast reactor. To demonstrate the utilization of TRU fuel in a fast reactor, an Advanced Burner Reactor (ABR) prototype of ARR is proposed, which would necessarily be started up using weapons grade (WG) Pu fuel. The WG Pu is distinguished by relatively highest proportions of Pu-239 and lesser amount of other actinides. The WG Pu will be used as the startup fuel along with TRU fuel in lead test assemblies. Because such fuel is not currently being produced in the US, a new facility (or new capability in an existing facility) is being considered for fabrication of WG Pu fuel for the ABR. This report is provided in response to ‘Data Call’ for the construction of startup fuel fabrication facility. It is anticipated that the facility will provide the startup fuel for 10-15 years and will take to 3 to 5 years to construct.

  13. Light Water Reactor Sustainability Program Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study

    SciTech Connect

    Curtis Smith; David Schwieder; Cherie Phelan; Anh Bui; Paul Bayless

    2012-08-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about LWR design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the RISMC Pathway R&D is to support plant decisions for risk-informed margins management with the aim to improve economics, reliability, and sustain safety of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. This report describes the RISMC methodology demonstration where the Advanced Test Reactor (ATR) was used as a test-bed for purposes of determining safety margins. As part of the demonstration, we describe how both the thermal-hydraulics and probabilistic safety calculations are integrated and used to quantify margin management strategies.

  14. Methods for quantifying uncertainty in fast reactor analyses.

    SciTech Connect

    Fanning, T. H.; Fischer, P. F.

    2008-04-07

    Liquid-metal-cooled fast reactors in the form of sodium-cooled fast reactors have been successfully built and tested in the U.S. and throughout the world. However, no fast reactor has operated in the U.S. for nearly fourteen years. More importantly, the U.S. has not constructed a fast reactor in nearly 30 years. In addition to reestablishing the necessary industrial infrastructure, the development, testing, and licensing of a new, advanced fast reactor concept will likely require a significant base technology program that will rely more heavily on modeling and simulation than has been done in the past. The ability to quantify uncertainty in modeling and simulations will be an important part of any experimental program and can provide added confidence that established design limits and safety margins are appropriate. In addition, there is an increasing demand from the nuclear industry for best-estimate analysis methods to provide confidence bounds along with their results. The ability to quantify uncertainty will be an important component of modeling that is used to support design, testing, and experimental programs. Three avenues of UQ investigation are proposed. Two relatively new approaches are described which can be directly coupled to simulation codes currently being developed under the Advanced Simulation and Modeling program within the Reactor Campaign. A third approach, based on robust Monte Carlo methods, can be used in conjunction with existing reactor analysis codes as a means of verification and validation of the more detailed approaches.

  15. Application of the "Full Cavitation Model" to the fundamental study of cavitation in liquid metal processing

    NASA Astrophysics Data System (ADS)

    Lebon, G. S. B.; Pericleous, K.; Tzanakis, I.; Eskin, D.

    2015-01-01

    Ultrasonic cavitation treatment of melt significantly improves the downstream properties and quality of conventional and advanced metallic materials. However, the transfer of this technology has been hindered by difficulties in treating large volumes of liquid metal. To improve the understanding of cavitation processing efficiency, the Full Cavitation Model, which is derived from a reduced form of the Rayleigh-Plesset equation, is modified and applied to the two-phase problem of bubble propagation in liquid melt. Numerical simulations of the sound propagation are performed in the microsecond time scale to predict the maximum and minimum acoustic pressure amplitude fields in the domain. This field is applied to the source term of the bubble transport equation to predict the generation and destruction of cavitation bubbles in a time scale relevant to the fluid flow. The use of baffles to limit flow speed in a launder conduit is studied numerically, to determine the optimum configuration that maximizes the residence time of the liquid in high cavitation activity regions. With this configuration, it is then possible to convert the batch processing of liquid metal into a continuous process. The numerical simulations will be validated against water and aluminium alloy experiments, carried out at Brunel University.

  16. First-principles entropy calculations for liquid metals and warm dense matter

    NASA Astrophysics Data System (ADS)

    Desjarlais, Michael

    2013-06-01

    The total entropy is not an explicit or easily accessible quantity in first-principles molecular dynamics simulations. It is, however, an extremely important quantity for the calculation of total free energies and derived properties such as equilibrium phase boundaries. In shock experiments the entropy of the shock state determines the release isentrope. Recent advances in the calculation of the entropy for liquid metals and warm dense matter directly from the velocity history in quantum molecular dynamics simulations are presented. The method, a generalization of the 2PT method for classical molecular dynamics, significantly increases the accuracy of the method for systems with electronic entropy, spin degrees of freedom, and the softer interactions characteristic of liquid metals and warm dense matter. The results are compared to data and the results of indirect methods, such as coexistence simulations to determine phase boundaries. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned subsidiary of Lockheed Martin Corporation, for the U.S. Department of Energy's National Nuclear Security Administration under contract DE-AC04-94AL85000.

  17. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

    SciTech Connect

    Ingersoll, D.T.

    2004-07-29

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

  18. 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.

  19. Safeguards and Non-proliferation Issues as Related to Advanced Fuel Cycle and Advanced Fast Reactor Development with Processing of Reactor Fuel

    SciTech Connect

    Rahmat Aryaeinejad; Jerry D. Cole; Mark W. Drigert; Dee E. Vaden

    2006-10-01

    The goal of this work is to establish basic data and techniques to enable safeguards appropriate to a new generation of nuclear power systems that will be based on fast spectrum reactors and mixed actinide fuels containing significant quantities of "minor" actinides, possibly due to reprocessing, and determination of what new radiation signatures and parameters need to be considered. The research effort focuses on several problems associated with the use of fuel having significantly different actinide inventories that current practice and on the development of innovative techniques using new radiation signatures and other parameters useful for safeguards and monitoring. In addition, the development of new distinctive radiation signatures as an aid in controlling proliferation of nuclear materials has parallel applications to support Gen-IV and current advanced fuel cycle initiative (AFCI) goals as well as the anticipated Global Nuclear Energy Partnership (GNEP).

  20. POWER REACTOR

    DOEpatents

    Zinn, W.H.

    1958-07-01

    A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.