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Sample records for prism liquid-metal reactor

  1. PRISM: An innovative liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Kruger, G.B.; Boardman, C.E.; Olich, E.E.; Switick, D.M.

    1986-01-01

    This paper describes an innovative sodium-cooled reactor concept employing small certified reactor modules coupled with a standardized steam generator system. The total plant employs nine PRISM reactors (power reactor inherently safe module) in three 415 MWe power blocks. The PRISM design concept utilizes inherent safety characteristics and modularity to improve licensability, reduce owner's risk, and reduce costs. The relatively small size of each reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. It is proposed that a single PRISM module be used in a full-scale integrated reactor safety test. Results from the test would be used to obtain NRC certification of the standard design

  2. Transient safety performance of the PRISM innovative liquid metal reactor

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Rhow, S.K.; Wu, T.

    1988-01-01

    The PRISM sodium-cooled reactor concept utilizes passive safety characteristics and modularity to increase performance margins, improve licensability, reduce owner's risk and reduce costs. The relatively small size of each reactor module (471 MWt) facilitates the use of passive self-shutdown and shutdown heat removal features, which permit design simplification and reduction of safety-related systems. Key to the transient performance is the inherent negative reactivity feedback characteristics of the core design resulting from the use of metal (U-Pu-Zr) swing, and very low control rod runout worth. Selected beyond design basis events relying only on these core design features are analyzed and the design margins summarized to demonstrate the advancement in reactor safety achieved with the PRISM design concept

  3. GE advanced liquid metal reactor S-PRISM

    International Nuclear Information System (INIS)

    Boardman, C.

    2001-01-01

    This presentation covers the following topics: Incentive for developing S-PRISM; Design and safety approach; Design description and competitive potential; Previous Licensing interactions; Planned approach to Licensing S-PRISM

  4. Progress on PRISM, an inherently safe, economic, and testable advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Tippets, F.E.; Salerno, L.N.; Boardman, C.E.; Kwant, W.; Murata, R.E.; Snyder, C.R.

    1987-01-01

    This paper reports progress on the design of PRISM (Power Reactor Inherently Safe Module) under the DOE-sponsored innovative reactor program now in its third year at General Electric. The purpose of this program is to develop a design for an inherently safe, reliable, and marketable liquid metal fast reactor power plant. The PRISM design approach includes the following key elements: Compact sodium-cooled pool-type reactor modules that are sized to enable factory fabrication, economical shipment to inland as well as water-side sites, and economical full-scale prototype testing for design certification; Nuclear safety-related envelope limited to the reactor modules and their service systems; Inherent, passive shutdown heat removal for loss-of-cooling events; Inherent, passive reactivity shutdown for failure-to-scram events

  5. Testing of seismic isolation bearings for advanced liquid metal reactor prism

    International Nuclear Information System (INIS)

    Tajirian, F.F.; Kelly, J.M.

    1988-01-01

    Seismic isolation can significantly mitigate earthquake loads on liquid metal reactors (LMR), thus reducing the impact of seismic loads on design. This improves plant safety margins for beyond-design basis seismic events and enhances adaptability of a standardized design to a variety of sites, with potential cost benefits. The PRISM (Power Reactor Inherently Safe Module) LMR incorporates a horizontal isolation system which consists of high damping steel laminated rubber bearings. The results of an experimental program to determine the mechanical properties of the rubber compound and the bearing performance under different loading conditions are presented. The test results demonstrate the excellent performance of the bearings and their suitability for isolating compact LMR plants

  6. Summary of advanced LMR [Liquid Metal Reactor] evaluations: PRISM [Power Reactor Inherently Safe Module] and SAFR [Sodium Advanced Fast Reactor

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  8. The PRISM concept for a safe, economic and testable liquid metal fast reactor plant

    International Nuclear Information System (INIS)

    Berglund, R.C.; Salerno, L.N.; Tippets, F.E.

    1987-01-01

    The PRISM project is underway at General Electric as part of an advanced reactor conceptual design program sponsored by the US Department of Energy. The PRISM concept emphasizes inherent safety, modular construction, and factory fabrication. These features are intended to improve the basis for public acceptance, reduce cost,improve licensability, and reduce the risk of schedule delays and cost increases during construction. A PRISM power plant comprises a number of reactor modules. The relatively small size of the reactor module facilitates the use of passive, inherent self-shutdown and shutdown heat removal features for safe accommodation of accidents. These inherent safety features permit simplification and reduction of conventional safety-related systems in the plant. Testing of a full-size prototype reactor module is planned in the late 1990's to demonstrate these inherent safety characteristics. It is intended that the results of the test be used to obtain certification of the design by the US Nuclear Regulatory Commission preparatory to use of reactor modules built to this standard design in licensed commercial plants

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

    Energy Technology Data Exchange (ETDEWEB)

    Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

    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.

  10. Liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Leigh, K.M.

    1980-01-01

    A liquid metal cooled nuclear reactor is described, wherein coolant is arranged to be flowed upwardly through a fuel assembly and having one or more baffles located above the coolant exit of the fuel assembly, the baffles being arranged so as to convert the upwardly directed motion of liquid metal coolant leaving the fuel assembly into a substantially horizontal motion. (author)

  11. Liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Scott, D.

    1981-01-01

    An improved method of constructing the diagrid used to support fuel assemblies of liquid metal fast breeder reactors, is described. The functions of fuel assembly support and coolant plenum are performed by discrete components of the diagrid each of which can serve the function of the other in the event of failure of one of the components. (U.K.)

  12. Liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Guidez, Joel; Jarriand, Paul.

    1975-01-01

    The invention concerns a fast neutron nuclear reactor cooled by a liquid metal driven through by a primary pump of the vertical drive shaft type fitted at its lower end with a blade wheel. To each pump is associated an exchanger, annular in shape, fitted with a central bore through which passes the vertical drive shaft of the pump, its wheel being mounted under the exchanger. A collector placed under the wheel comprises an open upward suction bell for the liquid metal. A hydrostatic bearing is located above the wheel to guide the drive shaft and a non detachable diffuser into which at least one delivery pipe gives, envelopes the wheel [fr

  13. Liquid-metal-cooled reactor

    International Nuclear Information System (INIS)

    Zhuchkov, I.I.; Filonov, V.S.; Zaitsev, B.I.; Artemiev, L.N.; Rakhimov, V.V.

    1976-01-01

    A liquid-metal-cooled reactor is described comprising two rotatable plugs, one of them, having at least one hole, being arranged internally of the other, a recharging mechanism with a guide tube adapted to be moved through the hole of the first plug by means of a drive, and a device for detecting stacks with leaky fuel elements, the recharging mechanism tube serving as a sampler

  14. Liquid metal reactor development

    International Nuclear Information System (INIS)

    Cho, Man; Kim, Yeong Cheol; Kim, Shi Hwan; Choi, Yeong Myeong; Sho, Dong Seop; Kim, Yeong In; Park, Joo Hwan; Kim, Yeong Kyoon; Song, Hoon; Kim, Yeong In; Cho, Chang Yeon; Cho, Seok Hong; Lee, Dong Jin; Kim, Jong Sook; Jeon, Hyeong Ryeon; Kim, Jeong Do; Kim, Deok In; Lee, Ui Jin; Kil, Chung Seop; Choi, Yeong Rok; Moon, Kap Seok; Yoo, Bong; Lee, Hyeong Yeon; Seo, Uk Hwan; Lee, Jae Han; Park, Yeon Pyo; Nam, Ho Yoon; Kim, Yong Ik; Min, Byeong Tae; Choi, Seok Ki; Kim, Yoo Kon; Lee, Yong Beom; Hwang, Jong Seon; An, Do Hui; Kang, Hui Seok; Choi, Byeong Hae; Kang, Yeong Hwan; Ryoo, Uh Seok; Joo, Ki Nam; Kim, Dae Hwan; Ji, Shee Hwan; Park, Deok Keun; Kim, Seong Soo; Maeng, Wan Yeong; Park, Shee Jin; Kim, Yeong Seok; Jang, Moon Hui; Hong, Joon Hwa; Han, Jeong Ho; No, Kyee Ho; Park, Ji Yeon; Jeong, Yong Hwan; Lee, Deok Hyeon; Jeong, Chung Hwan; Cho, Shee Hyeon; Kim, Dong Hwa; Seong, Ki Ung; Lee, Ki Yeong; Kim, Ui Kwang; Hong, Sang Hee

    1993-05-01

    On this year the study was performed in two parts : The establishment of LMR development plan, and the development of LMR coolant technology 1. The establishment of LMR technologies, the domestic political and technical environment, economics and technical maturity were duly considered for comparative analysis. In this year technologies specific to LMRs and technologies common to both PWRs and LMRs were identified to understand the inter-relationships between those two categorized technologies. Including those two categories, an overall LMR technology tree was drawn up taking into consideration technologies and tasks necessary to the pool type design of the primary and secondary cooling systems. And technology options that should be thoroughly evaluated their comparative feasibilities and applicabilities in trade-off study were derived as a preliminary procedure for the selection of the reactor type. 2. The development of LMR coolant technology. Many relevant basic technologies should be developed for LMR to have the inherent safety characteristics and to be economical. Since the sodium(Na) being used as the coolant in LMR has several thermo-hydraulic characteristics differing from water, the sodium handling technique which provides the maximum utilization of the thermo-hydraulic merits of the sodium and the protection measures against its defects is one of the most important technologies for the development of LMR. In the present study many problems associated with the establishment of the technology for measuring and controlling the impurity in the Na-facility have been investigated. The conceptual design of the purity control system in the Na-facility and related purity control system have been also made. The test-run of the Na-loop facility constructed last year has been performed, which provided the technology necessary for operation and repair of the Na-facility

  15. Liquid metal reactor absorber technology

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1990-10-01

    The selection of boron carbide as the reference liquid metal reactor absorber material is supported by results presented for irradiation performance, reactivity worth compatibility, and benign failure consequences. Scram response requirements are met easily with current control rod configurations. The trend in absorber design development is toward larger sized pins with fewer pins per bundle, providing economic savings and improved hydraulic characteristics. Very long-life absorber designs appear to be attainable with the application of vented pin and sodium-bonded concepts. 3 refs., 3 figs

  16. Evolution of the liquid metal reactor

    International Nuclear Information System (INIS)

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

    1989-01-01

    This paper reports on the integral fat reactor (IFR) concept. A key feature of the IFR concept is the metallic fuel, the original choice in liquid metal reactor development. An IFR development program is detailed by the authors

  17. Improvements in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1980-01-01

    A concrete containment vault for a liquid metal cooled nuclear reactor is described which is lined with thermal insulation to protect the vault against heat radiated from the reactor during normal operation of the reactor but whose efficiency of heat insulation is reduced in an emergency to enable excessive heat from the reactor to be dissipated through the vault. (UK)

  18. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    Scott, D.

    1979-01-01

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  19. Pool type liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Guthrie, B.M.

    1978-08-01

    Various technical aspects of the liquid metal fast breeder reactor (LMFBR), specifically pool type LMFBR's, are summarized. The information presented, for the most part, draws upon existing data. Special sections are devoted to design, technical feasibility (normal operating conditions), and safety (accident conditions). A survey of world fast reactors is presented in tabular form, as are two sets of reference reactor parameters based on available data from present and conceptual LMFBR's. (auth)

  20. Liquid metal fast reactor transient design

    International Nuclear Information System (INIS)

    Horak, C.; Purvis, E. III

    2000-01-01

    An examination has been made of how the currently available computing capabilities could be used to reduce Liquid Metal Fast Reactor design, manufacturing, and construction cost. While the examination focused on computer analyses some other promising means to reduce costs were also examined. (author)

  1. Development of Korea advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Park, C.K.

    1998-01-01

    Future nuclear power plants should not only have the features of improved safety and economic competitiveness but also provide a means to resolve spent fuel storage problems by minimizing volume of high level wastes. It is widely believed that liquid metal reactors (LMRs) have the highest potential of meeting these requirements. In this context, the LMR development program was launched as a national long-term R and D program in 1992, with a target to introduce a commercial LMR around 2030. Korea Advanced Liquid Metal Reactor (KALIMER), a 150 MWe pool-type sodium cooled prototype reactor, is currently under the conceptual design study with the target schedule to complete its construction by the mid-2010s. This paper summarizes the KALIMER development program and major technical features of the reactor system. (author)

  2. Improvements in liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1980-01-01

    Improvements in the design of the thermally insulating material used to shield the concrete containment walls in liquid metal cooled nuclear reactors are described in detail. The insulating material is composed of two layers and is placed between the primary vessel (usually steel) and the steel lined concrete containment vault. The outer layer, which clads the inner wall surface of the vault, is generally impervious to liquid metal coolant whilst the inner layer is pervious to the coolant. In normal operation, both layers protect the concrete from heat radiated from the reactor. In the event of a breach of the containment vessel, the resulting leakage of liquid metal coolant permeates the inner layer of insulating material, provides a means of heat transfer by conduction and hence reduces the overall insulating properties of the two layers. The outer layer continues to protect the wall surface of the vault from substantial direct contact with the liquid metal. Thus the two apparently conflicting requirements of good thermal insulation during normal operation and of heat transfer during loss of coolant accidents are satisfied by this novel design. Suggestions are given for possible materials for use as the insulating layers. (U.K.)

  3. Developing remote techniques for liquid metal reactors

    International Nuclear Information System (INIS)

    Fenemore, Peter

    1987-01-01

    Three devices have been designed in Britain to meet the need for special remote equipment and techniques required to inspect the reactor vessel and internals of liquid metal reactors. The ''Links Manipulator Under-Sodium Viewing System'' - a device to be used for the surveillance of reactor internals, which are submerged in sodium. An ''Automatic Guided Vehicle'' - a free roving vehicle to be used to survey the externals of the reactor vessel. The ''Snake Manipulator'' - an articulated arm used to gain access to restricted areas. (author)

  4. A low cost liquid metal reactor design

    International Nuclear Information System (INIS)

    Arnold, W.H.; Anderson, C.A.; Mangus, J.D.

    1984-01-01

    A new, compact Liquid Metal Reactor (LMR) plant arrangement designed by Westinghouse, featuring factory-fabricated modules and an integrated fuel cycle facility, has made it possible to project a commercially competitive LMR plant for the near future. This innovative liquid metal-cooled plant design will allow a combination of capital, fuel, operation and maintenance costs that could be lower than today's fossil-fueled or light water reactor plant costs, and incorporate features which enhance public safety even beyond current high standards. Following early core loadings, the plant feeds only on depleted uranium. No shipment of fuel is required. And the plant can be tailored to produce enough plutonium to meet its need or to provide fuel for other nuclear plants

  5. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Gatley, J.A.

    1979-01-01

    Breeder fuel sub-assemblies with electromagnetic brakes and fluidic valves for liquid metal cooled fast breeder reactors are described. The electromagnetic brakes are of relatively small proportions and the valves are of the controlled vortex type. The outlet coolant temperature of at least some of the breeder sub-assemblies are maintained by these means substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (UK)

  6. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Thatcher, G.; Mitchell, A.J.

    1981-01-01

    Fuel sub-assemblies for liquid metal-cooled fast breeder reactors are described which each incorporate a fluid flow control valve for regulating the rate of flow through the sub-assembly. These small electro-magnetic valves seek to maintain the outlet coolant temperature of at least some of the breeder sub-assemblies substantially constant throughout the life of the fuel assembly without severely pressurising the sub-assembly. (U.K.)

  7. Advanced liquid metal reactor plant control system

    International Nuclear Information System (INIS)

    Dayal, Y.; Wagner, W.; Zizzo, D.; Carroll, D.

    1993-01-01

    The modular Advanced Liquid Metal Reactor (ALMR) power plant is controlled by an advanced state-of-the-art control system designed to facilitate plant operation, optimize availability, and protect plant investment. The control system features a high degree of automatic control and extensive amount of on-line diagnostics and operator aids. It can be built with today's control technology, and has the flexibility of adding new features that benefit plant operation and reduce O ampersand M costs as the technology matures

  8. The US Liquid Metal Reactor Development Program

    International Nuclear Information System (INIS)

    Till, C.E.; Arnold, W.H.; Griffith, J.D.

    1988-01-01

    The US Liquid Metal Reactor Development Program has been restructured to take advantage of the opportunity today to carry out R and D on truly advanced reactor technology. The program gives particular emphasis to improvements to reactor safety. The new directions are based on the technology of the Integral Fast Reactor (IFR). Much of the basis for superior safety performance using IFR technology has been experimentally verified and aggressive programs continue in EBR-II and TREAT. Progress has been made in demonstrating both the metallic fuel and the new electrochemical processes of the IFR. The FFTF facility is converting to metallic fuel; however, FFTF also maintains a considerable US program in oxide fuels. In addition, generic programs are continuing in steam generator testing, materials development, and, with international cooperation, aqueous reprocessing. Design studies are carried out in conjunction with the IFR technology development program. In summary, the US maintains an active development program in Liquid Metal Reactor technology, and new directions in reactor safety are central to the program

  9. Radionuclide trap for liquid metal cooled reactors

    International Nuclear Information System (INIS)

    McGuire, J.C.; Brehm, W.F.

    1978-10-01

    At liquid metal cooled reactor operating temperatures, radioactive corrosion product transport and deposition in the primary system will be sufficiently high to limit access time for maintenance of system components. A radionuclide trap has been developed to aid in controlling radioactivity transport. This is a device which is located above the reactor core and which acts as a getter, physically immobilizing radioactive corrosion products, particularly 54 Mn. Nickel is the getter material used. It is most effective at temperatures above 450 0 C and effectiveness increases with increasing temperature. Prototype traps have been tested in sodium loops for 40,000 hours at reactor primary temperatures and sodium velocities. Several possible in-reactor trap sites were considered but a location within the top of each driver assembly was chosen as the most convenient and effective. In this position the trap is changed each time fuel is changed

  10. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Durston, J.G.

    1976-01-01

    It is stated that in a liquid metal cooled fast breeder reactor wherein the core, intermediate heat exchangers and liquid metal pumps are immersed in a pool of coolant such as Na, the intermediate heat exchangers are suspended from the roof, and ducting is provided in the form of a core tank or shroud interconnected with 'pods' housing the intermediate exchangers for directing coolant from the core over the heat exchanger tubes and thence back to the main pool of liquid metal. Seals are provided between the intermediate heat exchanger shells and the walls of their 'pods' to prevent liquid metal flow by-passing the heat exchanger tube bundles. As the heat exchangers must be withdrawable for servicing, and because linear differential thermal expansion of the heat exchanger and its 'pod' must be accommodated the seals hitherto have been of the sliding kind, generally known as 'piston ring type seals'. These present several disadvantages; for example sealing is not absolute, and the metal to metal seal gives rise to wear and fretting by rubbing and vibration. This could lead to seizure or jamming by the deposition of impurities in the coolant. Another difficulty arises in the need to accommodate lateral thermal expansion of the ducting, including the core tank and 'pods'. Hitherto some expansion has been allowed for by the use of expansible bellow pairs in the interconnections, or alternatively by allowing local deformations of the core tank 'pods'. Such bellows must be very flexible and hence constitute a weak section of the ducting, and local deformations give rise to high stress levels that could lead to premature failure. The arrangement described seeks to overcome these difficulties by use of a gas pocket trapping means to effect a seal against vertical liquid flow between the heat exchanger shell and the wall of the heat exchanger housing. Full details of the arrangement are described. (U.K.)

  11. Liquid metal cooled nuclear reactor constructions

    International Nuclear Information System (INIS)

    Aspden, G.J.; Allbeson, K.F.

    1984-01-01

    In a liquid metal cooled nuclear reactor with a nuclear fuel assembly in a coolant-containing primary vessel housed within a concrete containment vault, there is thermal insulation to protect the concrete, the insulation being disposed between vessel and concrete and being hung from metal structure secured to and projecting from the concrete, the insulation consisting of a plurality of adjoining units each unit incorporating a pack of thermal insulating material and defining a contained void co-extensive with said pack and situated between pack and concrete, the void of each unit being connected to the voids of adjoining units so as to form continuous ducting for a fluid coolant. (author)

  12. Liquid metal pump for nuclear reactors

    International Nuclear Information System (INIS)

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

    1975-01-01

    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

  13. Liquid metal reactor core material HT9

    International Nuclear Information System (INIS)

    Kim, S. H.; Kuk, I. H.; Ryu, W. S. and others

    1998-03-01

    A state-of-the art is surveyed on the liquid metal reactor core materials HT9. The purpose of this report is to give an insight for choosing and developing the materials to be applied to the KAERI prototype liquid metal reactor which is planned for the year of 2010. In-core stability of cladding materials is important to the extension of fuel burnup. Austenitic stainless steel (AISI 316) has been used as core material in the early LMR due to the good mechanical properties at high temperatures, but it has been found to show a poor swelling resistance. So many efforts have been made to solve this problem that HT9 have been developed. HT9 is 12Cr-1MoVW steel. The microstructure of HT9 consisted of tempered martensite with dispersed carbide. HT9 has superior irradiation swelling resistance as other BCC metals, and good sodium compatibility. HT9 has also a good irradiation creep properties below 500 dg C, but irradiation creep properties are degraded above 500 dg C. Researches are currently in progress to modify the HT9 in order to improve the irradiation creep properties above 500 dg C. New design studies for decreasing the core temperature below 500 dg C are needed to use HT9 as a core material. On the contrary, decrease of the thermal efficiency may occur due to lower-down of the operation temperature. (author). 51 refs., 6 tabs., 19 figs

  14. Liquid metal tribology in fast breeder reactors

    International Nuclear Information System (INIS)

    Wild, E.; Mack, K.J.; Gegenheimer, M.

    1984-11-01

    Liquid Metal Cooled Fast Breeder Reactors (LMFBR) require mechanisms operating in various sodium liquid and sodium vapor environments for extended periods of time up to temperatures of 900 K under different chemical properties of the fluid. The design of tribological systems in those reactors cannot be based on data and past experience of so-called conventional tribology. Although basic tribological phenomena and their scientific interpretation apply in this field, operating conditions specific to nuclear reactors and prevailing especially in the nuclear part of such facilities pose special problems. Therefore, in the framework of the R and D-program accompanying the construction phase of SNR 300 experiments were carried out to provide data and knowledge necessary for the lay-out of friction systems between mating surfaces of contacting components. Initially, screening tests isolated material pairs with good slipping properties and maximum wear resistance. Those materials were subjected to comprehensive parameter investigations. A multitude of laboratory scale tests have been performed under largely reactor specific conditions. Unusual superimpositions of parameters were analyzed and separated to find their individual influence on the friction process. The results of these experiments were made available to the reactor industry as well as to factories producing special tribo-materials. (orig.) [de

  15. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Duncombe, E.; Thatcher, G.

    1979-01-01

    The invention described relates to a liquid metal cooled fast breeder nuclear reactor in which the fuel assembly has an inner zone comprised mainly of fissile material and a surrounding outer zone comprised mainly of breeder material. According to the invention the sub-assemblies in the outer zone include electro-magnetic braking devices (magnets, pole pieces and armature) for regulating the flow of coolant through the sub-assemblies. The magnetic fields of the electro-magnetic breaking devices are temperature sensitive so that as the power output of the breeder sub-assemblies increases the electro-magnetic resistance to coolant flow is reduced thereby maintaining the temperature of the coolant outlets from the sub-assemblies substantially constant. (UK)

  16. Evaluating advanced LMR [liquid metal reactor] reactivity feedbacks using SSC

    International Nuclear Information System (INIS)

    Slovik, G.C.; Van Tuyle, G.J.; Kennett, R.J.; Cheng, H.S.

    1988-01-01

    Analyses of the PRISM and SAFR Liquid Metal Reactors with SSC are discussed from a safety and licensing perspective. The PRISM and SAFR reactors with metal fuel are designed for inherent shutdown responses to loss-of-flow and loss-of-heat-sink events. The demonstration of this technology was performed by EBR-II during experiments in April 1986 by ANL (Planchon, et al.). Response to postulated TOPs (control rod withdrawal) are made acceptable largely by reducing reactivity swings, and therefore minimizing the size of possible ractivity insertions. Analyses by DOE and the contractors GE, RI, and ANL take credit for several reactivity feedback mechanisms during transient calculations. These feedbacks include Doppler, sodium density, and thermal expansion of the grid plates, the load pads, the fuel (axial) and the control rod which are now factored into the BNL SSC analyses. The bowing feedback mechanism is not presently modeled in the SSC due to its complexity and subsequent large uncertainty. The analysis is conservative by not taking credit for this negative feedback mechanism. Comparisons of BNL predictions with DOE contractors are provided

  17. Inducer pumps for liquid metal reactor plants

    International Nuclear Information System (INIS)

    Jackson, E.D.

    2002-01-01

    Pumps proposed for liquid metal reactor plants typically use centrifugal impellers as the rotating element and are required to maintain a relatively low speed to keep the suction specific speed low enough to operate at the available net positive suction head (HPSH) and to avoid cavitation damage. These low speeds of operation require that the pump diameter increase and/or multiple stages be used to achieve the design head. This frequently results in a large, heavy, complex pump design. In addition, the low speed results in a larger drive motor size so that the resultant penalty to the plant designer is multiplied. The heavier pump can also result in further complications as, e.g., the difficulty in maintaining the first critical speed sufficiently above the pump operating range to provide margin for rotor dynamic stability. To overcome some of these disadvantages, it was proposed the use of inducer pumps for Liquid Metal Fast Breeder Reactor (LMFBR) plants. This paper discusses some of the advantages of the inducer pump and the development history of designing and testing these pumps both in water and sodium. The inducer pump is seen to be a sound concept with a strong technology base derived from the aerospace and ship propulsion industries. The superior suction performance capability of the inducer offers significant system design advantages, primarily a smaller, lighter weight, less complex pump design with resulting saving in cost. Extensive testing of these pumps has been conducted in both sodium and water to demonstrate the long-life capability with no cavitation damage occurring in those designs based on Rockwell's current design criteria. These tests have utilized multiple inspection and measurement approaches to accurately assess and identify any potential for cavitation damage, and these approaches have all concluded that no damage is occurring. Therefore, it is concluded that inducer pumps can be safely designed for long life operation in sodium with

  18. Liquid metal cooled fast breeder nuclear reactors

    International Nuclear Information System (INIS)

    Barnes, S.

    1976-01-01

    Reference is made to liquid metal cooled fast breeder reactors of the 'pool' kind. In this type of reactor the irradiated fuel is lowered into a transfer rotor for removal to storage facilities, this rotor normally having provision for the temporary storage of 20 irradiated fuel assemblies, each within a stainless steel bucket. For insertion or withdrawal of a fuel assembly the rotor is rotated to bring the fuel assembly to a loading or discharging station. The irradiated fuel assembly is withdrawn from the rotor within its bucket and the total weight is approximately 1000 kg, which is lifted about 27 m. In the event of malfunction the combination falls back into the rotor with considerable force. In order to prevent damage to the rotor fracture pins are provided, and to prevent damage to the reactor vessel and other parts of the reactor structure deformable energy absorbing devices are provided. After a malfunction the fractured pins and the energy absorbing devices must be replaced by remote control means operated from outside the reactor vault - a complex operation. The object of the arrangement described is to provide improved energy absorbing means for fuel assemblies falling into a fuel transfer rotor. The fuel assemblies are supported in the rotor by elastic means during transfer to storage and a hydraulic dash pot is provided in at least one position below the rotor for absorbing the energy of a falling fuel assembly. It is preferable to provide dash pots immediately below a receiving station for irradiated fuel assemblies and immediately below a discharge station. Each bucket is carried in a container that is elastically supported in the transfer rotor on a helical coil compression spring, so that, in the event of a malfunction the container and bucket are returned to their normal operating position after the force of the falling load has been absorbed by the dash pot. The transfer rotor may also be provided with recoil springs to absorb the recoil energy

  19. Advancing liquid metal reactor technology with nitride fuels

    International Nuclear Information System (INIS)

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

    1991-08-01

    A review of the use of nitride fuels in liquid metal fast reactors is presented. Past studies indicate that both uranium nitride and uranium/plutonium nitride possess characteristics that may offer enhanced performance, particularly in the area of passive safety. To further quantify these effects, the analysis of a mixed-nitride fuel system utilizing the geometry and power level of the US Advanced Liquid Metal Reactor as a reference is described. 18 refs., 2 figs., 2 tabs

  20. Characterization of liquid metal reactor materials

    International Nuclear Information System (INIS)

    Kuk, I. H.; Ryu, W. S.; Kim, H. H. and others

    1999-03-01

    The objectives of this report were to assess the material requirements for LMR environment, to select the optimum candidates for KALIMER components, to characterize the performance for establishing a database of the structural materials for KALIMER, and to develop the basic material technologies for the localization of the advanced materials. Stainless steel ingots were melted by VIM and hot-rolled to plate with the thickness of 15mm. The plate was solution-treated for 1 hr at 1100 deg C and then water-quenched. Specimens were taken parallel to the rolling direction of the plate. The effects of nitrogen and phosphorus were analyzed on the high temperature mechanical properties of 316MRP (Liquid Metal Reactor, Primary candidate material) stainless steels with the different nitrogen content from 0.04 to 0.15% and with the different phosphorus content from 0.002 to 0.02%. Heat treatment was performed to investigate the changes in microstructure and mechanical properties of Cr-Mo steels for LMR heat transfer tube materials and core materials. The Cr-Mo steels were normalized at the temperatures between 900 deg C and 1200 deg C for 1hrs and tempered at the temperatures between 500 deg C and 800 deg C for 2hrs. Conventional optical and electron micrographic studies were carried out to investigate the martensite lath structure, carbide indentification and carbide shape. Vickers microhardness was measured at room temperature using 10g load. Tensile properties were tested at high temperature. Charpy V-notch impact tests were also carried out at temperature between -120 deg C and +180 deg C. (author). 72 refs., 28 tabs., 244 figs

  1. A large economic liquid metal reactor for United States utilities

    International Nuclear Information System (INIS)

    Rodwell, E.

    1985-01-01

    The United States has demonstrated its ability to build and operate small and medium sized liquid metal reactors and continues to operate the Experimental Breeder Reactor II and the Fast Flux Test Facility to demonstrate long life fuel designs. Similar-sized liquid metal reactors in Europe have been followed by a step-up to the 1200 MWe capacity of the Superphenix plant. To permit the United States to make a similar step-up in capacity, a 1320 MWe liquid metal reactor plant has been designed with the main emphasis on minimizing the specific capital cost in order to be competitive with light water reactor plant and fossil plant alternatives. The design is based on a four parallel heat transport loops arrangement and complies with current regulatory requirements. The primary heat transport loops are now being integrated into the reactor vessel to achieve further reduction in the capital cost

  2. Blanket of a hybrid thermonuclear reactor with liquid- metal cooling

    International Nuclear Information System (INIS)

    Terent'ev, I.K.; Fedorovich, E.P.; Paramonov, P.M.; Zhokhov, K.A.

    1982-01-01

    Blanket design of a hybrid thermopuclear reactor with a liquid metal coolant is described. To decrease MHD-resistance for uranium zone fuel elements a cylindrical shape is suggested and movement of liquid-metal coolant in fuel element packets is presumed to be in perpendicular to the magnetic field and fuel element axes direction. The first wall is cooled by water, blanket-by lithium-lead alloy

  3. Status of development and licensing support for advanced liquid metal reactors in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, D R [Argonne National Laboratory, Argonne, IL (United States); Gyorey, G [General Electric, San Jose, CA (United States)

    1991-07-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR 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 U.S. program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment. (author)

  4. Status of development and licensing support for advanced liquid metal reactors in the United States

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Gyorey, G.

    1991-01-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR 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. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment

  5. Safety aspects of the US advanced LMR [liquid metal reactor] design

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Gyorey, G.L.; Marchaterre, J.F.; Rosen, S.

    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

  6. Status of development and licensing support for advanced liquid metal reactors in the United States

    International Nuclear Information System (INIS)

    Pedersen, D.R.; Gyorey, G.

    1991-01-01

    The cornerstones of the United States Advanced Liquid Metal Cooled Reactor (ALMR) program sponsored by the Department of Energy are: the ALMR 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 U.S. program is to produce a standard, commercial ALMR, including the associated fuel cycle. The paper addresses the status of the IFR program, the ALMR program and the interaction of the ALMR program with the regulatory environment. (author)

  7. Status of liquid metal reactor development in the United States of America

    International Nuclear Information System (INIS)

    Griffith, J.D.; Horton, K.E.

    1990-01-01

    The United States have made substantial progress in achieving Advanced Liquid Metal Reactor (ALMR) program objectives. A decision was made in 1988 to select the General Electric ALMR concept known as PRISM (Power Reactor Innovative Safe Module) for advanced conceptual design. A 3-year contract was awarded to General Electric in January of last year for concentrated trade-off studies and advanced design development. The strategy is to integrate those advancements that best meet program objectives into a national ALMR system concept. (author). 10 figs, 1 tab

  8. Performance and safety design of the advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Berglund, R.C.; Magee, P.M.; Boardman, C.E.; Gyorey, G.L.

    1991-01-01

    The Advanced Liquid Metal Reactor (ALMR) program led by General Electric is developing, under U.S. Department of Energy sponsorship, a conceptual design for an advanced sodium-cooled liquid metal reactor plant. This design is intended to improve the already excellent level of plant safety achieved by the nuclear power industry while at the same time providing significant reductions in plant construction and operating costs. In this paper, the plant design and performance are reviewed, with emphasis on the ALMR's unique passive design safety features and its capability to utilize as fuel the actinides in LWR spent fuel

  9. Emergency cooling system for a liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Murata, Ryoichi; Fujiwara, Toshikatsu.

    1980-01-01

    Purpose: To suitably cool liquid metal as coolant in emergency in a liquid metal cooled reactor by providing a detector for the pressure loss of the liquid metal passing through a cooling device in a loop in which the liquid metal is flowed and communicating the detector with a coolant flow regulator. Constitution: A nuclear reactor is stopped in nuclear reaction by control element or the like in emergency. If decay heat is continuously generated for a while and secondary coolant is insufficiently cooled with water or steam flowed through a steam and water loop, a cooler is started. That is, low temperature air is supplied by a blower through an inlet damper to the cooler to cool the secondary coolant flowed into the cooler through a bypass pipe so as to finally safely stop an entire plant. Since the liquid metal is altered in its physical properties by the temperature at this time, it is detected to regulate the opening of the valve of the damper according to the detected value. (Sekiya, K.)

  10. Fuel transfer manipulator for liquid metal nuclear reactors

    International Nuclear Information System (INIS)

    Sturges, R.H.

    1983-01-01

    A manipulator for transferring fuel assemblies between inclined fuel chutes of a liquid metal nuclear reactor installation. Hoisting means are mounted on a mount supported by beams pivotably attached by pins to the mount and to the floor in such a manner that pivoting of the beams causes movement and tilting of a hoist tube between positions of alignment with the inclined chutes. (author)

  11. Application of a bistable convection loop to LMFBR [liquid metal fast breeder reactor] emergency core cooling

    International Nuclear Information System (INIS)

    Anand, G.; Christensen, R.N.

    1990-01-01

    The concept of passive safety features for nuclear reactors has been developed in recent years and has gained wide acceptance. A literature survey of current reactors with passive features indicates that these reactors have some passive features but still do not fully meet the design objectives. Consider a current liquid-metal reactor design like PRISM. During normal operation, liquid sodium enters the reactor at ∼395 degree C and exits at ∼550 degree C. In the event of loss of secondary cooling with or without scram, the primary coolant (liquid sodium) initially acts as a heat sink and its temperature increases. For events without scram, the negative reactivity induced by the increase in temperature shuts the reactor down. When the average temperature of the sodium reaches ∼600 to 650 degree C, it overflows from the reactor vessel, activating the auxiliary cooling system. The auxiliary cooling system uses natural circulation of air around the reactor guard vessel. An alternative to the current design incorporates a bistable convection loop (BCL). The incorporation of the BCL concept remarkably improves the safety of the nuclear reactors. Application of the BCL concept to liquid-metal fast breeder reactors is described in this paper

  12. Simplified analysis of trasients in pool type liquid metal reactors

    International Nuclear Information System (INIS)

    Botelho, D.A.

    1987-01-01

    The conceptual design of a liquid metal fast breeder reactor will require a great effort of development in several technical disciplines. One of them is the thermal-hydraulic design of the reactor and of the heat and fluid transport components inside the reactor vessel. A simplified model to calculate the maximum sodium temperatures is presented in this paper. This model can be used to optimize the layout of components inside the reactor vessel and was easily programmed in a small computer. Illustrative calculations of two transients of a typical hot pool type fast reactor are presented and compared with the results of other researchers. (author) [pt

  13. Advanced liquid metal fast breeder reactor designs

    International Nuclear Information System (INIS)

    Sayles, C.W.

    1978-01-01

    Fast Breeder reactor power plants in the 1000-1200 MW(e) range are being built overseas and are being designed in this country. While these reactors have many characteristics in common, a variety of different approaches have been adopted for some of the major features. Some of those alternatives are discussed

  14. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    International Nuclear Information System (INIS)

    Park, Jee Won; Jeong, C. J.; Yang, M. S.

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs

  15. Current liquid metal cooled fast reactor concepts: use of the dry reprocess fuel

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jee Won; Jeong, C. J.; Yang, M. S

    2003-03-01

    Recent Liquid metal cooled Fast Reactor (LFR) concepts are reviewed for investigating the potential usability of the Dry Reprocess Fuel (DRF). The LFRs have been categorized into two different types: the sodium cooled and the lead cooled systems. In each category, overall design and engineering concepts are collected which includes those of S-PRISM, AFR300, STAR, ENHS and more. Specially, the nuclear fuel types which can be used in these LFRs, have been summarized and their thermal, physical and neutronic characteristics are tabulated. This study does not suggest the best-matching LFR for the DRF, but shows good possibility that the DRF fuel can be used in future LFRs.

  16. Unconventional liquid metal cooled fast reactors

    International Nuclear Information System (INIS)

    Spinrad, B.I.; Rohach, A.F.; Razzaque, M.M.

    1989-06-01

    This report describes the rationale for, design of and analytical studies on an unconventional sodium-cooled power reactor, called the Trench Reactor. It derives its name from the long, narrow sodium pool in which the reactor is placed. Unconventional features include: pool shape; reactor shape (also long and narrow); reflector control; low power density; hot-leg primary pumping; absence of a cold sodium pool; large core boxes rather than a large number of subassemblies; large diameter metal fuel; vessel suspension from cables; and vessel cooling by natural circulation of building atmosphere (nitrogen) at all times. These features all seem feasible. They result in a system that is capable of at least a ten year reload interval and shows good safety through direct physical response to loss-of-heat-sink, loss-of-flow and limited-reactivity nuclear transients. 43 figs., 43 tabs

  17. Liquid metal cooled experimental fast reactor simulator

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine; Braz Filho, Francisco; Borges, Eduardo M.; Rosa, Mauricio A.P.; Rocamora, Francisco; Hirdes, Viviane R.

    1997-01-01

    This paper is a continuation of the work that has been done in the area of fast reactor component dynamic analysis, as part of the REARA project at the IEAv/CTA-Brazil. A couple of preceding papers, presented in other meetings, introduced major concept design components of the REARA reactor. The components are set together in order to represent a full model of the power plant. Full model transient results will be presented, together with several parameters to help us to better establish the REARA experimental plant concept. (author). 8 refs., 6 figs., 3 tabs

  18. Sloshing behaviors in liquid metal reactors

    International Nuclear Information System (INIS)

    Choun, Young Sun; Yoo, Bong; Lee, Jae Han

    1998-04-01

    It is very difficult to predict sloshing response in the LMFBR which contains many internal components, since the sloshing response of a coolant depends not only upon the frequency contents of input ground motion, but also upon the configuration of reactor vessel and internal components, arrangement of internal components, and excitation direction of ground motion. Therefore, clear understanding of sloshing in the reactor vessel and development of proper technologies to predict sloshing response are necessary to establish the safety of the LMFBR against earthquake. The purpose of this report is to review sloshing behaviors in the LMFBR for the clear understanding of sloshing characteristics and to propose some interesting topics to be studied in future. (author). 34 refs., 12 tabs., 50 figs

  19. Addressing safety issues in a hybrid liquid metal reactor

    International Nuclear Information System (INIS)

    Garabedyan, D.; Ehvart, Eh.

    1988-01-01

    Hybrid design of a fast reactor with a liquid-metal coolant, combining the advantages of traditional loop (prevailing in the USA) and integral (prevailing in Europe) arrangements is described. Just as a loop reactor the hybrid one has separate arrangement of the core and the equipment of the primary circuit heat exchange. At the same time similar to the reactor with integral arrangement, the option considered has no complex pipeline system. This reduces sharply the possibility of sodium leakages which cause fires and personnel irradiation

  20. Passive safety and the advanced liquid metal reactors

    International Nuclear Information System (INIS)

    Hill, D.J.; Pedersen, D.R.; Marchaterre, J.F.

    1988-01-01

    Advanced Liquid Metal Reactors being developed today in the USA are designed to make maximum use of passive safety features. Much of the LMR safety work at Argonne National Laboratory is concerned with demonstrating, both theoretically and experimentally, the effectiveness of the passive safety features. The characteristics that contribute to passive safety are discussed, with particular emphasis on decay heat removal systems, together with examples of Argonne's theoretical and experimental programs in this area

  1. Status of liquid metal cooled fast reactor technology

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-04-01

    During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants Refs, figs, tabs

  2. Status of liquid metal cooled fast reactor technology

    International Nuclear Information System (INIS)

    1999-04-01

    During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants

  3. Performance of metallic fuels in liquid-metal fast reactors

    International Nuclear Information System (INIS)

    Seidel, B.R.; Walters, L.C.; Kittel, J.H.

    1984-01-01

    Interest in metallic fuels for liquid-metal fast reactors has come full circle. Metallic fuels are once again a viable alternative for fast reactors because reactor outlet temperature of interest to industry are well within the range where metallic fuels have demonstrated high burnup and reliable performance. In addition, metallic fuel is very tolerant of off-normal events of its high thermal conductivity and fuel behavior. Futhermore, metallic fuels lend themselves to compact and simplified reprocessing and refabrication technologies, a key feature in a new concept for deployment of fast reactors called the Integral Fast Reactor (IFR). The IFR concept is a metallic-fueled pool reactor(s) coupled to an integral-remote reprocessing and fabrication facility. The purpose of this paper is to review recent metallic fuel performance, much of which was tested and proven during the twenty years of EBR-II operation

  4. Safety characteristics of the US advanced liquid metal reactor core

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Gyorey, G.L.; Lipps, A.J.; Wu, T.

    1991-01-01

    The U.S. Advanced Liquid Metal Reactor (ALMR) design employs innovative, passive features to provide an unprecedented level of public safety and the ability to demonstrate this safety to the public. The key features employed in the core design to produce the desired passive safety characteristics are: a small core with a tight restraint system, the use of metallic U-Pu-Zr fuel, control rod withdrawal limiters, and gas expansion modules. In addition, the reactor vessel and closure are designed to have the capability to withstand, with large margins, the maximum possible core disruptive accident without breach and radiological release. (author)

  5. The US Liquid-Metal Reactor Program - overview and status

    International Nuclear Information System (INIS)

    Quinn, J.E.; Gyorey, G.L.; Salerno, L.N.

    1992-01-01

    The US Advanced Liquid-Metal Reactor (ALMR) Program has three major elements being developed in an integrated fashion to produce a system meeting the US long-term nuclear energy needs. Reactor design, one of those elements, is the focus of this paper. The other two elements, the integral fast reactor metal-fuel cycle and the light water reactor (LWR) spent-fuel actinide recycle, will be addressed in companion papers. The ALMR is adaptable to multiple missions with few modifications such as the core arrangements. The missions identified to date are (a) the extension of the existing uranium resources through breeding and highly efficient uranium utilization, (b) the recycle and utilization of the long-life actinides in LWR spent fuel as fissile material for the ALMR, and (c) the conversion of excess weapons fissil material into electricity. In addition to these missions, the reactor design is adaptable to either the metal-fuel cycle or the oxide fuel cycle

  6. Liquid metal coolant disposal from UKAEA reactors at Dounreay

    International Nuclear Information System (INIS)

    Adam, E.R.

    1997-01-01

    As part of the United Kingdom's Fast Reactor Development programme two reactors were built and operated at Dounreay in the North of Scotland. DFR (Dounreay Fast Reactor) was operated from 1959-1977 and PFR (Prototype Fast Reactor) was operated from 1974-1994. Both reactors are currently undergoing Stage 1 Decommissioning and are installing plant to dispose of the bulk coolant (DFR ∼ 60 tonne; PFR ∼ 1500 tonne). The coolant (NaK) remaining at DFR is mainly in the primary circuit which contains in excess of 500 TBq of Cs137. Disposal of 40 tonnes of secondary coolant has already been carried out. The paper will describe the processes used to dispose of this secondary circuit coolant and how it is intended the remaining primary circuit coolant will be handled. The programme to process the primary coolant will also be described which involves the conversion of the liquid metal to caustic and its decontamination. No PFR coolant Na has been disposed off to date. The paper will describe the current decommissioning programme activities relating to liquid metal disposal and treatment describing the materials to be disposed of and the issue of decontamination of the effluents. (author)

  7. Fuel rod for liquid metal-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Vinz, P.

    1976-01-01

    In fuel rods for nuclear reactors with liquid-metal cooling (sodium), with stainless steel tubes with a nitrated surface as canning, superheating or boiling delay should be avoided. The inner wall of the can is provided along its total length with a helical fin of stainless steel wire (diameter 0.05 to 0.5 mm) to be wetted by hot sodium. This fin is mounted under prestressing and has a distance in winding of 1/10 of the wire diameter. (UWI) [de

  8. Development of inelastic design method for liquid metal reactor plants

    International Nuclear Information System (INIS)

    Takahashi, Yukio; Take, Kohji; Kaguchi, Hitoshi; Fukuda, Yoshio; Uno, Tetsuro.

    1991-01-01

    Effective utilization of inelastic analysis in structural design assessment is expected to play an important role for avoiding too conservative design of liquid metal reactor plants. Studies have been conducted by the authors to develop a guideline for application of detailed inelastic analysis in design assessment. Both fundamental material characteristics tests and structural failure tests were conducted. Fundamental investigations were made on inelastic analysis method and creep-fatigue life prediction method based on the results of material characteristics tests. It was demonstrated through structural failure tests that the design method constructed based on these fundamental investigations can predict failure lives in structures subjected to cyclic thermal loadings with sufficient accuracy. (author)

  9. Hydrogen permeation resistant layers for liquid metal reactors

    International Nuclear Information System (INIS)

    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

  10. Specific power of liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Dobranich, D.

    1987-10-01

    Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs

  11. Materials requirements for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Bennett, J.W.; Horton, K.E.

    1978-01-01

    Materials requirements for Liquid Metal Fast Breeder Reactors (LMFBRs) are quite varied with requisite applications ranging from ex-reactor components such as piping, pumps, steam generators and heat exchangers to in-reactor components such as heavy section reactor vessels, core structurals, fuel pin cladding and subassembly flow ducts. Requirements for ex-reactor component materials include: good high temperature tensile, creep and fatigue properties; compatibility with high temperature flowing sodium; resistance to wear, stress corrosion cracking, and crack propagation; and good weldability. Requirements for in-reactor components include most of those cited above for ex-reactor components as supplemented by the following: resistance to radiation embrittlement, swelling and radiation enhanced creep; good neutronics; compatibility with fuel and fission product materials; and resistance to mass transfer via flowing sodium. Extensive programs are currently in place in a number of national laboratories and industrial contractors to address the materials requirements for LMFBRs. These programs are focused on meeting the near term requirements of early LMFBRs such as the Fast Flux Test Facility and the Clinch River Breeder Reactor as well as the longer term requirements of larger near-commercial and fully-commercial reactors

  12. A compact, inherently safe liquid metal reactor plant concept for terrestrial defense power applications

    International Nuclear Information System (INIS)

    Magee, P.M.; Dubberley, A.E.; Lutz, D.E.; Palmer, R.S.

    1987-01-01

    A compact, inherently safe, liquid metal reactor concept based on the GE PRISM innovative LMR design has been developed for terrestrial defense power applications in the 2-50 MWe range. The concept uses a small, sodium-cooled, U-5%Zr metal fueled reactor contained within two redundant steel vessels. The core is designed to operate at a low power density and temperature (925 F) and can operate 30 years without refueling. One two primary coolant loops, depending upon the plant size, transport heat from the core to sodium-to-air, double-wall heat exchangers. Power is produced by a gas turbine operated in a closed ''bottoming'' cycle that employs intercoolers between the compressor stages and a recuperator. Inherent safety is provided by passive means only; operator action is not required to ensure plant safety even for events normally considered Beyond Design Basis Accidents. In addition to normal shutdown heat removal via the sodium-to-air heat exchangers, the design utilizes an inherently passive radiant vessel auxiliary cooling system similar to that designed for PRISM. The use of an air cycle gas turbine eliminates the cost and complexity of the sodium-water reactor pressure relief system required for a steam cycle sodium-cooled reactor

  13. Liquid metal reactor applications of the CONTAIN code

    International Nuclear Information System (INIS)

    Carroll, D.E.; Bergeron, K.D.; Gido, R.; Valdez, G.D.; Scholtyssek, W.

    1988-01-01

    The CONTAIN code is the NRC's best-estimate code for the evaluation of the conditions that may exist inside a reactor containment building during a severe accident. Included in the phenomena modeled are thermal-hydraulics, radiant and convective heat transfer, aerosol loading and transient response, fission product transport and heating effects, and interactions of sodium and corium with the containment atmosphere and structures. CONTAIN has been used by groups in Japan and West Germany to assess its ability to analyze accident consequences for liquid metal reactor (LMR) plants. In conjunction with this use, collaborative efforts to improve the modeling have been pursued. This paper summarizes the current state of the version of CONTAIN that has been enhanced with extra capabilities for LMR applications. A description of physical models is presented, followed by a review of validation exercises performed with CONTAIN. Some demonstration calculations of an integrated LMR application are presented

  14. Vessel supporting structure for liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Mahe, Armel; Jullien, Georges

    1974-01-01

    The supporting structure described is for a liquid metal cooled nuclear reactor, the vessel being of the type suspended to the end slab of the reactor. It includes a ring connected at one of its two ends to a single shell and at the other end to two shells. One of these three shells connected to the lower end of the ring forms the upper part of the vessel to be supported. The two other shells are embedded in two sperate parts of the slab. The ring and shell assembly is housed in an annular space provided in the end slab and separating it into two parts, namely a central part and a peripheral part [fr

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

    International Nuclear Information System (INIS)

    Sayre, R.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 a 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

  16. The US Advanced Liquid-Metal Reactor Program

    International Nuclear Information System (INIS)

    Brolin, E.C.

    1992-01-01

    Based on National Energy Strategy projections, utilities will be required to substantially increase electric generating capacity over the next 40 yr to meet economic growth requirements and replace retiring capacity. Although aggressive conservation measures can save up to 85 GW(electric), ∼195 GW(electric) of additional generating capcity will still be needed by 2010. Assuming startup of new plants around 2000, US Department of Energy (DOE) analyses show that nuclear power can contribute 195 GW(electric) of capacity by 2030, or ∼20% of total electric generation. The DOE is involved in a number of strategies designed to revitalize the nuclear power industry and enable it to meet this projected need for additional capacity. Among these is an integrated overall strategy for advanced reactor development and high-level waste management. A high priority in pursuit of this strategy is the Advanced Liquid-Metal Reactor (ALMR) Program

  17. Automated scoping methodology for liquid metal natural circulation small reactor

    International Nuclear Information System (INIS)

    Son, Hyung M.; Suh, Kune Y.

    2014-01-01

    Highlights: • Automated scoping methodology for natural circulation small modular reactor is developed. • In-house code is developed to carry out system analysis and core geometry generation during scoping. • Adjustment relations are obtained to correct the critical core geometry out of diffusion theory. • Optimized design specification is found using objective function value. • Convex hull volume is utilized to quantify the impact of different constraints on the scope range. - Abstract: A novel scoping method is proposed that can automatically generate design variable range of the natural circulation driven liquid metal cooled small reactor. From performance requirements based upon Generation IV system roadmap, appropriate structure materials are selected and engineering constraints are compiled based upon literature. Utilizing ASME codes and standards, appropriate geometric sizing criteria on constituting components are developed to ensure integrity of the system during its lifetime. In-house one dimensional thermo-hydraulic system analysis code is developed based upon momentum integral model and finite element methods to deal with non-uniform descritization of temperature nodes for convection and thermal diffusion equation of liquid metal coolant. In order to quickly generate critical core dimensions out of given unit cell information, an adjustment relation that relates the critical geometry estimated from one-group diffusion and that from MCNP code is constructed and utilized throughout the process. For the selected unit cell dimension ranges, burnup calculations are carried out to check the cores can generate energy over the reactor lifetime. Utilizing random method, sizing criteria, and in-house analysis codes, an automated scoping methodology is developed. The methodology is applied to nitride fueled integral type lead cooled natural circulation reactor concept to generate design scopes which satisfies given constraints. Three dimensional convex

  18. Liquid metal reactor development. Development of LMR coolant technology

    Energy Technology Data Exchange (ETDEWEB)

    Nam, H. Y.; Choi, S. K.; Hwang, J. s.; Lee, Y. B.; Choi, B. H.; Kim, J. M.; Kim, Y. G.; Kim, M. J.; Lee, S. D.; Kang, Y. H.; Maeng, Y. Y.; Kim, T. R.; Park, J. H.; Park, S. J.; Cha, J. H.; Kim, D. H.; Oh, S. K.; Park, C. G.; Hong, S. H.; Lee, K. H.; Chun, M. H.; Moon, H. T.; Chang, S. H.; Lee, D. N.

    1997-07-15

    Following studies have been performed during last three years as the 1.2 phase study of the mid and long term nuclear technology development plan. First, the small scale experiments using the sodium have been performed such as the basic turbulent mixing experiment which is related to the design of a compact reactor, the flow reversal characteristics experiment by natural circulation which is necessary for the analysis of local flow reversal when the electromagnetic pump is installed, the feasibility test of the decay heat removal by wall cooling and the operation of electromagnetic pump. Second, the technology of operation mechanism of sodium facility is developed and the technical analysis and fundamental experiments of sodium measuring technology has been performed such as differential pressure measuring experiment, local flow rate measuring experimenter, sodium void fraction measuring experiment, under sodium facility, the free surface movement experiment and the side orifice pressure drop experiment. A new bounded convection scheme was introduced to the ELBO3D thermo-hydraulic computer code designed for analysis of experimental result. A three dimensional computer code was developed for the analysis of free surface movement and the analysis model of transmission of sodium void fraction was developed. Fourth, the small scale key components are developed. The submersible-in-pool type electromagnetic pump which can be used as primary pump in the liquid metal reactor is developed. The SASS which uses the Curie-point electromagnet and the mock-up of Pantograph type IVTM were manufactured and their feasibility was evaluated. Fifth, the high temperature characteristics experiment of stainless steel which is used as a major material for liquid metal reactor and the material characteristics experiment of magnet coil were performed. (author). 126 refs., 98 tabs., 296 figs.

  19. Radioactive waste management at a Liquid Metal Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Abrams, C.S.; Fryer, R.H.; Witbeck, L.C.

    1979-01-01

    This paper presents the radioactive waste production and management at a Liquid Metal Fast Breeder Reactor-II (EBR-II), which is operated for the US Department of Energy by the Argonne National Laboratory at the Idaho National Engineering Laboratory (INEL). Since this facility, in addition to supplying power has been used to demonstrate the breeder, fuel cycling, and recently operations with defective fuel elements, various categories of waste have been handled safely over some 14 years of operation. Liquid wastes are processed such that the resulting effluent can be discharged to an uncontrolled area. Solid wastes up to 10,000 R/hr are packaged and shipped contamination-free to a disposal site or interim storage with exposures to personnel approximately 10 mrem. Gaseous waste discharges are low such as 143 Ci of noble gases in 1978 and do not have a significant effect on the environment even with operations with breached fuel

  20. Approaches to measurement of thermal-hydraulic parameters in liquid-metal-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Sackett, J.I.

    1983-01-01

    This lecture considers instrumentation for liquid-metal-cooled fast breeder reactors (LMFBR's). Included is instrumentation to measure sodium flow, pressure, temperature, acoustic noise, and sodium purity. It is divided into three major parts: (1) measurement requirements for sodium cooled reactor systems, (2) in-core and out-of-core measurements in liquid metal systems, and (3) performance measurements of water steam generators

  1. Conceptual design strategy for liquid-metal-wall inertial-fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-02-01

    The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade provides a basis for formulating a conceptual design strategy for such chambers. Both the attractive and unattractive features of a LMW chamber are enumerated, and a design strategy is formulated which accommodates the engineering constraints while minimizing the liquid-metal flow rate.

  2. Conceptual design strategy for liquid-metal-wall inertial-fusion reactors

    International Nuclear Information System (INIS)

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

    1981-02-01

    The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade provides a basis for formulating a conceptual design strategy for such chambers. Both the attractive and unattractive features of a LMW chamber are enumerated, and a design strategy is formulated which accommodates the engineering constraints while minimizing the liquid-metal flow rate

  3. Technology which led to the westinghouse inherently safe liquid metal reactor

    International Nuclear Information System (INIS)

    Schmidt, J.E.; Coffield, R.D.; Doncals, R.A.; Kalinowski, J.E.; Markley, R.A.

    1985-01-01

    The Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor programs resulted in an understanding of liquid metal reactor behavior that is being used to design inherent safety capability into liquid metal reactors. Technological advances give the same beneficial operating characteristics of conventional liquid metal reactors, however, the addition of inherently safe design features precludes the initiation of hypothetical core disruptive accidents. These innovative features permit inherent safety capability to be demonstrated with more than adequate margins. Also, the variety of inherent safety features provides the designers with options in selecting inherent design features for a specific reactor application

  4. Liquid metal cooled reactors: Experience in design and operation

    International Nuclear Information System (INIS)

    2007-12-01

    on key fast reactor technology aspects in an integrative sense useful to engineers, scientists, managers, university students and professors. This publication has been prepared to contribute toward the IAEA activity to preserve the knowledge gained in the liquid metal cooled fast reactor (LMFR) technology development. This technology development and experience include aspects addressing not only experimental and demonstration reactors, but also all activities from reactor construction to decommissioning. This publication provides a survey of worldwide experience gained over the past five decades in LMFR development, design, operation and decommissioning, which has been accumulated through the IAEA programmes carried out within the framework of the TWG-FR and the Agency's INIS and NKMS

  5. Evaluation of liquid metal protection of a limiter/divertor in fusion reactors

    International Nuclear Information System (INIS)

    Hassanein, A.M.; Smith, D.L.

    1988-01-01

    The liquid metal protection concept is proposed mainly to prolong the lifetime of a divertor or a limiter in a fusion reactor. This attractive idea for protection requires studying a wide range of problems associated with the use of liquid-metals in fusion reactors. In this work the protection by liquid-metals has concentrated on predictions of the loss rate of the film to the plasma, the operating surface temperatures required for the film, and the potential tritium inventory requirement. The effect of plasma disruptions on the liquid metal film is also evaluated. Other problems such as liquid metal compatibility with structural materials, magnetic field effects, and the effect of liquid metal contamination on plasma performance are discussed. Three candidate liquid-metals are evaluated, i.e., lithium, gallium, and tin. A wide range of reactor operating conditions valid for both near term machines (INTOR and ITER) and for the next generation commercial reactors (TPSS) are considered. This study has indicated that the evaporation rate for candidate liquid metals can be kept below the sputtering range for reasonable operating temperatures and plasma edge conditions. At higher temperatures, evaporation dominates the losses. Impurity transport calculations indicate that impurities from the plate should not reach the main plasma. One or two millimeters of liquid films can protect the structure from severe plasma disruptions. Depending on the design of the liquid metal protection system, the tritium inventory in the liquid film is predicted to be on the order of a few grams. 16 refs., 5 figs

  6. Safety design analyses of Korea Advanced Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Suk, S.D.; Park, C.K.

    2000-01-01

    The national long-term R and D program updated in 1997 requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 MWe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self consistent design meeting a set of the major safety design requirements for accident prevention. Some of current emphasis include those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve supporting R and D programs of substance. This paper summarizes some of the results of engineering and design analyses performed for the safety of KALIMER. (author)

  7. Conceptual safety design analysis of Korea advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Suk, S. D.; Park, C. K.

    1999-01-01

    The national long-term R and D program, updated in 1977, requires Korea Atomic Energy Research Institute (KAERI) to complete by the year 2006 the basic design of Korea Advanced Liquid Metal Reactor (KALIMER), along with supporting R and D work, with the capability of resolving the issue of spent fuel storage as well as with significantly enhanced safety. KALIMER is a 150 Mwe pool-type sodium cooled prototype reactor that uses metallic fuel. The conceptual design is currently under way to establish a self-consistent design meeting a set of major safety design requirements for accident prevention. Some of the current emphasis includes those for inherent and passive means of negative reactivity insertion and decay heat removal, high shutdown reliability, prevention of and protection from sodium chemical reaction, and high seismic margin, among others. All of these requirements affect the reactor design significantly and involve extensive supporting R and D programs. This paper summarizes some of the results of conceptual engineering and design analyses performed for the safety of KALIMER in the area of inherent safety, passive decay heat removal, sodium water reaction, and seismic isolation. (author)

  8. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the Technical Meeting is to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials

  9. Potential applications of robotics in advanced liquid-metal reactors

    International Nuclear Information System (INIS)

    Carroll, D.G.; Thompson, M.L.

    1990-01-01

    The advanced liquid-metal reactor (ALMR) design includes a range of robots and automation devices. They extend from stationary robots that are a part of the current design to more exotic concepts with mobile, autonomous units, which may become part of the design. Development of robotic application requirements is enhanced by using computer models of work spaces in three dimensions. The primary goals of the more autonomous machines are to: (1) extent and/or enhance one's capabilities in a hazardous environment; some tasks could encounter high temperatures (up to 800 degree F), high radiation (fields up to several hundred thousand roentgens per hour), rooms filled with inert gas and/or sodium aerosol, or combinations of these; (2) reduce operating and maintenance cost through inservice inspection (ISI) of various parts of the reactor, through consideration of as-low-as-reasonably achievable radiation levels, and through automation of some maintenance/processing operations. This paper discusses some applications in the fuel cycle, in refueling operations, and in inspection

  10. Flow induced vibrations in liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    1989-01-01

    Flow induced vibrations are well known phenomena in industry. Engineers have to estimate their destructive effects on structures. In the nuclear industry, flow induced vibrations are assessed early in the design process, and the results are incorporated in the design procedures. In many cases, model testing is used to supplement the design process to ensure that detrimental behaviour due to flow induced vibrations will not occur in the component in question. While these procedures attempt to minimize the probability of adverse performance of the various components, there is a problem in the extrapolation of analytical design techniques and/or model testing to actual plant operation. Therefore, sodium tests or vibrational measurements of components in the reactor system are used to provide additional assurance. This report is a general survey of experimental and calculational methods in this area of structural mechanics. The report is addressed to specialists and institutions in industrialized and developing countries who are responsible for the design and operation of liquid metal fast breeder reactors. 92 refs, 90 figs, 8 tabs

  11. Apparatus for sealing a rotatable shield plug in a liquid metal nuclear reactor

    International Nuclear Information System (INIS)

    Winkleblack, R.K.

    1980-01-01

    An apparatus for sealing a rotatable shield plug in a nuclear reactor having liquid metal coolant is described. The apparatus includes a dip -ring seal adapted to provide a fluid barrier between the liquid metal and the atmosphere and to permit rotation of the shield plug. The apparatus also includes a static seal for the rotatable shield plug located between the dip-ring seal and the liquid metal. The static seal isolates the dip-ring seal from the liquid metal vapor during operation at power and can be disengaged for rotation of the shield plug

  12. Improved alloys for a liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    1981-01-01

    An alloy is specified suitable for use at elevated temperatures and especially in a liquid metal fast breeder reactor consisting essentially of a nickel-chromium steel having a specified range of composition of C, Mn, Si, Zr, V, Ni, Cr, Ti, Al, Mo, B, and the balance iron with incidental impurities, the alloy exhibiting a swelling at peak swelling temperature of less than 10% wherein the matrix composition has after heat treatment at a temperature within the range of 1000 0 C to 1100 0 C for about one half hour followed by aging at a temperature within the range of from 700 0 C to 815 0 C for a time period of between 10 to 24 hours, the longer hours being associated with the lower temperatures and vice-versa, and after the removal of the non-equilibrium gamma prime and other precipitated phases a composition within a specified range of composition of Ni, Cr, Ti, Al, Mo, the balance being essentially iron. (U.K.)

  13. Liquid-metal-gas heat exchanger for HTGR type reactors

    International Nuclear Information System (INIS)

    Werth, G.

    1980-01-01

    The aim of this study is to investigate the heat transfer characteristics of a liquid metal heat exchanger (HE) for a helium-cooled high temperature reactor. A tube-type heat exchanger is considered as well as two direct exchangers: a bubble-type heat exchanger and a heat exchanger according to the spray principle. Experiments are made in order to determine the gas content of bubble-type heat exchangers, the dependence of the droplet diameter on the nozzle diameter, the falling speed of the droplets, the velocity of the liquid jet, and the temperature variation of liquid jets. The computer codes developed for HE calculation are structured so that they may be used for gas/liquid HE, too. Each type of HE that is dealt with is designed by accousting for a technical and an economic assessment. The liquid-lead jet spray is preferred to all other types because of its small space occupied and its simple design. It shall be used in near future in the HTR by the name of lead/helium HE. (GL) [de

  14. Dispersion strengthened ferritic alloy for use in liquid-metal fast breeder reactors (LMFBRS)

    International Nuclear Information System (INIS)

    Fischer, J.J.

    1978-01-01

    A dispersion-strengthened ferritic alloy is provided which has high-temperature strength and is readily fabricable at ambient temperatures, and which is useful as structural elements of liquid-metal fast breeder reactors. 4 tables

  15. Small liquid metal reactor for an initial phase of fast breeder reactor introduction

    International Nuclear Information System (INIS)

    Ishiguro, Y.; Nascimento, J.A. do.

    1985-01-01

    Safety and burnup characteristics of a 1000 MWth liquid metal reactor have been examined for various fuel types. With metallic Pu/Th fuel containing a small amount of zirconium hydride, low sodium-void reactivity, a high Doppler coefficient, and small burnup reactivity swings can be achieved. A conservative design is considered for an initial phase of fast breeder reactor development and possible modifications are discussed. (Author) [pt

  16. The role of SASSYS-1 in LMR [Liquid Metal Reactor] safety analysis

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  18. Liquid metal reactor head designs in the USA - heat and mass transfer considerations

    International Nuclear Information System (INIS)

    Burke, T.M.

    1986-01-01

    Development of liquid metal reactor plants in the United States over the past 30 years has resulted in an evolution of reactor head designs as reflected in the SRE, Hallam, EBR-II and FFTF plants. This evolution has probably been affected to some extent by the fact that, in contrast to most other countries, there is no single organization in the United States which has been responsible for the design of liquid metal reactor plants. The current U.S. LMR design efforts involve two innovative design consortiums (guided by the US Department of Energy) and a joint industry venture on the Large Scale Prototype Breeder. It is therefore somewhat difficult to provide a statement on the philosophy of the reactor head design in the U.S. This paper however briefly describes the existing and proposed U.S. liquid metal reactor head designs and in the process, attempt to provide some insight on the basis for those designs

  19. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Joo, Young Sang; Lee, Jae Han

    2007-02-01

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

  20. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  1. Status of liquid metal reactor development in the United States of America

    Energy Technology Data Exchange (ETDEWEB)

    Griffith, Jerry D [Reactor Systems Development and Technology, Office of Nuclear Energy, U.S. Department of Energy (United States); Horton, Kenneth E [Division of International Programs, Office of Nuclear Energy, U.S. Department of Energy (United States)

    1992-07-01

    The U.S. civilian nuclear power research and development program continues to focus on advanced large and mid-size light water reactors, advanced liquid metal fast reactors, and modular high temperature gas cooled reactors. This paper discusses the Advanced Liquid Metal Reactor program, which is composed of a small, passively safe fast reactor coupled with a metal fuel cycle that incorporates actinide recycle, and an emerging effort to process LWR spent fuel for LMR fissile material, and to enhance the LWR waste management. The liquid metal reactor concept has a sound technology base, with some three decades of research and development both in this and other countries. An existing network of government and industry research facilities and engineering test centers in the United States is currently providing test capabilities and the technical expertise required to conduct an aggressive advanced reactor development program. Notable among the research facilities is the Experimental Breeder Reactor-Il (EBR-II) at Argonne National Laboratory (ANL) in Idaho and the Fast Flux Test Facility (FFTF) at Hanford, Washington. Subsequent to the DOE directive to shut down the Fast Flux Test Facility in early 1990, significant effort was placed in finding international financial support for this reactor. This initiative was not successful. Therefore, although there may be a potential future mission for the FFTF, the Secretary of Energy announced on March 13, 1992 that the FFTF will be put in a standby condition starting April 1, 1992. Current U.S. Advanced Liquid Metal Reactor (ALMR) activity is focused on providing a reactor and fuel cycle system with improved safety margins, better economics, and an attractive waste management (actinide recycle) option. Special attention is being directed to passive safety features, large design margins, modular plant construction, standardized plant design leading to simplified licensing and shorter construction schedules, factory fabrication

  2. Status of liquid metal reactor development in the United States of America

    International Nuclear Information System (INIS)

    Griffith, Jerry D.; Horton, Kenneth E.

    1992-01-01

    The U.S. civilian nuclear power research and development program continues to focus on advanced large and mid-size light water reactors, advanced liquid metal fast reactors, and modular high temperature gas cooled reactors. This paper discusses the Advanced Liquid Metal Reactor program, which is composed of a small, passively safe fast reactor coupled with a metal fuel cycle that incorporates actinide recycle, and an emerging effort to process LWR spent fuel for LMR fissile material, and to enhance the LWR waste management. The liquid metal reactor concept has a sound technology base, with some three decades of research and development both in this and other countries. An existing network of government and industry research facilities and engineering test centers in the United States is currently providing test capabilities and the technical expertise required to conduct an aggressive advanced reactor development program. Notable among the research facilities is the Experimental Breeder Reactor-Il (EBR-II) at Argonne National Laboratory (ANL) in Idaho and the Fast Flux Test Facility (FFTF) at Hanford, Washington. Subsequent to the DOE directive to shut down the Fast Flux Test Facility in early 1990, significant effort was placed in finding international financial support for this reactor. This initiative was not successful. Therefore, although there may be a potential future mission for the FFTF, the Secretary of Energy announced on March 13, 1992 that the FFTF will be put in a standby condition starting April 1, 1992. Current U.S. Advanced Liquid Metal Reactor (ALMR) activity is focused on providing a reactor and fuel cycle system with improved safety margins, better economics, and an attractive waste management (actinide recycle) option. Special attention is being directed to passive safety features, large design margins, modular plant construction, standardized plant design leading to simplified licensing and shorter construction schedules, factory fabrication

  3. US/DOE Man-Machine Integration program for liquid metal reactors

    International Nuclear Information System (INIS)

    D'Zmura, A.P.; Seeman, S.E.

    1985-03-01

    The United States Department of Energy (DOE) Man-Machine Integration program was started in 1980 as an addition to the existing Liquid Metal Fast Breeder Reactor safety base technology program. The overall goal of the DOE program is to enhance the operational safety of liquid metal reactors by optimum integration of humans and machines in the overall reactor plant system and by application of the principles of human-factors engineering to the design of equipment, subsystems, facilities, operational aids, procedures and environments. In the four years since its inception the program has concentrated on understanding the control process for Liquid Metal Reactors (LMRs) and on applying advanced computer concepts to this process. This paper describes the products that have been developed in this program, present computer-related programs, and plans for the future

  4. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Working Material

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the TM on “Liquid metal reactor concept: core design and structural materials” was to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials. Main results arising from national and international R&D programmes and projects in the field were reviewed, and new activities to be carried out under the IAEA aegis were identified on the basis of the analysis of current research and technology gaps

  5. PRISM reactor. An option for plutonium disposition?

    Energy Technology Data Exchange (ETDEWEB)

    Fehlinger, Sebastian; Friess, Friederike; Kuett, Moritz [IANUS, Technische Universitaet Darmstadt (Germany)

    2015-07-01

    The Power Reactor Innovative Small Module (PRISM) is sodium cooled fast reactor model. The energy output depends on the core configuration, however with an energy output of approximately 300 MWe, the PRISM reactor belongs to the class of small modular reactors. Beside using the reactor as a breeder reactor or for the transmutation of nuclear waste, it might also be used as a burner reactor for separated plutonium. This includes for example U.S.-American excess weapon-grade plutonium as well as separated reactor-grade plutonium. Recently, there has been an ongoing discussion in GB to use the PRISM reactor to dispose their excess civilian plutonium. Depending on the task, the core configuration varies slightly. We will present different layouts and the matching MCNP models, these models can then be used to conduct depletion calculations. From these results, analysis of the change in the plutonium isotopics in the spent fuel, the amount of fissioned plutonium, and the possible annual plutonium throughputs is possible.

  6. Demonstrating electromagnetic control of free-surface, liquid-metal flows relevant to fusion reactors

    Science.gov (United States)

    Hvasta, M. G.; Kolemen, E.; Fisher, A. E.; Ji, H.

    2018-01-01

    Plasma-facing components (PFC’s) made from solid materials may not be able to withstand the large heat and particle fluxes that will be produced within next-generation fusion reactors. To address the shortcomings of solid PFC’s, a variety of liquid-metal (LM) PFC concepts have been proposed. Many of the suggested LM-PFC designs rely on electromagnetic restraint (Lorentz force) to keep free-surface, liquid-metal flows adhered to the interior surfaces of a fusion reactor. However, there is very little, if any, experimental data demonstrating that free-surface, LM-PFC’s can actually be electromagnetically controlled. Therefore, in this study, electrical currents were injected into a free-surface liquid-metal that was flowing through a uniform magnetic field. The resultant Lorentz force generated within the liquid-metal affected the velocity and depth of the flow in a controllable manner that closely matched theoretical predictions. These results show the promise of electromagnetic control for LM-PFC’s and suggest that electromagnetic control could be further developed to adjust liquid-metal nozzle output, prevent splashing within a tokamak, and alter heat transfer properties for a wide-range of liquid-metal systems.

  7. Report on generation IV technical working group 3 : liquid metal reactors

    International Nuclear Information System (INIS)

    Lineberry, M. J.; Rosen, S. L.; Sagayama, Y.

    2002-01-01

    This paper reports on the first round of R and D roadmap activities of the Generation IV (Gen IV) Technical Working Group (TWG) 3, on liquid metal-cooled reactors. Liquid metal coolants give rise to fast spectrum systems, and thus the reactor systems considered in this TWG are all fast reactors. Gas-cooled fast reactors are considered in the context of TWG 2. As is noted in other Gen IV papers, this first round activity is termed ''screening for potential'', and includes collecting the most complete set of liquid metal reactor/fuel cycle system concepts possible and evaluating the concepts against the Gen IV principles and goals. Those concepts or concept groups that meet the Gen IV principles and which are deemed to have reasonable potential to meet the Gen IV goals will pass to the next round of evaluation. Although we sometimes use the terms ''reactor'' or ''reactor system'' by themselves, the scope of the investigation by TWG 3 includes not only the reactor systems, but very importantly the closed fuel recycle system inevitably required by fast reactors. The response to the DOE Request for Information (RFI) on liquid metal reactor/fuel cycle systems from principal investigators, laboratories, corporations, and other institutions, was robust and gratifying. Thirty three liquid metal concept descriptions, from eight different countries, were ultimately received. The variation in the scope, depth, and completeness of the responses created a significant challenge for the group, but the TWG made a very significant effort not to screen out concepts early in the process

  8. A conceptual design strategy for liquid-metal-wall inertial fusion reactors

    International Nuclear Information System (INIS)

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

    1981-01-01

    The liquid-metal-wall chamber has emerged as an attractive reactor concept for inertial fusion energy conversion. The principal feature of this concept is a thick, free-flowing blanket of liquid metal used to protect the structure of the reactor. The development and design of liquid-metal-wall chambers over the past decade are reviewed from the perspective of formulating a conceptual design strategy for such chambers. The basis for the design strategy is set by enumerating both the attractive and unattractive features of a LMW chamber. Past concepts are then reviewed to identify conceptual design approaches and physical configurations that enhance the positive aspects and minimize the negative aspects. A detailed description of the engineering considerations is given, including such topics as the selection of a liquid metal, control of radiation damage, selection of structural material, control of tritium breeding and extraction, control of wall stress, and designing for a given rep-rate. Finally, a design strategy is formulated which accomodates the engineering constraints while minimizing the liquid-metal flow rate. (orig.)

  9. Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Solomon, K.A.

    1979-07-01

    In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

  10. Steam water cycle chemistry of liquid metal cooled innovative nuclear power reactors

    International Nuclear Information System (INIS)

    Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

    2012-09-01

    The Federal Target Program (FTP) of Russian Federation 'Nuclear Energy Technologies of the New Generation for 2010-2015 and for Perspective up to 2020' is aimed at development of advanced nuclear energy technologies on the basis of closed fuel cycle with fast reactors. There are advanced fast reactor technologies of the 4. generation with liquid metal cooled reactors. Development stages of maturity of fast sodium cooled reactor technology in Russia includes experimental reactors BR-5/10 (1958-2002) and BOR-60 (since 1969), nuclear power plants (NPPs) with BN-350 (1972-1999), BN-600 (since 1980), BN-800 (under construction), BN-1200 (under development). Further stage of development of fast sodium cooled reactor technology in Russia is commercialization. Lead-bismuth eutectic fast reactor technology has been proven at industrial scale for nuclear submarines in former Soviet Union. Lead based technology is currently under development and need for experimental justification. Current status and prospects of State Corporation 'Rosatom' participation in GIF activities was clarified at the 31. Meeting of Policy Group of the International Forum 'Generation-IV', Moscow, May 12-13, 2011. In June, 2010, 'Rosatom' joined the Sodium Fast Reactor Arrangement as an authorized representative of the Russian Government. It was also announced the intention of 'Rosatom' to sign the Memorandum on Lead Fast Reactor based on Russia's experience with lead-bismuth and lead cooled fast reactors. In accordance with the above FTP some innovative liquid metal cooled reactors of different design are under development in Russia. Gidropress, well known as WER designer, develops innovative lead-bismuth eutectic cooled reactor SVBR-100. NIKIET develops innovative lead cooled reactor BRESTOD-300. Some other nuclear scientific centres are also involved in this activity, e.g. Research and Development Institute for Power Engineering (RDIPE). Optimum

  11. Compilation of data and descriptions for United States and foreign liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Appleby, E.R.

    1975-08-01

    This document is a compilation of design and engineering information pertaining to liquid metal cooled fast breeder reactors which have operated, are operating, or are currently under construction, in the United States and abroad. All data has been taken from publicly available documents, journals, and books

  12. Induction apparatus monitoring structural strains in liquid-metal-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Dean, S.A.; Evans, R.A.

    1981-01-01

    An improved method of monitoring induced torsional and linear strains in the internal structures of liquid metal cooled nuclear reactors is described. An electrical induction apparatus indicates the variation of magnetic coupling caused by a ferromagnetic member of the apparatus being subjected to such strains. (U.K.)

  13. Coolant circulation system for a liquid metal nuclear reactor

    International Nuclear Information System (INIS)

    DeLuca, R.A.; Garabedian, G.

    1988-01-01

    This patent describes a liquid metal circulation system comprising an electromagnetic pump comprised of: (a) an elongated cylindrical pump support housing; (b) a cylindrical pressure dome structure coaxially situated and supported within the pump support housing, having a closed, hemispherical upper end and an open, cylindrical lower end; (c) a cylindrical pump coaxially situated within the pressure dome structure including: (1) a central core body of laminated transformer steel having six peripherally equally spaced helical grooves on its outer surface extending the entire length of the central core body, (2) a multiplicity of square, ceramic insulated copper wires situated in the helical grooves, (3) electrical leads extending from the terminal ends of the square copper wires through the upper end of the pressure dome structure and to a three-phase low voltage/high amperage power source, (4) an austenitic stainless steel jacket covering the outer surface of the central core body and covering the helically coiled square copper wires, the outer stainless steel jacket and the inner surface of the pressure dome structure defining an annular flow passage

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

    Science.gov (United States)

    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).

  15. The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

  16. The uncertainty analysis of a liquid metal reactor for burning minor actinides from light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    The neutronics analysis of a liquid metal reactor for burning minor actinides has shown that uncertainties in the nuclear data of several key minor actinide isotopes can introduce large uncertainties in the predicted performance of the core. A comprehensive sensitivity and uncertainty analysis was performed on a 1200 MWth actinide burner designed for a low burnup reactivity swing, negative doppler coefficient, and low sodium void worth. Sensitivities were generated using depletion perturbation methods for the equilibrium cycle of the reactor and covariance data was taken ENDF-B/V and other published sources. The relative uncertainties in the burnup swing, doppler coefficient, and void worth were conservatively estimated to be 180%, 97%, and 46%, respectively. 5 refs., 1 fig., 3 tabs. (Author)

  17. Modeling of liquid-metal corrosion/deposition in a fusion reactor blanket

    International Nuclear Information System (INIS)

    Malang, S.; Smith, D.L.

    1984-04-01

    A model has been developed for the investigation of the liquid-metal corrosion and the corrosion product transport in a liquid-metal-cooled fusion reactor blanket. The model describes the two-dimensional transport of wall material in the liquid-metal flow and is based on the following assumptions: (1) parallel flow in a straight circular tube; (2) transport of wall material perpendicular to the flow direction by diffusion and turbulent exchange; in flow direction by the flow motion only; (3) magnetic field causes uniform velocity profile with thin boundary layer and suppresses turbulent mass exchange; and (4) liquid metal at the interface is saturated with wall material. A computer code based on this model has been used to analyze the corrosion of ferritic steel by lithium lead and the deposition of wall material in the cooler part of a loop. Three cases have been investigated: (1) ANL forced convection corrosion experiment (without magnetic field); (2) corrosion in the MARS liquid-metal-cooled blanket (with magnetic field); and (3) deposition of wall material in the corrosion product cleanup system of the MARS blanket loop

  18. Safety aspects of LMR [liquid metal-cooled reactor] core design

    International Nuclear Information System (INIS)

    Cahalan, J.E.

    1986-01-01

    Features contributing to increased safety margins in liquid metal-cooled reactor (LMR) design are identified. The technical basis is presented for the performance of a pool-type reactor system with an advanced metallic alloy fuel in unprotected accidents. Results are presented from analyses of anticipated transients without scram, including loss-of-flow (LOF), transient overpower (TOP), and loss-of-heat-sink (LOHS) accidents

  19. Application of probabilistic risk assessment to advanced liquid metal reactor designs

    International Nuclear Information System (INIS)

    Carroll, W.P.; Temme, M.I.

    1987-01-01

    The United States Department of Energy (US DOE) has been active in the development and application of probabilistic risk assessment methods within its liquid metal breeder reactor development program for the past eleven years. These methods have been applied to comparative risk evaluations, the selection of design features for reactor concepts, the selection and emphasis of research and development programs, and regulatory discussions. The application of probabilistic methods to reactors which are in the conceptual design stage presents unique data base, modeling, and timing challenges, and excellent opportunities to improve the final design. We provide here the background and insights on the experience which the US DOE liquid metal breeder reactor program has had in its application of probabilistic methods to the Clinch River Breeder Reactor Plant project, the Conceptual Design State of the Large Development Plant, and updates on this design. Plans for future applications of probabilistic risk assessment methods are also discussed. The US DOE is embarking on an innovative design program for liquid metal reactors. (author)

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

    International Nuclear Information System (INIS)

    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

  1. An overview of IPPE research on liquid metal fast reactor thermohydraulics

    International Nuclear Information System (INIS)

    Sorokin, A. P.; Efanov, A. D.; Zhukov, A. V.; Bogoslovskaia, G. P.

    2003-01-01

    The paper presents brief information on the most significant researches in the fields of liquid metal hydrodynamics and heat transfer performed in the State Scientific Center of Russian Federation 'Institute for Physics and Power Engineering' named after A.I.Leypunski applied to sodium-cooled fast reactors. Experimental methods for studying liquid metal thermohydraulics and applied measurement techniques are overviewed briefly in the paper. Some results of fundamental thermohydraulic investigations, such as quasi-universal character of velocity and temperature profile in liquid metals, if considered normally to the channel wall etc. are presented. Specific features of heat transfer in liquid metal cooled fuel subassembly are mentioned, among them there are: high level of coolant temperature; significant influence of an interchannel exchange on velocity and temperature distribution; an availability of contact thermal resistance; large azimuthal non-uniformity of velocity and temperature; 'conjugate' problem of heat transfer in combined geometry of fuel pin; an absence of stabilization of heat transfer in non-standard channels; an influence of non-uniform heat generation. Special attention is given to the temperature fields in fuel subassembly subjected to deformation because of radioactive swelling and creeping, as well as in case of blockage of a part of subassembly cross section. Some results of thermohydraulic investigation are demonstrated for intermediate heat exchangers, pressurized head collectors. Also the developed methods and codes of thermohydraulic calculations applied to fast reactor core are considered: subchannel approach, porous body model

  2. Independent verification: operational phase liquid metal breeder reactors

    International Nuclear Information System (INIS)

    Bourne, P.B.

    1981-01-01

    The Fast Flux Test Facility (FFTF) recently achieved 100-percent power and now is in the initial stages of operation as a test reactor. An independent verification program has been established to assist in maintaining stable plant conditions, and to assure the safe operation of the reactor. Independent verification begins with the development of administrative procedures to control all other procedures and changes to the plant configurations. The technical content of the controlling procedures is subject to independent verification. The actual accomplishment of test procedures and operational maneuvers is witnessed by personnel not responsible for operating the plant. Off-normal events are analyzed, problem reports from other operating reactors are evaluated, and these results are used to improve on-line performance. Audits are used to confirm compliance with established practices and to identify areas where individual performance can be improved

  3. Liquid Metal Fast Breeder Reactor plant maintenance and equipment design

    International Nuclear Information System (INIS)

    Swannack, D.L.

    1982-01-01

    This paper provides a summary of maintenance equipment considerations and actual plant handling experiences from operation of a sodium-cooled reactor, the Fast Flux Test Facility (FFTF). Equipment areas relating to design, repair techniques, in-cell handling, logistics and facility services are discussed. Plant design must make provisions for handling and replacement of components within containment or allow for transport to an ex-containment area for repair. The modular cask assemblies and transporter systems developed for FFTF can service major plant components as well as smaller units. The plant and equipment designs for the Clinch River Breeder Reactor (CRBR) plant have been patterned after successful FFTF equipment

  4. Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-04-20

    The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treated separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s

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

    International Nuclear Information System (INIS)

    Modarres, D.

    1991-01-01

    This patent describes a device for detecting sodium leaks from a reactor vessel of a liquid sodium cooled nuclear reactor the reactor vessel being concentrically surrounded by a a containment vessel so as to define an airtight gap containing argon. It comprises: a light source for generating a first light beam, the first light beam having first and second predominant wavelengths, the first wavelength being substantially equal to an absorption line of sodium and the second wavelength being chosen such that it is not absorbed by sodium and argon; an optical multiplexer optically coupled to the light source; optically coupled to the multiplexer, each of the sensors being embedded in the containment vessel of the reactor, each of the sensors projecting the first light beam into the gap and collecting the first light beam after it has reflected off of a surface of the reactor vessel; a beam splitter optically coupled to each of the sensors through the multiplexer, the beam splitter splitting the first light beam into second and third light beams of substantially equal intensities; a first filter dispersed within a path of second light beam for filtering the second wavelength out of the third light beam; first and second detector beams disposed with in the paths of the second and third light beams so as to detect the intensities of the second and third light beams, respectively; and processing means connected to the first and second detector means for calculating the amount of the first wavelength which is absorbed when passing through the argon

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

    International Nuclear Information System (INIS)

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

    1999-01-01

    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

  7. The development on the methodology of the initiating event frequencies for liquid metal reactor KALIMER

    International Nuclear Information System (INIS)

    Jeong, K. S.; Yang, Z. A.; Ah, Y. B.; Jang, W. P.; Jeong, H. Y.; Ha, K. S.; Han, D. H.

    2002-01-01

    In this paper, the PSA methodology of PRISM,Light Water Reactor, Pressurized Heavy Water Reactor are analyzed and the methodology of Initiating Events for KALIMER are suggested. Also,the reliability assessment of assumptions for Pipes Corrosion Frequency is set up. The reliability assessment of Passive Safety System, one of Main Safety System of KALIMER, are discussed and analyzed

  8. Status of liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    1985-01-01

    This document represents a compilation of the information on the status of fast breeder reactor development. It is intended to provide complete and authoritative information for academic, energy, industrial and planning organizations in the IAEA Member States. The Report also provides extended reference and bibliography lists. A summarized overview of the national programmes of LMFBR development is given in Chapter II. Chapter III on LMFBR experience provides a brief description and purpose of all fast reactors - experimental, demonstration and commercial size - that have been or are planned for construction and operation. Fast reactor physics is dealt with in Chapter IV. Besides the basic facts and definitions of neutronics and the compilation and measurement of nuclear data, a broad range of the calculation methods, codes, and the state of the art is described. In Chapter V, fuels and materials are described. The emphasis is on the design and development experience gained with mixed oxide fuel pins and subassemblies. Structural materials, blanket elements and absorber materials are also discussed. Chaper VI presents a broad overview of the technical and engineering aspects of LMFBR power plants. LMFBR core design is described in detail, followed by the components of the main heat transport system, the refuelling equipment, and auxiliary systems. Chapter VII on safety is a compilation of the current safety design concepts of LMFBRs and new trends in safety criteria and safety goals. The chapter concludes with risk analyses of LMFBR technology. In Chapter VIII, the systems approach has been emphasized in the consideration of the whole LMFBR fuel cycle. Special emphasis is placed on safeguards aspects and the environmental impact of the LMFBR fuel cycle. Chapter IX describes deployment considerations of LMFBRs. Special emphasis is placed on economic aspects of the LMFBR power plant and its related fuel cycle. Finally, Chapter X provides an overall summary and a

  9. Control Rod Driveline Reactivity Feedback Model for Liquid Metal Reactors

    International Nuclear Information System (INIS)

    Kwon, Young-Min; Jeong, Hae-Yong; Chang, Won-Pyo; Cho, Chung-Ho; Lee, Yong-Bum

    2008-01-01

    The thermal expansion of the control rod drivelines (CRDL) is one important passive mitigator under all unprotected accident conditions in the metal and oxide cores. When the CRDL are washed by hot sodium in the coolant outlet plenum, the CRDL thermally expands and causes the control rods to be inserted further down into the active core region, providing a negative reactivity feedback. Since the control rods are attached to the top of the vessel head and the core attaches to the bottom of the reactor vessel (RV), the expansion of the vessel wall as it heats will either lower the core or raise the control rods supports. This contrary thermal expansion of the reactor vessel wall pulls the control rods out of the core somewhat, providing a positive reactivity feedback. However this is not a safety factor early in a transient because its time constant is relatively large. The total elongated length is calculated by subtracting the vessel expansion from the CRDL expansion to determine the net control rod expansion into the core. The system-wide safety analysis code SSC-K includes the CRDL/RV reactivity feedback model in which control rod and vessel expansions are calculated using single-nod temperatures for the vessel and CRDL masses. The KALIMER design has the upper internal structures (UIS) in which the CRDLs are positioned outside the structure where they are exposed to the mixed sodium temperature exiting the core. A new method to determine the CRDL expansion is suggested. Two dimensional hot pool thermal hydraulic model (HP2D) originally developed for the analysis of the stratification phenomena in the hot pool is utilized for a detailed heat transfer between the CRDL mass and the hot pool coolant. However, the reactor vessel wall temperature is still calculated by a simple lumped model

  10. Thermal hydraulic considerations in liquid-metal-cooled components of tokamak fusion reactors

    International Nuclear Information System (INIS)

    Picologlou, B.F.; Reed, C.B.; Hua, T.Q.

    1989-01-01

    The basic considerations of MHD thermal hydraulics for liquid-metal-cooled blankets and first walls of tokamak fusion reactors are discussed. The liquid-metal MHD program of Argonne National Laboratory (ANL) dedicated to analytical and experimental investigations of reactor relevant MHD flows and development of relevant thermal hydraulic design tools is presented. The status of the experimental program and examples of local velocity measurements are given. An account of the MHD codes developed to date at ANL is also presented as is an example of a 3-D thermal hydraulic analysis carried out with such codes. Finally, near term plans for experimental investigations and code development are outlined. 20 refs., 8 figs., 1 tab

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

    International Nuclear Information System (INIS)

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

    1991-01-01

    In this paper, the conceptual development and analysis of a proposed mixed nitride driver and blanket fuel system for a prototypic advanced liquid metal reactor design is performed. As a first step, an intensive literature survey is completed on the development and testing of nitride fuel systems. Based on the results of this survey, prototypic mixed nitride fuel and blanket pins is designed and analyzed using the SIEX computer code. The analysis predicts 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

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

    International Nuclear Information System (INIS)

    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

  13. Seismic analysis of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Gilbert, R.J.; Martelli, A.

    1989-06-01

    This report is a general survey of the recent methods to predict the seismic structural behaviour of LMFBRs. It shall put into evidence the impact of seismic analysis on the design of the different structures of the reactor. This report is addressed to specialists and institutions of governmental organizations in industrialized and developing countries responsible for the design and operation of LMFBRs. The information presented should enable specialists in the R and D institutions and industries likely to be involved, to establish the correct course of the design and operation of LMFBRs. Also, the safety aspect of seismic risk are emphasized in the report. Refs and figs

  14. Liquid metal cooled fast breeder nuclear reactor constructions

    International Nuclear Information System (INIS)

    Chesworth, G.; Hind, J.R.; Hodgson, D.; Seed, G.

    1981-01-01

    In a nuclear reactor of the pool kind the primary vessel and fuel assembly are carried from the roof of the containment vault by tie straps. The primary vessel incorporates an annular yoke of 'k' cross-section the tie straps being attached to the upwardly directed vertical leg and the downwardly directed inclined leg. The upper and lower strakes of the primary vessel are extensions of the remaining legs. Load supporting welds therefore are of intermittent nature thereby limiting the effects of weld crack propagation

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  16. Reactor advantages of the belt pinch and liquid metal walls

    International Nuclear Information System (INIS)

    Kotschenreuther, M.; Manickam, J.; Menard, J.; Rappaport, H.; Zheng Linjin; Dorland, B.; Miller, R.; Turnbull, A.

    2001-01-01

    MHD stability of highly elongated tokamaks (termed a belt pinch) are considered for high bootstrap fraction cases. By employing high triangularity or indentation, and invoking wall stabilization, and β can be increased by a factor of roughly 3 by increasing κ from 2 to 4. Axisymmetric stability up to κ=4 tolerable by employing a shell which conforms more closely to the boundary than in present experiments. Engineering difficulties with a close fitting shell in a reactor environment may be overcome by employing a liquid lithium alloy shell. Rapid metal flows can lead to potentially deleterious plasma shifts and damping of the flow. (author)

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

    International Nuclear Information System (INIS)

    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

  18. Preapplication safety evaluation report for the Sodium Advanced Fast Reactor (SAFR) liquid-metal reactor

    International Nuclear Information System (INIS)

    King, T.L.; Landry, R.R.; Throm, E.D.; Wilson, J.N.

    1991-12-01

    This safety evaluation report (SER) presents the final results of a preapplication design review for the Sodium Advanced Fast Reactor (SAFR) liquid metal reactor (Project 673). The SAFR conceptual design was submitted by the US Department of Energy (DOE) in accordance with the US Nuclear Regulatory Commission (NRC) ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 FR 24643 which provides for the early Commission review and interaction). The standard SAFR plant design consists of four identical reactor modules, referred to as ''paks,'' each with a thermal output rating of 900 MWt, coupled with four steam turbine-generator sets. The total electrical output was held to be 1400 MWe. This SER represents the NRC staff's preliminary technical evaluation of the safety features in the SAFR design. It must be recognized that final conclusions in all matters discussed in this SER require approval by the Commission. During the NRC staff review of the SAFR conceptual design, DOE terminated work on this design in September 1988. This SER documents the work done to that date and no additional work is planned for the SAFR

  19. Comparison of calculational methods for liquid metal reactor shields

    International Nuclear Information System (INIS)

    Carter, L.L.; Moore, F.S.; Morford, R.J.; Mann, F.M.

    1985-09-01

    A one-dimensional comparison is made between Monte Carlo (MCNP), discrete ordinances (ANISN), and diffusion theory (MlDX) calculations of neutron flux and radiation damage from the core of the Fast Flux Test Facility (FFTF) out to the reactor vessel. Diffusion theory was found to be reasonably accurate for the calculation of both total flux and radiation damage. However, for large distances from the core, the calculated flux at very high energies is low by an order of magnitude or more when the diffusion theory is used. Particular emphasis was placed in this study on the generation of multitable cross sections for use in discrete ordinates codes that are self-shielded, consistent with the self-shielding employed in the generation of cross sections for use with diffusion theory. The Monte Carlo calculation, with a pointwise representation of the cross sections, was used as the benchmark for determining the limitations of the other two calculational methods. 12 refs., 33 figs

  20. Liquid Metal Fast Breeder Reactor Program: Argonne facilities

    International Nuclear Information System (INIS)

    Stephens, S.V.

    1976-09-01

    The objective of the document is to present in one volume an overview of the Argonne National Laboratory test facilities involved in the conduct of the national LMFBR research and development program. Existing facilities and those under construction or authorized as of September 1976 are described. Each profile presents brief descriptions of the overall facility and its test area and data relating to its experimental and testing capability. The volume is divided into two sections: Argonne-East and Argonne-West. Introductory material for each section includes site and facility maps. The profiles are arranged alphabetically by title according to their respective locations at Argonne-East or Argonne-West. A glossary of acronyms and letter designations in common usage to describe organizations, reactor and test facilities, components, etc., involved in the LMFBR program is appended

  1. Concept of the new generation high safety liquid metal reactor (LMFR)

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Zverkov, Y.A.; Morozov, A.G.; Orlov, V.V.; Ponomarev-Stepnoi, N.N.; Proshkin, A.A.; Slesarev, I.S.; Subbotin, S.A.

    1988-01-01

    The comparative analysis of the inner stability of the liquid metal reactors to severe accidents was made using the asymptotic reactivity balance. The group of the BN-reactors, Superphenix, IFR, LMFR were considered. This paper lists the characteristics of the reactors, used in the self-protectiveness analysis. The authors present the maximum coolant temperatures in post-accident asymptotic state for IFRs as on of the possible designs of a high safety fast reactor with metal fuel, U-Pu-Zr and LMFR. As is known, these values are very important for assessment of the ATWS accidence consequences. The authors consider the following situations and their combinations: loss of reactor coolant flow-LOFWS, loss of heat sink-LOHSWS, uncontrolled reactor sodium overcooling (down to the freezing point)-OVCWS, uncontrolled excess reactivity insertion-TOPWS. The calculation results demonstrate a high stability of the IFR and LMFR reactors to the most severe accidence sequences

  2. Liquid metal reactor development. Development of LMR design technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Cheol; Kim, Y I; Kim, Y G; Kim, E K; Song, H; Chung, H T; Sim, Y S; Min, B T; Kim, Y S; Wi, M H; Yoo, B; Lee, J H; Lee, H Y; Kim, J B; Koo, G H; Hahn, D H; Na, B C; Hwang, W; Nam, C; Ryu, W S; Lim, G S; Kim, D H; Kim, J D; Gil, C S

    1997-07-01

    This project was performed in five parts, the scope and contents of which are as follows: The nuclear data processing system was established and the KFS group constant library was improved and verified. Basic computation system was constructed by either developing or adding its function. Input/output (I/O) interface processing was developed to establish an integrated calculation system for LMR core nuclear rand thermal-hydraulic design and analysis. An experimental data analysis was performed to validate the constructed core neutronic calculation system. Using the established core calculation system and design technology, preliminary core design and performance analysis on the domestic LMR core design concept were carried out. To develop the basic technology of the LMR system analysis, LMR system behavior characteristics evaluation, thermal -fluid system analysis in the reactor pool, preliminary overall plant analysis and computer codes development have been performed. A porous model and simple one-dimensional model have been evaluated for the reactor pool analysis. The evaluation of the residual heat removal system on different design concepts has been also conducted. For the development of high temperature structural analysis, the heat transfer and thermal stress analyses were performed using finite element program with user subroutine that has been developed with an implementation of the Chaboche constitutive model for inelastic analysis capability, and the evaluation of creep-fatigue and ratcheting behavior of high temperature structure was carried out using this program. for development of the seismic isolation system and to predict the shear behavior for the laminated rubber bearing were established. And the behavior tests of isolation bearing and rubber specimens were carried out, and the seismic response tests for the isolation model structure were performed using the 30 ton shaking table. (author). 369 refs., 119 tabs., 320 figs.

  3. Liquid metal reactor development. Development of LMR design technology

    International Nuclear Information System (INIS)

    Kim, Young Cheol; Kim, Y. I.; Kim, Y. G.; Kim, E. K.; Song, H.; Chung, H. T.; Sim, Y. S.; Min, B. T.; Kim, Y. S.; Wi, M. H.; Yoo, B.; Lee, J. H.; Lee, H. Y.; Kim, J. B.; Koo, G. H.; Hahn, D. H.; Na, B. C.; Hwang, W.; Nam, C.; Ryu, W. S.; Lim, G. S.; Kim, D. H.; Kim, J. D.; Gil, C. S.

    1997-07-01

    This project was performed in five parts, the scope and contents of which are as follows: The nuclear data processing system was established and the KFS group constant library was improved and verified. Basic computation system was constructed by either developing or adding its function. Input/output (I/O) interface processing was developed to establish an integrated calculation system for LMR core nuclear rand thermal-hydraulic design and analysis. An experimental data analysis was performed to validate the constructed core neutronic calculation system. Using the established core calculation system and design technology, preliminary core design and performance analysis on the domestic LMR core design concept were carried out. To develop the basic technology of the LMR system analysis, LMR system behavior characteristics evaluation, thermal -fluid system analysis in the reactor pool, preliminary overall plant analysis and computer codes development have been performed. A porous model and simple one-dimensional model have been evaluated for the reactor pool analysis. The evaluation of the residual heat removal system on different design concepts has been also conducted. For the development of high temperature structural analysis, the heat transfer and thermal stress analyses were performed using finite element program with user subroutine that has been developed with an implementation of the Chaboche constitutive model for inelastic analysis capability, and the evaluation of creep-fatigue and ratcheting behavior of high temperature structure was carried out using this program. for development of the seismic isolation system and to predict the shear behavior for the laminated rubber bearing were established. And the behavior tests of isolation bearing and rubber specimens were carried out, and the seismic response tests for the isolation model structure were performed using the 30 ton shaking table. (author). 369 refs., 119 tabs., 320 figs

  4. Decay heat removal for the liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Zemanick, P.P.; Brown, N.W.

    1975-01-01

    The functional and reliability requirements of the decay heat removal systems are described. The reliability requirement and its rationale as adequate assurance that public health and safety are safeguarded are presented. The means by which the reliability of the decay heat removal systems are established to meet their requirement are identified. The heat removal systems and their operating characteristics are described. The discussion includes the overflow heat removal service and its role in decay heat removal if needed. The details of the systems are described to demonstrate the elements of redundancy and diversity in the systems design. The quantitative reliability assessment is presented, including the reliability model, the most important assumptions on which the analysis is based, sources of failure data, and the preliminary numerical results. Finally, the qualitative analyses and administrative controls will be discussed which ensure reliability attainment in design, fabrication, and operation, including minimization of common mode failures. A component test program is planned to provide reliability data on selected critical heat removal system equipment. This test plan is described including a definition of the test parameters of greatest interest and the motivation for the test article selection. A long range plan is also in place to collect plant operational data and the broad outlines of this plan are described. A statement of the high reliability of the Clinch River Breeder reactor Plant decay heat removal systems and a summary of the supporting arguments is presented. (U.S.)

  5. Decay Heat Removal for the Liquid Metal Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zemanick, P. P.; Brown, N. W.

    1975-10-15

    The functional and reliability requirements of the decay heat removal systems are described. The reliability requirement and its rationale as adequate assurance that public health and safety are safeguarded are presented. The means by which the reliability of the decay heat removal systems are established to meet their requirement are identified. The heat removal systems and their operating characteristics are described. The discussion includes the overflow heat removal service and its role in decay heat removal if needed. The details of the systems are described to demonstrate the elements of redundancy and diversity in the systems design. The quantitative reliability assessment is presented, including the reliability model, the most important assumptions on which the analysis is based, sources of failure data, and the preliminary numerical results. Finally, the qualitative analyses and administrative controls will be discussed which ensure reliability attainment in design, fabrication, and operation, including minimization of common mode failures. A component test program is planned to provide reliability data on selected critical heat removal system equipment. This test plan is described including a definition of the test parameters of greatest interest and the motivation for the test article selection. A long range plan is also in place to collect plant operational data and the broad outlines of this plan are described. The paper closes with a statement of the high reliability of the Clinch River Breeder Reactor Plant decay heat removal systems and a summary of the supporting arguments. (author)

  6. Thermal convection of liquid metal in the titanium reduction reactor

    Science.gov (United States)

    Teimurazov, A.; Frick, P.; Stefani, F.

    2017-06-01

    The structure of the convective flow of molten magnesium in a metallothermic titanium reduction reactor has been studied numerically in a three-dimensional non-stationary formulation with conjugated heat transfer between liquid magnesium and solids (steel walls of the cavity and titanium block). A nonuniform computational mesh with a total of 3.7 million grid points was used. The Large Eddy Simulation technique was applied to take into account the turbulence in the liquid phase. The instantaneous and average characteristics of the process and the velocity and temperature pulsation fields are analyzed. The simulations have been performed for three specific heating regimes: with furnace heaters operating at full power, with furnace heaters switched on at the bottom of the vessel only, and with switched-off furnace heaters. It is shown that the localization of the cooling zone can completely reorganize the structure of the large-scale flow. Therefore, by changing heating regimes, it is possible to influence the flow structure for the purpose of creating the most favorable conditions for the reaction. It is also shown that the presence of the titanium block strongly affects the flow structure.

  7. Component and Technology Development for Advanced Liquid Metal Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Mark [Univ. of Wisconsin, Madison, WI (United States)

    2017-01-30

    The following report details the significant developments to Sodium Fast Reactor (SFR) technologies made throughout the course of this funding. This report will begin with an overview of the sodium loop and the improvements made over the course of this research to make it a more advanced and capable facility. These improvements have much to do with oxygen control and diagnostics. Thus a detailed report of advancements with respect to the cold trap, plugging meter, vanadium equilibration loop, and electrochemical oxygen sensor is included. Further analysis of the university’s moving magnet pump was performed and included in a section of this report. A continuous electrical resistance based level sensor was built and tested in the sodium with favorable results. Materials testing was done on diffusion bonded samples of metal and the results are presented here as well. A significant portion of this work went into the development of optical fiber temperature sensors which could be deployed in an SFR environment. Thus, a section of this report presents the work done to develop an encapsulation method for these fibers inside of a stainless steel capillary tube. High temperature testing was then done on the optical fiber ex situ in a furnace. Thermal response time was also explored with the optical fiber temperature sensors. Finally these optical fibers were deployed successfully in a sodium environment for data acquisition. As a test of the sodium deployable optical fiber temperature sensors they were installed in a sub-loop of the sodium facility which was constructed to promote the thermal striping effect in sodium. The optical fibers performed exceptionally well, yielding unprecedented 2 dimensional temperature profiles with good temporal resolution. Finally, this thermal striping loop was used to perform cross correlation velocimetry successfully over a wide range of flow rates.

  8. Gap and impact of LMR [Liquid Metal Reactor] piping systems and reactor components

    International Nuclear Information System (INIS)

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

    1987-01-01

    Because of high operation temperature, the LMR (Liquid Metal Reactor) plant is characterized by the thin-walled piping and components. Gaps are often present to allow free thermal expansion during normal plant operation. Under dynamic loadings, such as seismic excitation, if the relative displacement between the components exceeds the gap distance, impacts will occur. Since the components and piping become brittle over their design lifetime, impact is of important concern for it may lead to fractures of components and other serious effects. This paper deals with gap and impact problems in the LMR reactor components and piping systems. Emphasis is on the impacts due to seismic motion. Eight sections are contained in this paper. The gap and impact problems in LMR piping systems are described and a parametric study is performed on the effects of gap-induced support nonlinearity on the dynamics characteristics of the LMR piping systems. Gap and impact problems in the LMR reactor components are identified and their mathematical models are illustrated, and the gap and impact problems in the seismic reactor scram are discussed. The mathematical treatments of various impact models are also described. The uncertainties in the current seismic impact analyses of LMR components and structures are presented. An impact test on a 1/10-scale LMR thermal liner is described. The test results indicated that several clusters of natural modes can be excited by the impact force. The frequency content of the excited modes depends on the duration of the impact force; the shorter the duration, the higher the frequency content

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

    Energy Technology Data Exchange (ETDEWEB)

    Poddubnyi, I. I., E-mail: poddubnyyii@nikiet.ru [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, V. G.; Sviridov, E. V. [Russian Academy of Science, Joint Institute of High Temperatures (Russian Federation); Leshukov, A. Yu. [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Aleskovskiy, K. V. [National Research University Moscow Power Engineering Institute (MPEI) (Russian Federation); Obukhov, D. M. [Joint Stock Company Efremov Institute of Electrophysical Apparatus (Russian Federation)

    2016-12-15

    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

  10. Remote Inspection Techniques for Reactor Internals of Liquid Metal Reactor by using Ultrasonic Waveguide Sensor

    International Nuclear Information System (INIS)

    Joo, Young Sang; Kim, Seok Hun; Lee, Jae Han

    2006-02-01

    The primary components such as a reactor core, heat exchangers, pumps and internal structures of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection and continuous monitoring as major in-service inspection (ISI) methods of reactor internal structures. Reactor core and internal structures of LMR can not be visually examined due to an opaque liquid sodium. The under-sodium viewing and remote inspection techniques by using an ultrasonic wave should be applied for the in-service inspection of reactor internals. The remote inspection techniques using ultrasonic wave have been developed and applied for the visualization and ISI of reactor internals. The under sodium viewing technique has a limitation for the application of LMR due to the high temperature and irradiation environment. In this study, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium viewing and remote inspection. The Lamb wave propagation of a waveguide sensor has been analyzed and the zero-order antisymmetric A 0 plate wave was selected as the application mode of the sensor. The A 0 plate wave can be propagated in the dispersive low frequency range by using a liquid wedge clamped to the waveguide. A new technique is presented which is capable of steering the radiation beam angle of a waveguide sensor without a mechanical movement of the sensor assembly. The steering function of the ultrasonic radiation beam can be achieved by a frequency tuning method of the excitation pulse in the dispersive range of the A 0 mode. The technique provides an opportunity to overcome the scanning limitation of a waveguide sensor. The beam steering function has been evaluated by an experimental verification. The ultrasonic C-scanning experiments are performed in water and the feasibility of the ultrasonic waveguide sensor has been verified. The various remote inspection

  11. Liquid metal versus gas cooled reactor concepts for a turbo electric powered space vehicle

    International Nuclear Information System (INIS)

    Carre, F.; Proust, E.; Schwartz, J.P.

    1985-01-01

    Recent CNES/CEA prospective studies of an orbit transfer vehicule to be launched by ARIANE V, emphasize the advantage of the Brayton cycle over the thermionics and thermoelectricity, in minimizing the total mass of 100 to 300 kWsub(e) power systems under the constraint specific to ARIANE of a radiator area limited to 95 m 2 . The review of candidate reactor concepts for this application, finally recommends both liquid metal and gas cooled reactors, for their satisfactory adaptation to a reference Brayton cycle and for the available experience from the terrestrial operation of comparable systems

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

    International Nuclear Information System (INIS)

    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

  13. Physics considerations in the design of liquid metal reactors for transuranium element consumption

    International Nuclear Information System (INIS)

    Khalil, H.; Hill, R.; Fujita, E.; Wade, D.

    1992-01-01

    The management of transuranic nuclides in liquid metal reactors (LMR's) is considered based on the use of the Integral Fast Reactor (IFR) concept. Unique features of the IFR fuel cycle with respect to transuranic management are identified. These features are exploited together with the hard spectrum of LMR's to demonstrate the neutronic feasibility of a wide range of transuranic management options ranging from efficient breeding to pure consumption. Core physics aspects of the development of a low sodium void worth transuranic burner concept are described. Neutronics performance parameters and reactivity feedback characteristics estimated for this core concept are presented

  14. Technology development program for safe shipment of spent fuel from liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Freedman, J.M.; Humphreys, J.R.

    1975-10-01

    A comprehensive plan to develop shipping cask technology is described. Technical programs in the disciplines of heat transfer, structures and containment, spent fuel characterization, hot laboratory verification, shielding, and hazards analysis are discussed. Both short- and long-term goals in each discipline are delineated and how the disciplines interrelate is shown. The technologies developed will be used in the design, fabrication, and testing of truck-mounted and rail-car casks. These casks will be used for safely transporting short-cooled, high-burnup Liquid Metal Fast Breeder Reactor (LMFBR) spent fuel from reactors to reprocessing plants

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

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

    2009-01-01

    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.

  17. Thermal bonding of light water reactor fuel using nonalkaline liquid-metal alloy

    International Nuclear Information System (INIS)

    Wright, R.F.; Tulenko, J.S.; Schoessow, G.J.; Connell, R.G. Jr.; Dubecky, M.A.; Adams, T.

    1996-01-01

    Light water reactor (LWR) fuel performance is limited by thermal and mechanical constraints associated with the design, fabrication, and operation of fuel in a nuclear reactor. A technique is explored that extends fuel performance by thermally bonding LWR fuel with a nonalkaline liquid-metal alloy. Current LWR fuel rod designs consist of enriched uranium oxide fuel pellets enclosed in a zirconium alloy cylindrical clad. The space between the pellets and the clad is filled by an inert gas. Because of the low thermal conductivity of the gas, the gas space thermally insulates the fuel pellets from the reactor coolant outside the fuel rod, elevating the fuel temperatures. Filling the gap between the fuel and clad with a high-conductivity liquid metal thermally bonds the fuel to the cladding and eliminates the large temperature change across the gap while preserving the expansion and pellet-loading capabilities. The application of liquid-bonding techniques to LWR fuel is explored to increase LWR fuel performance and safety. A modified version of the ESCORE fuel performance code (ESBOND) is developed to analyze the in-reactor performance of the liquid-metal-bonded fuel. An assessment of the technical feasibility of this concept for LWR fuel is presented, including the results of research into materials compatibility testing and the predicted lifetime performance of liquid-bonded LWR fuel. The results show that liquid-bonded boiling water reactor peak fuel temperatures are 400 F lower at beginning of life and 200 F lower at end of life compared with conventional fuel

  18. Potential of multi-purpose liquid metallic fuelled fast reactor (MPFR) as a hydrogen production system

    International Nuclear Information System (INIS)

    Endo, H.; Ninokata, H.; Netchaev, A.; Sawada, T.

    2001-01-01

    Nuclear energy is the only effective alternative energy source to fossil fuels in the next century. Therefore future nuclear power plants should satisfy the following three requirements: i) multiple energy conversion capability with high temperature not only for electricity generation but also for hydrogen production, ii) extended siting capability so as to eliminate on-site refuelling, and iii) passive safety features. An aim of this paper is to describe the basic concept of the multi-purpose liquid metallic fuelled fast reactor system (MPFR). The MPFR introduces the U-Pu-X (X: Mn, Fe, Co) liquid metallic alloy with Ta and Ta/TaC structural materials, and satisfies all of the conditions listed above based on the following characteristics of the liquid metallic fuel: high temperature operation between 650 deg C (sodium-cooled system) and 1 200 deg C (lead-cooled system), a core lifetime of 15-30 years without radiation damage of fuel materials, and enhanced passive safety by the thermal expansion of liquid fuel and the avoidance of re-criticality due to local core fuel dispersion at fuel failure events. (authors)

  19. Study of MHD problems in liquid metal blankets of fusion reactors

    International Nuclear Information System (INIS)

    Michael, I.

    1984-12-01

    This study describes in a concise form the state of knowledge regarding MHD problems to be expected in case of use of liquid metal in the blankets of fusion reactors with magnetic confinement. MHD pressure losses and MHD friction coefficients in the straight channel, in bent sections and in case of variation of the channel cross section play a major role because the high MHD flow resistances call for high pumping powers. Influencing the velocity profile transverse to the main flow direction of the liquid metal by application of an external, strong magnetic field bears consequences on the release and transport of corrosion products in the liquid metal circuit and on the heat transfer. Possibilities of reducing the MHD effects are discussed. However, it becomes obvious that an account of the lack of experimental results there are still major gaps in the knowledge of MHD effects occurring in strong magnetic fields. These gaps can be greatly reduced by implementation of an experimental program as proposed in this report. (orig.) [de

  20. Comparative sodium void effects for different advanced liquid metal reactor fuel and core designs

    International Nuclear Information System (INIS)

    Dobbin, K.D.; Kessler, S.F.; Nelson, J.V.; Gedeon, S.R.; Omberg, R.P.

    1991-01-01

    An analysis of metal-, oxide-, and nitride-fueled advanced liquid metal reactor cores was performed to investigate the calculated differences in sodium void reactivity, and to determine the relationship between sodium void reactivity and burnup reactivity swing using the three fuel types. The results of this analysis indicate that nitride fuel has the least positive sodium void reactivity for any given burnup reactivity swing. Thus, it appears that a good design compromise between transient overpower and loss of flow response is obtained using nitride fuel. Additional studies were made to understand these and other nitride advantages. (author)

  1. Utility industry evaluation of the metal fuel facility and metal fuel performance for liquid metal reactors

    International Nuclear Information System (INIS)

    Burstein, S.; Gibbons, J.P.; High, M.D.; O'Boyle, D.R.; Pickens, T.A.; Pilmer, D.F.; Tomonto, J.R.; Weinberg, C.J.

    1990-02-01

    A team of utility industry representatives evaluated the liquid metal reactor metal fuel process and facility conceptual design being developed by Argonne National Laboratory (ANL) under Department of Energy sponsorship. The utility team concluded that a highly competent ANL team was making impressive progress in developing high performance advanced metal fuel and an economic processing and fabrication technology. The utility team concluded that the potential benefits of advanced metal fuel justified the development program, but that, at this early stage, there are considerable uncertainties in predicting the net overall economic benefit of metal fuel. Specific comments and recommendations are provided as a contribution towards enhancing the development program. 6 refs

  2. An overview of FFTF [Fast Flux Test Facility] contributions to Liquid Metal Reactor Safety

    International Nuclear Information System (INIS)

    Waltar, A.E.; Padilla, A. Jr.

    1990-11-01

    The Fast Flux Test Facility has provided a very useful framework for testing the advances in Liquid Metal Reactor Safety Technology. During the licensing phase, the switch from a nonmechanistic bounding technique to the mechanistic approach was developed and implemented. During the operational phase, the consideration of new tests and core configurations led to use of the anticipated-transients-without-scram approach for beyond design basis events and the move towards passive safety. The future role of the Fast Flux Test Facility may involve additional passive safety and waste transmutation tests. 26 refs

  3. Secondary coolant circuit for liquid-metal cooled reactor and steam generator for such a circuit

    International Nuclear Information System (INIS)

    Brachet, A.; Figuet, J.; Guidez, J.; Lions, N.; Traiteur, R.; Zuber, T.

    1984-01-01

    An upper buffer tank and downstream buffer tank are disposed inside the steam generators. The downstream briffer tank is annular and it surrounds and communicates with a zone of the steam generator through which the liquid metal flows towards the bottom between the exchange zone and the outlet nozzle. The pressure of the inert gas blanket in the downstream buffer volume is more important than this one in the upper buffer volume. The invention applies to fast neutron nuclear reactor cooled by sodium [fr

  4. Conceptual study of a complementary scram system for liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Vanmaercke, S.; Van den Eynde, G.; Tijskens, E.; Bartosiewicz, Y.

    2009-01-01

    GEN-IV reactors promise higher safety and reliability as one of the major improvements over previous generations of reactors. To achieve that, all GEN-IV reactor concepts require two completely independent shutdown systems that rely on different operating principles. For liquid metal cooled reactors the first system is an absorber-rod based solution. The second system that by requirement should rely on another principle, is however quite a challenge to design. The second system used in current PWR reactors is to dissolve a neutron absorber, boric acid, into the primary coolant. This method cannot be used in liquid metal cooled reactors because of the high cost of cleaning the coolant. In this paper an overview of the existing literature on scram systems is given, each with their advantages and limitations. A promising new concept is also presented. This concept leads to a totally passive self activating device using small absorbing particles that flow into a dedicated channel to shutdown the reactor. The system consists of tubes filled with particles of an absorber material. During normal operation, these particles are kept above the active core by means of a metallic seal. In case of an accident, the system is activated by the temperature increase in the coolant. This leads to melting of the metal seal. The ongoing work conducted at SCK·CEN and UCL/TERM aims at assessing the reliability of this new concept both experimentally and numerically. This study is multidisciplinary as neutronic and thermal hydraulics issues are tackled. Most challenging is however the thermal hydraulics related to understanding and predicting the liberation and flow of the absorber particles during a shutdown. Simple experiments are envisaged to compare to numerical simulations using the Discrete Element Method for simulating the particles. In a later stage this will be coupled with Smoothed Particles Hydrodynamics for simulating the melting of the seal. Some preliminary experimental and

  5. Development of a risk-based inservice inspection program for a liquid metal reactor

    International Nuclear Information System (INIS)

    King, R.W.; Buschman, H.W.

    1996-01-01

    The emerging application of risk-based assessment technology to the operation and maintenance of nuclear power plants holds considerable promise for improving efficiency and reducing operating costs. EBR-II is liquid-metal-cooled fast reactor which operated for thirty years before shutting down in September 1994 due to program termination. Prior to the shutdown of EBR-II, an in-service inspection (ISI) program was developed that exploited certain advantages of the liquid-metal reactor design, e.g., demonstrated passive response to plant upset events, low pressure primary coolant and compatibility of the coolant and reactor materials. Many of the systems cannot be inspected due to inaccessibility of the components. However, application of a risk-based approach provided the basis for reducing or eliminating inspections in some areas that would otherwise be required. Development and implementation of the risk-based ISI program was interrupted by the DOE-mandated shutdown of EBR-II, so the potential benefits of this approach in terms of reduced O and M costs have yet to be realized. Through the development of this program, however it is clear that there is potential for substantial cost-savings while improving the risk-profile of the facility through this approach

  6. An overview of the U.S. Department of Energy's program for liquid metal reactor seismic technology

    International Nuclear Information System (INIS)

    Jetter, R.I.; Seidensticker, R.W.

    1988-01-01

    During the past decade, the U.S. Department of Energy (DOE) has sponsored the development of seismic design technology in support of Liquid Metal Reactors (LMR's). This has been accomplished through 1) major projects such as the Fast Flux Test Facility (FFTF) and the Clinch River Breeder Reactor (CRBR), 2) base technology programs and 3) support to the design development of innovative LMR's, SAFR and PRISM. These developments have come in the areas of ground motion definition, soil-structure interaction, seismic isolation, fluid-structure interaction and structural analysis methods and criteria for equipment and components such as piping, reactor core and vessels. The initial developments in seismic design technology by DOE and others were directed toward ensuring that the plant, equipment and components had sufficient seismic resistance to ensure availability after an Operations Basis Earthquake (OBE) and to survive a Safe Shutdown Earthquake (SSE). During this period, the emphasis on conservative design had significant cost impacts. The current focus is directed toward a better understanding of seismic design margins and the development of methods to reduce seismic loads on plant and equipment and to enhance siting flexibility. From this perspective, the DOE is currently reassessing the needs and priorities for future seismic technology development. Coordination with University research programs and ongoing seismic technology development sponsored by other governmental agencies and institutions is an integral part of this planning process. The purpose of this paper is to highlight the current status of DOE's seismic technology program for LMR's and to provide an overview of future areas of interest. (author). 7 refs

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    Jones, R.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

  9. Systems analysis of a 100-MWe modular liquid metal cooled reactor

    International Nuclear Information System (INIS)

    Morris, E.E.; Rhow, S.K.; Switick, D.M.

    1985-01-01

    The response of a 100-MWe modular liquid metal cooled reactor to unprotected loss of flow and/or loss of primary heat removal accidents is analyzed using the systems analysis code SASSYS. The reactor response is tracked for the first 1000 s following a postulated upset in the primary heat removal system. The calculations do not take credit for the functioning of any decay heat removal other than through the secondary system. In addition to the power rating, other features of the reactor are an average sodium temperature rise of 148 K, a sodium void worth (counting the core and upper axial blanket) of 1.89 $, and 3.6 $ of Doppler feedback due to a uniform e-fold fuel temperature increase

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

    International Nuclear Information System (INIS)

    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

  11. Research towards ultrasonic systems to assist in-vessel manipulations in liquid metal cooled reactors

    International Nuclear Information System (INIS)

    Dierckx, Marc; Van-Dyck, Dries

    2013-06-01

    We describe the state of the art of the research towards ultrasonic measurement methods for use in lead-bismuth cooled liquid metal reactors. Our current research activities are highly focused on specific tasks in the MYRRHA system, which is a fast spectrum research reactor cooled with the eutectic mixture of lead and bismuth (LBE) and is conceived as an accelerator driven system capable of operating in both sub-critical and critical mode. As liquid metal is opaque to light, normal visual feedback during fuel manipulations in the reactor vessel is not available and must therefore be replaced by a system that is not hindered by the opacity of the coolant. In this respect ultrasonic measurement techniques have been proposed and even developed in the past for operation in sodium cooled reactors. To our knowledge, no such systems have ever been deployed in lead based reactors and we are the first to have a research program in this direction as will be detailed in this paper. We give an overview of the acoustic properties of LBE and compare them with the properties of sodium and water to theoretically show the feasibility of ultrasonic systems operating in LBE. In the second part of the paper we discuss the results of the validation experiments in water and LBE. A typical scene is ultrasonically probed by a mechanical scanning system while the signals are processed to render a 3D visualization on a computer screen. It will become clear that mechanical scanning is capable of producing acceptable images but that it is a time consuming process that is not fit to solve the initial task to providing feedback during manipulations in the reactor vessel. That is why we propose to use several dedicated ultrasonic systems each adapted to a specific task and capable to provide real-time feedback of the ongoing manipulations, as is detailed in the third and final part of the paper. (authors)

  12. Design of self-cooled, liquid-metal blankets for tokamak and tandem mirror reactors

    International Nuclear Information System (INIS)

    Cha, Y.S.; Gohar, Y.; Hassanein, A.M.; Majumdar, S.; Picologlou, B.F.; Smith, D.L.; Szo, D.K.

    1985-01-01

    Results of the self-cooled, liquid-metal blanket design from the Blanket Comparison and Selection Study (BCSS) are summarized. The objectives of the BCSS project are to define a small number (about three) of blanket concepts that should be the focus of the blanket research and development (RandD) program, identify and prioritize the critical issues for the leading blanket concepts, and provide technical input necessary to develop a blanket RandD program plan. Two liquid metals (lithium and lithium-lead (17Li-83Pb)) and three structural materials (primary candidate alloy (PCA), ferritic steel (FS) (HT-9), and vanadium alloy (V-15 Cr-5 Ti)) are included in the evaluations for both tokamaks and tandem mirror reactors (TMRs). TMR is of the tube configuration similar to the Mirror Advanced Reactor Study design. Analyses were performed in the following generic areas for each blanket concept: MHD, thermal hydraulics, stress, neutronics, and tritium recovery. Integral analyses were performed to determine the design window for each blanket design. The Li/Li/V blanket for tokamak and the Li/Li/V, LiPb/LiPb/V, and Li/Li/HT-9 blankets for the TMR are judged to be top-rated concepts. Because of its better thermophysical properties and more uniform nuclear heating profile, liquid lithium is a better coolant than liquid 17Li83Pb. From an engineering point of view, vanadium alloy is a better structural material than either FS or PCA since the former has both a higher allowable structural temperature and a higher allowable coolant/structure interface temperature than the latter. Critical feasibility issues and design constraints for the self-cooled, liquid-metal blanket concepts are identified and discussed

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

    Science.gov (United States)

    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. Centralized Reliability Data Organization (CREDO) assessment of critical component unavailability in liquid metal reactors

    International Nuclear Information System (INIS)

    Koger, K.H.; Haire, M.J.; Humphrys, B.L.; Manneschmidt, J.F.; Setoguchi, K.; Nakai, R.

    1988-01-01

    The Centralized Reliability Data Organization (CREDO) is the largest repository of liquid metal reactor (LMR) component reliability data in the world. It is jointly sponsored by the US Dept. of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. The CREDO data base contains information on a population of more than 20,000 components and approximately 1500 event records. A conservative estimation is that the total component operating hours is approaching 2.2 billion hours. The work reported here focuses on the availability information contained in CREDO and the development of availability critical items lists. That is, individual components are ranked in prioritized lists from worst to best performers from an availability standpoint. Availability as used here is an inherent characteristics of the component and is not necessarily related to plant operability. A major observation is that a few components have a much higher unavailability factor than the average. The top fifteen components contribute 93%, 77%, and 87% of the total system unavailability for EBR-II, FFTF, and JOYO respectively. Critical components common to all three sites are mechanical pumps and electromagnetic pumps. Application of resources to these components with the highest unavailability will have the greatest effect on overall availability. All three sites demonstrate that low maintainability (i.e., long repair times), rather than unreliability (i.e., high failure rates), are the main contributors, by about a two-to-one margin, to liquid metal system unavailability

  15. Status of liquid metal reactor development in the United States of America

    International Nuclear Information System (INIS)

    Griffith, J.D.; Horton, K.E.

    1989-01-01

    The United States has made substantial progress in achieving LMR programme objectives. A decision was made in 1988 to select the General Electric ALMR concept known as PRISM (Power Reactor Innovative Safe Module) for advanced conceptual design. A 3-year contract was awarded to General Electric in January of this year for concentrated trade-off studies and advanced design development. The strategy is to integrate those advancements that best meet programme objectives into a national ALMR system concept. (author). 8 figs

  16. The role of structural integrity in liquid metal fast breeder reactor safety

    International Nuclear Information System (INIS)

    Holmes, J.A.G.

    1982-01-01

    Extensive studies have demonstrated the favourable safety characteristics of liquid metal fast breeder reactors, which are attributable to both their inherent features and the engineered safeguards which are included. This requires demonstration that there is no risk of sudden catastrophic failure of the core support system allowing the core to drop off the control rods to give a prompt critical reactivity excursion. An important part of our work in support of the safety case for the U.K. Commercial Demonstration Fast Reactor is to demonstrate that such a failure is virtually incredible. This covers design features, study of the fracture behaviour of stainless steel structures, and inspection and monitoring during fabrication and service. The paper gives a broad description of the relevant design features and supporting work programme

  17. Thermal performance of fresh mixed-oxide fuel in a fast flux LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Ethridge, J.L.; Baker, R.B.

    1985-01-01

    A test was designed and irradiated to provide power-to-melt (heat generation rate necessary to initiate centerline fuel melting) data for fresh mixed-oxide UO 2 -PuO 2 fuel irradiated in a fast neutron flux under prototypic liquid metal reactor (LMR) conditions. The fuel pin parameters were selected to envelope allowable fabrication ranges and address mass production of LMR fuel using sintered-to-size techniques. The test included fuel pins with variations in fabrication technique, pellet density, fuel-to-cladding gap, Pu concentration, and fuel oxygen-to-metal ratios. The resulting data base has reestablished the expected power-to-melt in mixed-oxide fuels during initial reactor startup when the fuel temperatures are expected to be the highest. Calibration of heat transfer models of fuel pin performance codes with these data are providing more accurate capability for predicting steady-state thermal behavior of current and future mixed-oxide LMR fuels

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

    International Nuclear Information System (INIS)

    Cahalan, J.; Wigeland, R.; Friedel, G.; 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

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

    International Nuclear Information System (INIS)

    Gillett, J.E.; Wineman, A.L.

    1983-01-01

    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 each have a pin which rides 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 the locking plates into engagement with the housing in a raised or a lowered position of the support column such that the support column is then supported by the locking plate and not by the ball screw jacks. (author)

  20. The neutronics design and analysis of a liquid metal reactor for burning minor actinides

    International Nuclear Information System (INIS)

    Choi, H.B.; Downar, T.J.

    1992-01-01

    A liquid metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors (LWR). The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the Doppler coefficient, and the sodium void worth. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200 MWth core is able to transmute the annual minor actinide inventory of about 26 LWRs and still exhibit reasonable safety characteristics. Sensitivity analysis of the final core design indicates deficiencies in the minor actinide nuclear data can introduce large uncertainties in the prediction of the core safety performance parameters

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

    International Nuclear Information System (INIS)

    Hunsbedt, A.; Boardman, C.E.

    1993-01-01

    A dual passive cooling system for liquid metal cooled nuclear fission reactors is described, comprising the combination of: a reactor vessel for containing a pool of liquid metal coolant with a core of heat generating fissionable fuel substantially submerged therein, a side wall of the reactor vessel forming an innermost first partition; a containment vessel substantially surrounding the reactor vessel in spaced apart relation having a side wall forming a second partition; a first baffle cylinder substantially encircling the containment vessel in spaced apart relation having an encircling wall forming a third partition; a guard vessel substantially surrounding the containment vessel and first baffle cylinder in spaced apart relation having a side wall forming a forth partition; a sliding seal at the top of the guard vessel edge to isolate the dual cooling system air streams; a second baffle cylinder substantially encircling the guard vessel in spaced part relationship having an encircling wan forming a fifth partition; a concrete silo substantially surrounding the guard vessel and the second baffle cylinder in spaced apart relation providing a sixth partition; a first fluid coolant circulating flow course open to the ambient atmosphere for circulating air coolant comprising at lent one down comer duct having an opening to the atmosphere in an upper area thereof and making fluid communication with the space between the guard vessel and the first baffle cylinder and at least one riser duct having an opening to the atmosphere in the upper area thereof and making fluid communication with the space between the first baffle cylinder and the containment vessel whereby cooling fluid air can flow from the atmosphere down through the down comer duct and space between the forth and third partitions and up through the space between the third and second partition and the riser duct then out into the atmosphere; and a second fluid coolant circulating flow

  2. Inter-subchannel heat transfer modeling for a subchannel analysis of liquid metal-cooled reactors

    International Nuclear Information System (INIS)

    Hae-Yong, Jeong; Kwi-Seok, Ha; Young-Min, Kwon; Yong-Bum, Lee; Dohee, Hahn

    2007-01-01

    In a subchannel approach, the temperature, pressure and velocity in a subchannel are averaged, and one representative thermal-hydraulic condition specifies the state of a subchannel. To enhance the predictability of a subchannel analysis code, it is required to model the inter-subchannel heat transfer between the adjacent subchannels as accurately as possible. One of the critical parameters which determine the thermal-hydraulic behavior of the coolant in subchannels is the heat conduction between two neighboring sub-channels. This portion of a heat transfer becomes more important in the design of an LMR (Liquid Metal-cooled Reactor) because of the high heat capacity of the liquid metal coolant. The other important part of heat transfer is the mixing of flow as a form of cross flow. Especially, the turbulent mixing caused by the eddy motion of fluid across the gap between the subchannels enhances the exchange of the momentum and the energy through the gap with no net transport of the mass. Major results of recent efforts on these modeling have been implemented in a subchannel analysis code MATRA-LMR-FB. The analysis shows that the accuracy of a subchannel analysis code is improved by enhancing the models describing the conduction heat transfer and the cross-flow mixing, especially at low flow rate. (authors)

  3. Seismic isolation development for the US advanced liquid-metal reactor program

    International Nuclear Information System (INIS)

    Gluekler, E.L.; Bigelow, C.C.; DeVita, V.; Kelly, J.M.; Seidensticker, R.W.; Tajirian, F.F.

    1989-01-01

    GE Nuclear Energy, in association with a US Industrial Team and support from the US National Laboratories and Universities, is developing a modular liquid-metal reactor concept for the US Department of Energy (DOE). The objective of this development is to provide, by the turn of the century, a reactor concept with optimized passive safety features that is economically competitive with other domestic energy sources, licensable, and ready for commercial deployment. One of the unique features of the concept is the seismic isolation of the reactor modules which decouples the reactor and their safety systems from potentially damaging ground motions and significantly enhances the structural resistance to high energy, as well as long duration earthquakes. Seismic isolation is accomplished with high damping natural rubber bearings. The reactors are located in individual silos below grade level and are supported by the isolator bearings at approximately their center of gravity. This application of seismic isolation is the first for a US nuclear power plant. A development program has been established to assure the full benefits from the utilization of this new approach and to provide adequate system characterization and qualification for licensing certification. The development program is described in this paper and selected results are presented. The initial testing indicated excellent performance of high damping natural rubber bearings

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

    International Nuclear Information System (INIS)

    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

  5. Seismic analysis of a large pool-type LMR [liquid metal reactor

    International Nuclear Information System (INIS)

    Wang, C.Y.; Gvildys, J.

    1989-01-01

    This paper describes the seismic study of a 450-MWe liquid metal reactor (LMR) under 0.3-g SSE ground excitation. Two calculations were performed using the new design configuration. They deal with the seismic response of the reactor vessel, the guard vessel and support skirt, respectively. In both calculations, the stress and displacement fields at important locations of those components are investigated. Assessments are also made on the elastic and inelastic structural capabilities for other beyond-design basis seismic loads. Results of the reactor vessel analysis reveal that the maximum equivalent stress is only about half of the material yield stress. For the guard vessel and support skirt, the stress level is very small. Regarding the analysis if inelastic structural capability, solutions of the Newmark-Hall ductility modification method show that the reactor vessel can withstand seismics with ground ZPAs ranging from 1.015 to 1.31 g, which corresponds to 3.37 to 4.37 times the basic 0.3-g SSE. Thus, the reactor vessel and guard vessel are strong enough to resist seismic loads. 4 refs., 10 figs., 5 tabs

  6. A seismic performance and cost comparison of top and bottom supported liquid metal reactor vessels

    International Nuclear Information System (INIS)

    Carlson, T.M.; Kiciman, O.K.; Petrozelli, J.F.

    1989-01-01

    It is the premise of this paper that the revision of a pool LMR from a TSRV configuration to a specific bottom supported reactor vessel (BSRV) configuration can resolve the above TSRV disadvantages related to load path length and diversity, thereby improving seismic performance and simultaneously reducing RV block costs by reducing weights. This paper demonstrates this premise by comparing a reference TSRV block with a specific BSRV block design. Recent capital cost estimates ($/kWe) for U.S. liquid metal reactor (LMR) plant designs reveal that the balance of plant costs could be reduced below that of the balance of plant costs for a comparable light water reactor plant. However, in regions of high seismicity, non-seismically isolated LMR nuclear steam supply system weights are costs per kWe are two to three times the weights and costs of light water reactor nuclear steam supply systems. While all portions of the LMR nuclear steam supply system require examination for potential cost reductions, the focus of this paper is the reactor vessel (RV) block for a large pool plant

  7. Review of In-Service Inspection and Repair Technique Developments for French Liquid Metal Fast Reactors

    International Nuclear Information System (INIS)

    Baque, F.

    2005-01-01

    In-service monitoring of nuclear plants is indispensable for both the Operator and the Regulator. The notion of in-service monitoring ranges from the continuous monitoring of the reactor in operation to the thorough in-service reactor inspection during programmed shutdowns. However, the highly specific environment found in French liquid metal fast reactor plants - Phenix and Superphenix - makes monitoring and inspection complicated because of the use of a sodium coolant that is hot, opaque, and difficult to drain.The Commissariat a l'Energie Atomique, in collaboration with its traditional French partners, Electricite de France utilities and FRAMATOME/Novatome Engineering, decided to conduct a 6-yr research and development program (1994-2000) to explore this problem vis-a-vis Superphenix, as well as the possibilities of intervening within the reactor block or on components in a sodium environment. Furthermore, the safety reevaluation of Phenix, conducted between 1994 and 2003, represented an excellent 'test bench' during which the limits of inspection processes - applied to an integrated reactor concept - were surpassed using techniques such as fuel subassembly head scanning, ultrasonic examination of the core support, and visual inspection of the cover-gas plenum following a partial sodium draining. Repair techniques were investigated for cleaning of sodium wet structure surfaces, cutting of damaged parts, and welding in sodium aerosol atmosphere. Both conventional and laser processes were tested

  8. Fission and corrosion product behaviour in liquid metal fast breeder reactors (LMFBRs)

    International Nuclear Information System (INIS)

    1993-02-01

    It is intended that this review will be useful not only to scientists but also to those concerned with design, day-to-day operation of plant, with liquid metal fast breeder reactors (LMFBRs), safety and decommissioning. Because of this, the review has been widened to include not only the mass transfer behaviour of the various radionuclides in experimental and operating systems, but also the monitoring of the various species, the methods of measurement and the development of methods to control the build-up of the more important long half-life species in operating plants. The information used in the review has been taken from open literature sources to provide an up-to-date presentation of the behaviour of the various isotopes in LMFBRs. 172 refs, 14 figs, 22 tabs

  9. Development of an emergency air-cleaning system for liquid-metal reactors

    International Nuclear Information System (INIS)

    Owen, R.K.

    1980-11-01

    A novel air cleaning concept has been developed for potential use in venting future commercial liquid metal fast breeder reactor containment buildings in the unlikely event of postulated core disruptive accidents. The passive concept consists of a submerged gravel bed to collect the bulk of particulate contaminates carried by the vented gas. A fibrous scrubber could be combined with the submerged gravel scrubber to enhance collection efficiencies for the smaller sized particles. The submerged gravel scrubber is unique in that water flow through the packed bed is induced by the gas flow, eliminating the need for an active liquid pump. In addition, design gas velocities through the packed bed are 10 to 20 times higher than for a conventional sand bed filter

  10. A vanadium alloy for the application in a liquid metal blanket of a fusion reactor

    Science.gov (United States)

    Borgstedt, H. U.; Grundmann, M.; Konys, J.; Perić, Z.

    1988-07-01

    The vanadium alloy V3Ti1Si has been corrosion tested in liquid lithium and the eutectic alloy Pb-17Li at 550°C. This alloy has a comparable corrosion resistance to the alloy V15Cr5Ti in lithium. In this molten metal it is superior to stainless steel AISI 316. In the Pb-17Li melt it is even superior to martensitic steels. The alloy has only a weak tendency to be dissolved. It is sensitive to an exchange of non-metallic elements, which causes the formation of a hardened surface layer. These chemical effects are influenced by the mass and surface ratios of the vanadium alloy to the molten metals and other structural materials. These ratios are unfavorable in the two test loops. The effects might be less pronounced in a vanadium alloy/liquid metal fusion reactor blanket.

  11. Probabilities of inherent shutdown of unprotected events in innovative liquid metal reactors

    International Nuclear Information System (INIS)

    Mueller, C.J.; Wade, D.C.

    1988-01-01

    The uncertainty in predicting the effectiveness of inherent shutdown in innovative liquid metal cooled reactors with metallic fuel results from three broad contributing areas of uncertainty: (1) the inability to exactly predict the frequency of ATWS events with potential to challenge the safety systems and require inherent shutdown; (2) the approximation of representing all such events by a selected set of ''generic scenarios''; and (3) the inability to exactly calculate the core response to the selected generic scenarios. This paper discusses the work being done to address each of these contributing areas, identifies the design and research approaches being used at Argonne National Laboratory to reducing the key contributions to uncertainties in inherent shutdown, and presents results. The conditional probabilities (given ATWS initiation) of achieving temperatures capable of defeating inherent shutdown are shown to range from /approximately/0.1% to negligible for current designs

  12. Application of acoustic agglomerators for emergency use in liquid-metal fast breeder reactor plants

    International Nuclear Information System (INIS)

    Shaw, D.T.; Rajendran, N.

    1979-01-01

    The use of acoustic agglomerators for the suppression of sodium-fire aerosols in the case of a hypothetical core disruptive accident of a liquid-metal fast breeder reactor is discussed. The basic principle for the enhancement of agglomeration of airborne particles under the influence of an acoustic field is first discussed, followed by theoretical predictions of the optimum operating conditions for such application. It is found that with an acoustic intensity of 160 dB (approx. 1 W/cm 2 ), acoustic agglomeration is expected to be several hundred times more effective than gravitational agglomeration. For particles with a radius larger than approx. 2 μm, hydrodynamic interaction becomes more important than the inertial capture. For radii between 0.5 and 2 μm, both mechanisms have to included in the theoretical predictions of the acoustic agglomeration rate

  13. PRISM [Power Reactor Inherently Safe Module] design concept enhances waste management

    International Nuclear Information System (INIS)

    Thompson, M.L.; Berglund, R.C.

    1989-01-01

    PRISM, a modular advanced liquid metal reactor (ALMR), has been designed conceptually by GE under the US Department of Energy sponsorship. The concept design and analyses have been primarily focused on passive safety and improved construction and operating costs. Significantly, the unique design of multiple modules and features of PRISM enhance waste management over conventional reactor systems. This paper provides an overview of PRISM of these enhancements. Inherent to the ALMR's, the sodium coolant precludes crud buildup on reactor surfaces and in components and waste for disposal. Preliminary evaluations indicate this fundamental feature results in factors of 2-4 less waste volume and 2-3 orders of magnitude less curies per megawatt-electric for ultimate disposal. For example, the tap designed for sodium cleanup is expected to be exchanged only once every thirty years. Also, inherent to ALMR's, burning waste actinides and selected fission products to preclude their accumulation and burial is very attractive. The hard neutron spectrum of ALMR burns the actinides efficiently and is not poisoned by the actinides and fission products. The modular design of PRISM components (and the fuel cycle equipment) permit replacement without expensive and potentially hazardous volume reduction. For example, the functional components of the reference electromagnetic pump and IHK can be removed intact for waste disposal. Although development of the reference metal fuel is not completed, it is estimated that (low-level) waste from recycle of the fuel will result in significantly less volume than would be generated by aqueous recycle of oxide fuel. 6 refs., 10 figs

  14. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    International Nuclear Information System (INIS)

    Geraldo, Issa Cherif; Bose, Tanmoy; Pekpe, Komi Midzodzi; Cassar, Jean-Philippe; Mohanty, A.R.; Paumel, Kévin

    2014-01-01

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected

  15. An evaluation of the concept of transuranic burning using liquid metal reactors

    International Nuclear Information System (INIS)

    Rodwell, E.; Shaw, R.A.; Williams, R.F.

    1991-03-01

    The evaluation investigates the potential benefits to radioactive waste disposal of separating the transuranic elements from spent reactor fuel before the spent fuel is disposed of in geologic repositories. The evaluation addresses the question whether the benefits to radioactive waste disposal would justify processing the fuel to separate the transuranics, plus a liquid metal reactor (LMR) deployment program to transmute the separated transuranics. The evaluation concludes that adoption of a process-before-disposal policy for all the spent fuel from the light water reactors (LWRs) would accrue only modest benefits with respect to the accumulation of uranium mill tailings, the national inventory of transuranics and the licensing of a geologic repository. It is likely that this process-before-disposal policy would incur a large cost penalty, encounter major institutional difficulties, multiply licensing hurdles, and amplify political and public opposition to the overall nuclear power program. However, plutonium from spent LWR fuel is projected to be substantially more economic than enriched uranium, as fissile material for startup of LMRs when LMR deployment becomes economically justified. At that time, the LMR would fulfill its role as a reactor system that would protect the nation from diminishing energy resources. Development tasks towards defining and developing the most cost-effective LMR and associated fuel cycle remain very important. 29 refs., 3 figs., 18 tabs

  16. Consequence analysis of core meltdown accidents in liquid metal fast reactor

    International Nuclear Information System (INIS)

    Suk, S.D.; Hahn, D.

    2001-01-01

    Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of work to demonstrate the inherent and ultimate safety of the conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 Mw pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method was developed using a modified Bethe-Tait method to simulate the kinetics and hydraulic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the method for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies was also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters. (author)

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

    International Nuclear Information System (INIS)

    Gillett, J.E.; Wineman, A.L.

    1984-01-01

    A jacking mechanism is described 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

  18. Advanced liquid metal reactor development at Argonne National Laboratory during the 1980s

    International Nuclear Information System (INIS)

    Wade, D.C.

    1990-01-01

    Argonne National Laboratory's (ANL'S) effort to pursue the exploitation of liquid metal cooled reactor (LMR) characteristics has given rise to the Integral Fast Reactor (IFR) concept, and has produced substantial technical advancement in concept implementation which includes demonstration of high burnup capability of metallic fuel, demonstration of injection casting fabrication, integral demonstration of passive safety response, and technical feasibility of pyroprocessing. The first half decade of the 90's will host demonstration of the IFR closed fuel cycle technology at the prototype scale. The EBR-II reactor will be fueled with ternary alloy fuel in HT-9 cladding and ducts, and pyroprocessing and injection casting refabrication of EBR-II fuel will be conducted using near-commercial sized equipment at the Fuel cycle Facility (FCF) which is co-located adjacent to EBR-II. Demonstration will start in 1992. The demonstration of passive safety response achievable with the IFR design concept, (already done in EBR-II in 1986) will be repeated in the mid 90's using the IFR prototype recycle fuel from the FCF. The demonstration of scrubbing of the reprocessing fission product waste stream, with recycle of the transuranics to the reactor for consumption, will also occur in the mid 90's. 30 refs

  19. Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

    International Nuclear Information System (INIS)

    2003-01-01

    In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

  20. The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

    International Nuclear Information System (INIS)

    Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

    1990-01-01

    This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

  1. The DSNP simulation language and its application to liquid-metal fast breeder reactor transient analyses

    International Nuclear Information System (INIS)

    Saphier, D.; Madell, J.T.

    1982-01-01

    A new, special purpose block-oriented simulation language, the Dynamic Simulator for Nuclear Power Plants (DSNP), was used to perform a dynamic analysis of several conceptual design studies of liquid metal fast breeder reactors. The DSNP being a high level language enables the user to transform a power plant flow chart directly into a simulation program using a small number of DSNP statements. In addition to the language statements, the DSNP system has its own precompiler and an extensive library containing models of power plant components, algorithms of physical processes, material property functions, and various auxiliary functions. The comparative analysis covered oxide-fueled versus metal-fueled core designs and loop- versus pool-type reactors. The question of interest was the rate of change of the temperatures in the components in the upper plenum and the primary loop, in particular the reactor outlet nozzle and the intermediate heat exchanger inlet nozzle during different types of transients. From the simulations performed it can be concluded that metal-fueled cores will have much faster temperature transients than oxide-fueled cores due mainly to the much higher thermal diffusivity of the metal fuel. The transients in the pool-type design (either with oxide fuel or metal fuel) will be much slower than in the loop-type design due to the large heat capacity of the sodium pool. The DSNP language was demonstrated to be well suited to perform many types of transient analysis in nuclear power plants

  2. SWEEP, a computer program for the analysis of CDA energetics in liquid metal reactors

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Lee, Yong Bum; Hahn, Do Hee

    2003-12-01

    The SWEEP computer code was developed in this study to evaluate the work energy arising from two-phase expansion of fuel or sodium during core disruptive accidents in KALIMER. In the SWEEP program, scoping calculations with a modified Bethe-Tait method is first carried out using SCHAMBETA module to provide the initial thermodynamic conditions for the subsequent analyses to estimate the mechanical work energy generated in the reactor system. To estimate the work energy due to fuel-vapor expansion, a bounding approach is adopted to calculate the work potential assuming isentropic expansion to atmospheric pressure during super-prompt critical power excursions. Work potentials are also calculated in the SWEEP code for sodium expansion using the simple thermodynamic models including the infinite heat transfer model during expansion(Hicks and Menzies method) or more realistic zero heat transfer model for a typical initial condition of core disruptive accident. Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of the work to demonstrate the inherent and ultimate safety of conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 MWe pool-type sodium cooled prototype fast reactor that uses U-TRU-Zr metallic fuel

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

    International Nuclear Information System (INIS)

    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-11 with processed IFR fuel should provide a substantial and unique database for validating the computer codes being used to simulate postulated accident conditions

  4. Loss of coolant accident mitigation for liquid metal cooled space reactors

    International Nuclear Information System (INIS)

    Georgevich, Vladimir; Best, Frederick; Erdman, Carl

    1989-01-01

    A loss of coolant accident (LOCA) in a liquid metal-cooled space reactor system has been considered as a possible accident scenario. Development of new concepts that will prevent core damage by LOCA caused elevated temperatures is the primary motivation of this work. Decay heat generated by the fission products in the reactor core following shutdown is sufficiently high to melt the fuel unless energy can be removed from the pins at a sufficiently rapid rate. There are two major reasons that prevent utilization of traditional emergency cooling methods. One is the absence of gravity and the other is the vacuum condition outside the reactor vessel. A concept that overcomes both problems is the Saturated Wick Evaporation Method (SWEM). This method involves placing wicking structures at specific locations in the core to act as energy sinks. One of its properties is the isothermal behaviour of the liquid in the wick. The absorption of energy by the surface at the isothermal temperature will direct the energy into an evaporation process and not in sensible heat addition. The use of this concept enables establishment of isothermal positions within the core. A computer code that evaluates the temperature distribution of the core has been developed and the results show that this design will prevent fuel meltdown. (author)

  5. Study on liquid-metal MHD power generation system with two-phase natural circulation. Applicability to fast reactor conditions

    International Nuclear Information System (INIS)

    Saito, Masaki

    2000-03-01

    Feasibility study of the liquid-metal MHD power generation system combined with the high-density two-phase natural circulation has been performed for the applicability to the simple, autonomic energy conversion system of the liquid-metal cooled fast reactor. The present system has many promising aspects not only in the energy conversion process, but also in safety and economical improvements of the liquid-metal cooled fast reactor. For example, the high cycle efficiency can be expected because of the similarity of the present cycle to the Ericsson cycle. Sodium-Water Interaction problem can be excluded by proper combination of the working fluids. As the economical feature, the present system is so simple that the liquid-metal main circular pump, the steam turbine generator, and even the steam generator can be excluded if the thermodynamic working fluid is injected directly into the high temperature liquid metal MHD working fluid. In addition, the present system has the potential to be applied to various heat sources including solar energy because of the high flexibility of the operation temperature. In the present paper, as the first step of the feasibility study, the cycle analyses were performed to examine the effects of the main system parameters on the fundamental characteristics of the system. It is found that the cycle efficiency of the present system is enough competitive with that of the conventional steam turbine system. It is, however, found that the cycle efficiency depends strongly on the gas-liquid slip ratio in the two-phase flow channel. As the conclusions, it is recommended to perform experimental study to obtain the fundamental data, such as the gas-liquid slip ratio in the high-density liquid-metal two-phase natural circulation. (author)

  6. The detection of sodium vapor bubble collapse in a liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Carey, W.M.; Gavin, A.P.; Bobis, J.P.; Sheen, S.H.; Anderson, T.T.; Doolittle, R.D.; Albrecht, R.W.

    1977-01-01

    Sodium boiling detection utilizing the sound pressure emanated during the collapse of a sodium vapour bubble in a subcooled media is discussed in terms of the sound characteristic, the reactor ambient noise background, transmission loss considerations and performance criteria. Data obtained in several loss of flow experiments on Fast Test Reactor Fuel Elements indicate that the collapse of the sodium vapour bubble depends on the presence of a subcooled structure or sodium. The collapse pressure pulse was observed in all cases to be on the order of a kPa, indicating a soft type of cavitational collapse. Spectral examination of the pulses indicates the response function of the test structure and geometry is important. The sodium boiling observed in these experiments was observed to occur at a low ( 0 C) liquid superheat with the rate of occurrence of sodium vapor bubble collapse in the 3 to 30 Hz range. Reactor ambient noise data were found to be due to machinery induced vibrations flow induced vibrations, and flow noise. These data were further found to be weakly stationary enhancing the possibility of acoustic surveillance of an operating Liquid Metal Fast Breeder Reactor. Based on these noise characteristics and extrapolating the noise measurements from the Fast Flux Test Facility Pump (FFTP), one would expect a signal to noise ratio of up to 20 dB in the absence of transmission loss. The requirement of a low false alarm probability is shown to necessitate post detection analysis of the collapse event sequence and the cross correlation with the second derivative of the neutronic boiling detection signal. Sodium boiling detection using the sounds emitted during sodium vapor bubble collapse are shown to be feasible but a need for in-reactor demonstration is necessary. (author)

  7. Seismic isolation development for the US advanced liquid-metal reactor program

    International Nuclear Information System (INIS)

    Gluekler, E.L.; Bigelow, C.C.; DeVita, V.; Kelly, J.M.; Seidensticker, R.W.; Tajirian, F.F.

    1991-01-01

    GE Nuclear Energy, in association with a US Industrial Team and support from the US National Laboratories and Universities, is developing a modular liquid-metal reactor concept for the US DOE. The objective of this development is to provide, by the turn of the century, a reactor with optimized passive safety features that is economically competitive with other domestic energy sources, licensable, and ready for commercial deployment. One of the unique features of the concept is the seismic isolation of the reactor modules which decouples the reactors and their safety systems from potentially damaging ground motions and significantly enhances the structural resistance to high energy, as well as long-duration earthquakes. Seismic isolation is accomplished with high-damping natural-rubber bearings. The reactors are located in individual silos below grade level and are supported by the isolator bearings at approximately their center of gravity. This application of seismic isolation is the first for a US nuclear power plant. A development program has been established to assure the full benefits from the utilization of this new approach and to provide adequate system characterization and qualification for licensing certification. The development program, which is supported by the US DOE, ANL, Energy Technology Engineering Center (ETEC), the University of California at Berkeley (UC-Berkeley), GE, and Bechtel National, Inc. (BNI), is described and selected results are presented. The initial testing indicated excellent performance of high-damping natural-rubber bearings. The development of seismic isolation guidelines is in progress as a joint activity between ENEA of Italy and the GE Team. (orig./HP)

  8. A liquid-metal reactor for burning minor actinides of spent light water reactor fuel. 1: Neutronics design study

    International Nuclear Information System (INIS)

    Choi, H.; Downar, T.J.

    1999-01-01

    A liquid-metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors (LWRs). The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the Doppler constant, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200-MW(thermal) core is able to consume the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics

  9. A pump/intermediate heat exchanger assembly for a liquid metal reactor

    International Nuclear Information System (INIS)

    Nathenson, R.D.; Alexion, C.C.; Sumpman, W.C.

    1987-01-01

    A heat exchanger and electromagnetic pump assembly is disclosed comprising a heat exchanger housing defining an annularly shaped cavity and supporting therein a plurality of heat transfer tubes. An electromagnetic pump disposed beneath the heat exchanger comprises a circular array of flow couplers. Each flow coupler comprises a pump duct receiving primary liquid metal and a generator duct receiving a pumped intermediate liquid metal. A first plenum chamber is in communication with the generator ducts of all the flow couplers and receives intermediate liquid metal from inlet duct. The generator ducts exit their flows of intermediate liquid metal to a second plenum chamber in communication with the heat exchanger annularly shaped cavity to permit the flow of the intermediate liquid metal therethrough. A third plenum chamber receives collectively the flows of the primary liquid metal from the tubes and directs the primary liquid metal to the pump ducts of the flow couplers. The annular magnetic field of the electromagnetic pump is produced by a circular array of electromagnets having hollow windings cooled by a flow of intermediate liquid metal via tubes and manifolds. The leads to the electromagnets pass through an annular space around the inlet duct. (author)

  10. Analysis of the Nonlinear Density Wave Two-Phase Instability in a Steam Generator of 600MWe Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Choi, Seok Ki; Kim, Seong O

    2011-01-01

    A 600 MWe demonstration reactor being developed at KAERI employs a once-through helically coiled steam generator. The helically coiled steam generator is compact and is efficient for heat transfer, however, it may suffer from the two-phase instability. It is well known that the density wave instability is the main source of instability among various types of instabilities in a helically coiled S/G in a LMR. In the present study a simple method for analysis of the density wave two phase instability in a liquid metal reactor S/G is proposed and the method is applied to the analysis of density wave instability in a S/G of 600MWe liquid metal reactor

  11. Primary coolant pipe rupture event in liquid metal cooled reactors. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2004-08-01

    In liquid-metal cooled fast reactors (LMFR) the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). However, the primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors (Indira Gandhi Centre for Atomic Research, Kalpakkam, India, 13-17 January 2003) was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the technical meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the technical meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

  12. Sensitivity Analysis of Core Neutronic Parameters in Electron Accelerator-driven Subcritical Advanced Liquid Metal Reactor

    Directory of Open Access Journals (Sweden)

    Marziye Ebrahimkhani

    2016-02-01

    Full Text Available Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy (Ee and source multiplication coefficient (ks, has been investigated. A Monte Carlo code (MCNPX_2.6 has been used to calculate neutronic parameters such as effective multiplication coefficient (keff, net neutron multiplication (M, neutron yield (Yn/e, energy constant gain (G0, energy gain (G, importance of neutron source (φ∗, axial and radial distributions of neutron flux, and power peaking factor (Pmax/Pave in two axial and radial directions of the reactor core for four fuel loading patterns. According to the results, safety margin and accelerator current (Ie have been decreased in the highest case of ks, but G and φ∗ have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loading pattern, with increasing Ee from 100 MeV up to 1 GeV, Yn/e and G improved by 91.09% and 10.21%, and Ie and Pacc decreased by 91.05% and 10.57%, respectively. The results indicate that placement of the Np–Pu assemblies on the periphery allows for a consistent keff because the Np–Pu assemblies experience less burn-up.

  13. Experimental research and development of main circulation pump bearings in reactor plants using heavy liquid-metal coolants

    International Nuclear Information System (INIS)

    Zudin, A.; Beznosov, A.; Chernysh, A.; Prikazchikov, G.

    2015-01-01

    At the present time, specialists in Russia are engaged in designing the BREST-OD-300 fast neutron lead-coolant reactor plant. There is currently no experience in designing and operating axial pumps of lead-coolant reactor plants, including one of their major units – bearing unit. Selection and substantiation of operating and structural parameters of plain friction bearings used in main circulation pumps of reactor plants running on heavy liquid-metal coolants are important tasks that are solved at the NNSTU. Development of a feasible procedure for designing bearings and its components operating within the structure of the main circulation pump of a reactor plant running on a heavy liquid-metal coolant as well as guidelines for an optimized structural scheme of such bearings set a goal of performing a range of theoretically-calculated and experimental works. The report contains testing data of a hydrostatic bearing with reciprocal fricative choking tested on the NNSTU FT-4 bench running on a lead coolant within the range of 420-500degC. There have been presented a scheme of a bench for testing a contact friction bearing on a high-temperature coolant and the results of investigation tests of bearings of such type at T = 450 ÷ 500degC. Material of the bearing sleeve is steel 08X18H10T, and a possibility is provided with regard to installation of the bearing sleeves and shaft made of non-metal materials (ceramic materials, silicified graphite, etc.). The presented testing data of plain friction bearings operating in a high-temperature heavy liquid-metal coolant will serve as a ground for making an alternative choice of a plain friction bearing for the main circulation pump of a reactor plant running on a heavy liquid-metal coolant. (author)

  14. Comparative studies of liquid metals for an alternative divertor target in a fusion reactor

    Science.gov (United States)

    Tabarés, F. L.; Oyarzabal, E.; Tafalla, D.; Martin-Rojo, A. B.; Pastor, I.; Ochando, M. A.; Medina, F.; Zurro, B.; McCarthy, K. J.; the TJ-II Team

    2017-12-01

    Two liquid metals (LM), Li and LiSn (20:80 at), presently considered as alternative materials for the divertor target of a fusion reactor, have been exposed to the plasma in a capillary porous system (CPS) arrangement in TJ-II. A negligible perturbation of the plasma has been recorded in both cases, even when stellarator plasmas are particularly sensitive to high Z elements due to the tendency to central impurity accumulation. The surface temperature of the LM CPS samples (made of a tungsten mesh impregnated in SnLi or Li) has been measured during the plasma pulse with ms resolution by pyrometry and the thermal balance during heating and cooling has been used to obtain the thermal parameters of the SnLi and Li CPS arrangements. Temperatures as high as 1150 K during TJ-II plasma exposure were observed for the LiSn solid case. Strong changes in the thermal conductivity of the alloy were recorded in the cooling phase at temperatures close to the nominal melting point. The deduced values for the thermal conductivity of the LiSn alloy/CPS sample were significantly lower than those predicted from their individual components.

  15. Uncertainty correlation in stochastic safety analysis of natural circulation decay heat removal of liquid metal reactor

    International Nuclear Information System (INIS)

    Takata, Takashi; Yamaguchi, Akira

    2009-01-01

    Since various uncertainties of input variables are involved and nonlinearly-correlated in the Best Estimate (BE) plant dynamics code, it is of importance to evaluate the importance of input uncertainty to the computational results and to estimate the accuracy of the confidence level of the results. In order to estimate the importance and the accuracy, the authors have applied the stochastic safety analysis procedure using the Latin Hypercube sampling method to Liquid Metal Reactor (LMR) natural circulation Decay Heat Removal (DHR) phenomenon in the present paper. 17 input variables are chosen for the analyses and 5 influential variables, which affect the maximum coolant temperature at the core in a short period of time (several tens seconds), are selected to investigate the importance by comparing with the full-scope parametric analysis. As a result, it has been demonstrated that a comparative small number of samples is sufficient enough to estimate the dominant input variable and the confidence level. Furthermore, the influence of the sampling method on the accuracy of the upper tolerance limit (confidence level of 95%) has been examined based on the Wilks' formula. (author)

  16. Irradiation performance of U-Pu-Zr metal fuels for liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    Tsai, H.; Cohen, A.B.; Billone, M.C.; Neimark, L.A.

    1994-10-01

    This report discusses a fuel system utilizing metallic U-Pu-Zr alloys which has been developed for advanced liquid metal-cooled reactors (LMRs). Result's from extensive irradiation testing conducted in EBR-II show a design having the following key features can achieve both high reliability and high burnup capability: a cast nominally U-20wt %Pu-10wt %Zr slug with the diameter sized to yield a fuel smear density of ∼75% theoretical density, low-swelling tempered martensitic stainless steel cladding, sodium bond filling the initial fuel/cladding gap, and an as-built plenum/fuel volume ratio of ∼1.5. The robust performance capability of this design stems primarily from the negligible loading on the cladding from either fuel/cladding mechanical interaction or fission-gas pressure during the irradiation. The effects of these individual design parameters, e.g., fuel smear density, zirconium content in fuel, plenum volume, and cladding types, on fuel element performance were investigated in a systematic irradiation experiment in EBR-II. The results show that, at the discharge burnup of ∼11 at. %, variations on zirconium content or plenum volume in the ranges tested have no substantial effects on performance. Fuel smear density, on the other hand, has pronounced but countervailing effects: increased density results in greater cladding strain, but lesser cladding wastage from fuel/cladding chemical interaction

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    Hutter, E.; Pardini, J.A.

    1977-01-01

    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. 3 claims, 6 figures

  19. Liquid-metal fast breeder reactor fuel rod performance and modeling at high burnup

    International Nuclear Information System (INIS)

    Verbeek, P.; Toebbe, H.; Hoppe, N.; Steinmetz, B.

    1978-01-01

    The fuel rod modeling codes IAMBUS and COMETHE were used in the analysis and interpretation of postirradiation examination results of mixed-oxide fuel pins. These codes were developed in the framework of the SNR-300 research and development (R and D) program at Interatom and Belgonucleaire, respectively. SNR-300 is a liquid-metal fast breeder reactor demonstration plant designed and presently constructed in consortial cooperation by Germany, Belgium, and the Netherlands. RAPSODIE I, the two-bundle irradiation experiment, was irradiated in the French test FBR RAPSODIE FORTISSIMO and is one of the key irradiation experiments within the SNR-300 R and D program. The comparison of code predictions with postirradiation examination results concentrates on clad diameter expansions, clad total axial elongations, fuel differential and total axial elongations, fuel restructuring, and fission gas release. Fuel rod modeling was considered in the light of benchmarking of the codes, and there was consideration of fuel rod design for operation at low and high burnup

  20. Structural design aspects of innovative designs under development in the current US Liquid Metal-Cooled Reactor program

    International Nuclear Information System (INIS)

    Seidensticker, R.W.

    1986-01-01

    The US Liquid Metal-Cooled Reactor (LMR) program has been restructured and is now focussed on the development of innovative plant designs which emphasize shorter construction times, increased use of passive, inherently safe features, cost-competitiveness with LWR plants, and minimization of safety-related systems. These changes have a considerable effect on the structural design aspects of the LMR plant. These structural problems and their solutions now under study form the main focus of this paper. (orig.)

  1. Prism reactor system design and analysis of postulated unscrammed events

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.

    1991-08-01

    Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

  2. PRISM reactor system design and analysis of postulated unscrammed events

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.

    1991-01-01

    Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristic and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. (author)

  3. PRISM reactor system design and analysis of postulated unscrammed events

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.; Rosztoczy, Z.; Lane, J.

    1991-01-01

    Key safety characteristics of the PRISM reactor system include the passive reactor shutdown characteristics and the passive shutdown heat removal system, RVACS. While these characteristics are simple in principle, the physical processes are fairly complex, particularly for the passive reactor shutdown. It has been possible to adapt independent safety analysis codes originally developed for the Clinch River Breeder Reactor review, although some limitations remain. In this paper, the analyses of postulated unscrammed events are discussed, along with limitations in the predictive capabilities and plans to correct the limitations in the near future. 6 refs., 4 figs

  4. Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

    International Nuclear Information System (INIS)

    Tsoulfanidis, N.; Jankhah, M.H.

    1979-01-01

    The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

  5. Analysis of metallic fuel pin behaviors under transient conditions of liquid metal reactors

    International Nuclear Information System (INIS)

    Nam, Cheol; Kwon, Hyoung Mun; Hwang, Woan

    1999-02-01

    Transient behavior of metallic fuel pins in liquid metal reactor is quite different to that in steady state conditions. Even in transient conditions, the fuel may behave differently depending on its accident situation and/or accident sequence. This report describes and identifies the possible and hypothetical transient events at the aspects of fuel pin behavior. Furthermore, the transient experiments on HT9 clad metallic fuel have been analyzed, and then failure assessments are performed based on accident classes. As a result, the failure mechanism of coolant-related accidents, such as LOF, is mainly due to plenum pressure and cladding thinning caused by eutectic penetration. In the reactivity-related accidents, such as TOP, the reason to cladding failure is believed to be the fuel swelling as well as plenum pressure. The probabilistic Weibull analysis is performed to evaluate the failure behavior of HT9 clad-metallic fuel pin on coolant related accidents.The Weibull failure function is derived as a function of cladding CDF. Using the function, a sample calculation for the ULOF accident of EBR-II fuel is performed, and the results indicate that failure probability is less the 0.3%. Further discussion on failure criteria of accident condition is provided. Finally, it is introduced the state-of-arts for developing computer codes of reactivity-related fuel pin behavior. The development efforts for a simple model to predict transient fuel swelling is described, and the preliminary calculation results compared to hot pressing test results in literature.This model is currently under development, and it is recommended in the future that the transient swelling model will be combined with the cladding model and the additional development for post-failure behavior of fuel pin is required. (Author). 36 refs., 9 tabs., 18 figs

  6. Relevance of passive safety testing at the fast flux test facility to advanced liquid metal reactors - 5127

    International Nuclear Information System (INIS)

    Wootan, D.W.; Omberg, R.P.

    2015-01-01

    Significant cost and safety improvements can be realized in advanced liquid metal reactor (LMR) designs by emphasizing inherent or passive safety through crediting the beneficial reactivity feedbacks associated with core and structural movement. This passive safety approach was adopted for the Fast Flux Test Facility (FFTF), and an experimental program was conducted to characterize the structural reactivity feedback. Testing at the Rapsodie and EBR-II reactors had demonstrated the beneficial effect of reactivity feedback caused by changes in fuel temperature and core geometry mechanisms in a liquid metal fast reactor in a holistic sense. The FFTF passive safety testing program was developed to examine how specific design elements influenced dynamic reactivity feedback in response to a reactivity input and to demonstrate the scalability of reactivity feedback results from smaller cores like Rapsodie and EBR-II to reactor cores that were more prototypic in scale to reactors of current interest. The U.S. Department of Energy, Office of Nuclear Energy Advanced Reactor Technology program is in the process of preserving, protecting, securing, and placing in electronic format information and data from the FFTF, including the core configurations and data collected during the passive safety tests. Evaluation of these actual test data could provide insight to improve analytical methods which may be used to support future licensing applications for LMRs. (authors)

  7. Development and computational simulation of thermoelectric electromagnetic pumps for controlling the fluid flow in liquid metal cooled space nuclear reactors

    International Nuclear Information System (INIS)

    Borges, E.M.

    1991-01-01

    Thermoelectric Electromagnetic (TEEM) Pumps can be used for controlling the fluid flow in the primary and secondary circuits of liquid metal cooled space nuclear reactor. In order to simulate and to evaluate the pumps performance, in steady-state, the computer program BEMTE has been developed to study the main operational parameters and to determine the system actuation point, for a given reactor operating power. The results for each stage of the program were satisfactory, compared to experimental data. The program shows to be adequate for the design and simulating of direct current electromagnetic pumps. (author)

  8. Structural effects on fusion reactor blankets due to liquid metals in magnetic fields

    International Nuclear Information System (INIS)

    Lehner, J.R.; Reich, M.; Powell, J.R.

    1976-01-01

    The transient stress distribution caused in the blanket structure when the plasma current suddenly switches off in a time short compared to the L/R decay time of the liquid metal blanket was studied. Poloidal field of the plasma will induce a current to flow in the liquid metal and blanket walls. Since the resistance of the liquid lithium will be much less than that of the metal walls, the current can be considered as flowing around the blanket near the cross section perimeter, but in the lithium

  9. Scram and nonlinear reactor system seismic analysis for a liquid metal fast reactor

    International Nuclear Information System (INIS)

    Morrone, A.; Brussalis, W.G.

    1975-01-01

    The paper presents the analysis and results for a LMFBR system which was analyzed for both scram times and seismic responses such as bending moments, accelerations and forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node (translational and rotational). The model was developed to incorporate as many reactor components as possible without exceeding computer limitations. It consists of 12 reactor components with a total of 71 nodes, 69 beam and pin-jointed elements and 27 gap elements. The gap elements were defined by their clearances, impact spring constants and impact damping constants based on a 50% coefficient of restitution. The horizontal excitation input to the model was the response of the containment building at the location of the reactor vessel supports. It consists of a ten seconds Safe Shutdown Earthquake acceleration-time history at 0.005 seconds intervals and with a maximum acceleration of 0.408 g. The analysis was performed with two Westinghouse special purpose computer programs. The first program calculated the reactor system seismic responses and stored the impact forces on tape. The impact forces on the control rod driveline were converted into vertical frictional forces by multiplying them by a coefficient of friction, and then used by the second program for the scram time determination. The results give time history plots of various seismic responses, and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about 4 times longer than that calculated without the earthquake. The bending moment and shear force responses were used as input for the structural analysis (stresses, deflections, fatigue) of the various components, in combination with the other applicable loading conditions. (orig./HP) [de

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

    International Nuclear Information System (INIS)

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

    1994-01-01

    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

  11. Analysis of gas-liquid metal two-phase flows using a reactor safety analysis code SIMMER-III

    International Nuclear Information System (INIS)

    Suzuki, Tohru; Tobita, Yoshiharu; Kondo, Satoru; Saito, Yasushi; Mishima, Kaichiro

    2003-01-01

    SIMMER-III, a safety analysis code for liquid-metal fast reactors (LMFRs), includes a momentum exchange model based on conventional correlations for ordinary gas-liquid flows, such as an air-water system. From the viewpoint of safety evaluation of core disruptive accidents (CDAs) in LMFRs, we need to confirm that the code can predict the two-phase flow behaviors with high liquid-to-gas density ratios formed during a CDA. In the present study, the momentum exchange model of SIMMER-III was assessed and improved using experimental data of two-phase flows containing liquid metal, on which fundamental information, such as bubble shapes, void fractions and velocity fields, has been lacking. It was found that the original SIMMER-III can suitably represent high liquid-to-gas density ratio flows including ellipsoidal bubbles as seen in lower gas fluxes. In addition, the employment of Kataoka-Ishii's correlation has improved the accuracy of SIMMER-III for gas-liquid metal flows with cap-shape bubbles as identified in higher gas fluxes. Moreover, a new procedure, in which an appropriate drag coefficient can be automatically selected according to bubble shape, was developed. Through this work, the reliability and the precision of SIMMER-III have been much raised with regard to bubbly flows for various liquid-to-gas density ratios

  12. Method of detecting stacks with leaky fuel elements in liquid-metal-cooled reactor and apparatus for effecting same

    International Nuclear Information System (INIS)

    Aristarkhov, N.N.; Efimov, I.A.; Zaistev, B.I.; Peters, I.G.; Tymosh, B.S.

    1976-01-01

    Described is a method of detecting stacks with leaky fuel elements in a liquid-metal-cooled reactor, consisting in that prior to withdrawing a coolant sample, gas is accumulated in the coolant of the stack being controlled, the reactor being shut down, separated from the sample by means of an inert carrier gas, and the radioactivity of the separated gas is measured. An apparatus for carrying out said method comprises a sampler in the form of a tube parallel to the reactor axis in the hole of a rotating plug and adapted to move along the reactor axis. Made in the top portion of the tube are holes for the introduction of the inert carrier gas and the removal thereof together with the gases evolved from the coolant, while the bottom portion of the tube is provided with a sealing member

  13. E-SCAPE: A scale facility for liquid-metal, pool-type reactor thermal hydraulic investigations

    Energy Technology Data Exchange (ETDEWEB)

    Van Tichelen, Katrien, E-mail: kvtichel@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Mirelli, Fabio, E-mail: fmirelli@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Greco, Matteo, E-mail: mgreco@sckcen.be [SCK-CEN, Boeretang 200, 2400 Mol (Belgium); Viviani, Giorgia, E-mail: giorgiaviviani@gmail.com [University of Pisa, Lungarno Pacinotti 43, 56126 Pisa (Italy)

    2015-08-15

    Highlights: • The E-SCAPE facility is a thermal hydraulic scale model of the MYRRHA fast reactor. • The focus is on mixing and stratification in liquid-metal pool-type reactors. • Forced convection, natural convection and the transition are investigated. • Extensive instrumentation allows validation of computational models. • System thermal hydraulic and CFD models have been used for facility design. - Abstract: MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

  14. Performance of the prism reactor's passive decay heat removal system

    International Nuclear Information System (INIS)

    Magee, P.M.; Hunsbedt, A.

    1989-01-01

    The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  16. Structure of liquid metal cooled nuclear reactor with loops and steady vessel

    International Nuclear Information System (INIS)

    Costes, D.

    1990-01-01

    This structure comprises, in a vessel containing liquid metal, a nuclear core steadied on an alimentation diagrid and external loops comprising heat exchanger and reinjection pump of sodium in the diagrid. The vessel has the bottom resting on the concrete surround with a thermal stratification of the sodium between the bottom and the diagrid. This disposition has for advantage to allow a vertical connection of the sodium reinjection channel. This channel is contained in a metal sheath with a sliding leak tightness [fr

  17. An innovative design approach to a cost effective commercial liquid metal reactor

    International Nuclear Information System (INIS)

    Carelli, M. D.; Garkisch, H. D.; Hundal, R.; Arie, K.; Handa, N.; Ota, H.; Matsuyama, S.; Todreas, N.; Hejzlar, P.; Wells, P.; Louis, T. S.

    2008-01-01

    One of the requirements for the Advanced Recycle Reactor (ARR) in the DOE GNEP program is to be economically competitive with advanced LWRs. Sodium cooled Liquid Metal Reactors have been designed in two basic configurations, pool and loop, and various designs had different features, but the capital cost has consistently been substantially higher than for LWRs in spite of the LMRs higher efficiency due to the higher coolant exit temperature. We have identified two major areas as responsible for the divergence in capital costs. First, LMRs have been designed within the HCDA (Hypothetical Core Disruptive Accident) philosophy. Basically, the plant has to survive a major accident regardless whether this accident has any mechanistically acceptable, albeit very small, probability of occurring. Thus, in spite of their inherent reactivity feedbacks, much more effective than in LWRs, LMRs have been designed with redundant systems and costly design solutions which a probabilistic, rather than a deterministic approach, would eliminate or greatly simplify. The other major consideration is the presence of the intermediate heat transport system to prevent a Na/H 2 O reaction involving primary (radioactive) sodium. No other reactor type has three heat transport systems and in fact the LMR intermediate heat transport system is a substantial contributor to the capital cost. For our ARR rated at 1000 MWt power, we have selected a pool configuration because of its enhanced safety and its inherent capability of yielding a more compact reactor. The intermediate heat transport system is eliminated. Immersed in the pool is a double wall steam generator where water/steam flows through the inner tubes and Helium is in the space between the inner and outer tubes for detection purposes. Models of two different types of double wall steam generators designed independently by Westinghouse and Toshiba had been tested in the 1970-1980's at ETEC with favorable results. As we fully realize the risks

  18. Water tube liquid metal control

    International Nuclear Information System (INIS)

    Campbell, J.W.E.

    1981-01-01

    An improved heat exchanger for use in liquid metal cooled nuclear power reactors is described in which the heat is transferred between the flow of liquid metal which is to be cooled and a forced flow of liquid which is wholly or partly evaporated. (U.K.)

  19. On the problem of heat and mass exchange between liquid metal surface and structural elements in fast reactors

    International Nuclear Information System (INIS)

    Rineisky, A.A.; Sorokin, A.P.; Yatsenko, M.G.

    1986-01-01

    For the development of means ensuring normal operating conditions of the fast reactor vessel some design procedures for calculating temperature conditions of its structural elements over the liquid metal surface are required. The radiative heat transfer from the liquid metal surface playing an important part at working temperatures (550 deg. C), the effect of experimentally detected fog formation process (not taken into account before) upon the radiative heat exchange has been considered. A simplified heat transfer model based upon separation of thin thermal boundary layers and of the main volume at a constant temperature has been proposed. Calculation relationships for the heat flux from the reactor vessel roof have been obtained by solving a one-dimensional equation of radiation transfer within the boundary layer and a three-dimensional one in the bulk volume at an approximation of moments with Marshak boundary conditions. Evaluations performed have shown a possibility of a considerable decrease of the vessel roof temperature due to fog formation. The observed asymmetry of the temperature distribution in the cover gas is explained in this case greater fog density near the evaporation surface and by a possibility of some radiative energy loss due to evaporation from the droplets surface. (author)

  20. Liquid metal level measurement

    International Nuclear Information System (INIS)

    Hale, J.C.; Leyland, K.S.

    1982-01-01

    A liquid metal level indicator is described which can be used to measure, in a stainless steel tank, the level of a nuclear reactor coolant such as sodium. The instrument, which is based on the eddy current induction effect, gives readings over substantially the full depth of the tank and indicates the sense of change of level. (U.K.)

  1. PRISM

    Science.gov (United States)

    2017-12-21

    societal issues , given countervailing moral , legal, and ethical interpretations, as well as suspicions PRISM 7, NO. 2 FEATURES | 13 COGNITIVE-EMOTIONAL...to tell our story. Additionally, U.S. practi- tioners are bound by asymmetric legal, moral , and ethical constraints that often keep them from being...the founder of SpaceX and Tesla, withdrew from two business councils providing advice to the administration on economic issues . Further driving

  2. Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

    International Nuclear Information System (INIS)

    1994-01-01

    The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned

  3. Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-01

    The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned.

  4. Supporting the national energy needs for the early 21st century with the advanced liquid metal reactor system (ALMRS)

    International Nuclear Information System (INIS)

    Hutchins, B.A.; Quinn, J.E.; Thompson, M.L.

    1991-01-01

    This paper presents a cost effective approach to providing a major contribution to the electricity needs of the United States in the early 21st century through an integrated Advanced Liquid Metal Reactor System (ALMRS). This system has several synergistic components which are under development by the United States Department of Energy (DOE): the modular, passively safe ALMR reactor design; metal fuel recycle (aka IFR); and the processing of LWR spent fuel to use as startup fuel for the ALMR. Each of these components contributes to an overall system behavior that will be able to provide an important portion of the United States' electrical energy needs beginning about the year 2010, while at the same time translating some fuel wastes of the LWR spent fuel to an asset. This paper describes each of these components and their synergism. Economic projections and busbar costs for this system are also presented

  5. Complex risk analysis for loss of electric power in liquid metal nuclear reactor by system dynamics (SD) method

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Tae Ho [Seoul National Univ. (Korea, Republic of). Dept. of Nuclear Engineering

    2012-07-15

    The power stabilization of the nuclear power plants (NPPs) is investigated in the aspect of the liquid metal coolant. The quantification of the risk analysis is performed by the system dynamics (SD) method which is processed by the feedback and accumulation complex algorithms. The Vensim software package is used for the simulations, which is supported by the Monte-Carlo method. There are 2 kinds of considerations as the economic and safety properties. The result shows the stability of the operations when the power can be decided. This shows the higher efficiency of the reactor. The failure frequency is 16/60 = 27%. In the event of Power Stabilized, the failure event is in the quite lower frequency rate. The commercial use of the reactor is important in the operations. (orig.)

  6. The US Advanced Liquid Metal Reactor and the Fast Flux Test Facility Phase IIA passive safety tests

    International Nuclear Information System (INIS)

    Shen, P.K.; Harris, R.A.; Campbell, L.R.; Dautel, W.A.; Dubberley, A.E.; Gluekler, E.L.

    1992-07-01

    This report discusses the safety approach of the Advanced Liquid Metal reactor program, sponsored by the US Department of Energy, which relies upon passive reactor responses to off-normal condition to limit power and temperature excursions to levels that allow safety margins. Gas expansion modules (GEM) have included in the design to provide negative reactivity to enhance these margins in the extremely unlikely event that pumping power is lost and the highly reliable scram system fails to operate. The feasibility and beneficial features of these devices were first demonstrated in the core of the Fast Flux Test Facility (FFTF) in 1986. Preapplication safety evaluations by the US Nuclear Regulatory Commission have identified areas that must be addressed if these devices are to be relied on. One of these areas is the response of the reactor when it is critical and the pumps are turned on, resulting in positive reactivity being added to the core. Tests to examine such transients have been performed as part of the continuing FFTF program to confirm the passive safety characteristics of liquid metal reactors (LMR). The primary tests consisted of starting the main coolant pumps, which forced sodium coolant into the GEMS, decreasing neutron leakage and adding positive reactivity. The resulting transients were shown to be benign and easily mitigated by the reactivity feedbacks inherent in the FFTF and all LMRs. Steady-state auxiliary tests of the GEM and feedback reactivity worths accurately predicted the transient results. The auxiliary GEM worth tests also demonstrated that the worth can be determined at a subcritical state, which allows for a verification of the GEM's availability prior to ascending to power

  7. Korea advanced liquid metal reactor development - Development of measuring techniques of the sodium two-phase flow

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Moo Hwan; Cha, Jae Eun [Pohang University of Science and Technology, Pohang (Korea)

    2000-04-01

    The technology which models and measures the behavior of bubble in liquid sodium is very important to insure the safety of the liquid metal reactor. In this research, we designed/ manufactured each part and loop of experimental facility for sodium two phase flow, and applied a few possible methods, measured characteristic of two phase flow such as bubbly flow. A air-water loop similar to sodium loop on each measuring condition was designed/manufactured. This air-water loop was utilized to acquire many informations which were necessary in designing the two phase flow of sodium and manufacturing experimental facility. Before the manufacture of a electromagnetic flow meter for sodium, the experiment using each electromagnetic flow mete was developed and the air-water loop was performed to understand flow characteristics. Experiments for observing the signal characteristics of flow were performed by flowing two phase mixture into the electromagnetic flow mete. From these experiments, the electromagnetic flow meter was designed and constructed by virtual electrode, its signal processing circuit and micro electro magnet. It was developed to be applicable to low conductivity fluid very successfully. By this experiment with the electromagnetic flow meter, we observed that the flow signal was very different according to void fraction in two phase flow and that probability density function which was made by statistical signal treatment is also different according to flow patterns. From this result, we confirmed that the electromagnetic flow meter could be used to understand the parameters of two phase flow of sodium. By this study, the experimental facility for two phase flow of sodium was constricted. Also the new electromagnetic flow meter was designed/manufactured, and experimental apparatus for two phase flow of air-water. Finally, this study will be a basic tool for measurement of two phase flow of sodium. As the fundamental technique for the applications of sodium at

  8. Experiments for liquid metal embrittlement of fusion reactor materials by liquid lithium

    International Nuclear Information System (INIS)

    Grundmann, M.; Borgstedt, H.U.

    1984-10-01

    The liquid metal embrittlement behaviour of two martensitic-ferritic steels [X22CrMoV121 (Nr. 1.4923) and X18CrMoVNb 121 (Nr. 1,4914)] and one austenite chromium-nickel-steel X5CrNi189 (Nr. 1.4301) was investigated. Tensile tests in liquid lithium at 200 and 250 0 C with two different strain rates on precorroded samples (1000 h at 550 0 C in lithium) were carried out. Reference values were gained from tensile tests in air (RT, 250 0 C). It is concluded that there is sufficient compatibility of the austenitic steel with liquid lithium. The use of the ferritic-martensitic steels in liquid lithium on the other hand, especially at temperatures of about 550 0 C, seems to be problematic. The experimental results led to a better understanding of LME, applying the theory of this material failure. (orig./IHOE) [de

  9. Capital cost evaluation of liquid metal reactor by plant type - comparison of modular type with monolithic type -

    International Nuclear Information System (INIS)

    Mun, K. H.; Seok, S. D.; Song, K. D.; Kim, I. C.

    1999-01-01

    A preliminary economic comparison study was performed for KALIMER(Korea Advanced LIquid MEtal Reactor)between a modular plant type with 8 150MWe modules and a 1200MWe monolithic plant type. In both cases of FOAK (First-Of-A-Kind) Plant and NOAK (Nth-Of-A-Kind) Plant, the result says that the economics of monolithic plant is superior to its modular plant. In case of NOAK plant comparison, however, the cost difference is not significant. It means that modular plant can compete with monolithic plant in capital cost if it makes efforts of cost reduction and technical progress on the assumption that the same type of NOAK plant will be constructed continuously

  10. State-of-the-art report on the development of liquid metal reactor fuel cladding materials in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Ho; Kuk, Il Hiun; Ryu, Woo Seog; Jang, Jin Sung; Rhee, Chang Kyu; Kim, Dae Whan; Park, Soon Dong; Kim, Woo Gon; Chung, Man Kyo; Han, Chang Hee

    1998-01-01

    PNC 1520 and PNC-FM5 have been developed as a cladding materials for LMR in Japan. PNC 1520 has superior swelling resistance and high temperature properties to PNC 31.6. And PNC-FMS steel has shown a high rupture stress as well as good neutron irradiation performance. In addition oxide dispersed ferritic steel (PNC-ODS) and 12Cr-8Mo steel have been developed. This report will give an insight for choosing and developing the materials to be applied to the KAERI prototype liquid metal reactor which is going to be operable in 2010 by analysis of the characteristics of cladding materials developed in Japan. (author). 39 refs., 2 tabs., 23 figs

  11. Study on liquid-metal MHD power generation system with two-phase natural circulation. Applicability to fast reactor conditions

    International Nuclear Information System (INIS)

    Saito, Masaki

    2001-03-01

    Feasibility study of the liquid-metal MHD power generation system combined with the high-density two-phase natural circulation has been performed for the applicability to the simple, autonomic energy conversion system of the liquid-metal cooled fast reactor. The present system has many promising aspects not only in the energy conversion process, but also in safety and economical improvements of the liquid-metal cooled fast reactor. In the previous report, as the first step of the feasibility study, the cycle analyses were performed to examine the effects of the main system parameters on the fundamental characteristics of the system. It was found that the cycle efficiency of the present system is enough competitive with that of the conventional steam turbine system. It was also found that the cycle efficiency depends strongly on the gas-liquid slip ratio in the two-phase flow channel. However, it is very difficult to estimate the gas-liquid slip ratio theoretically, especially in the heavy liquid metal two-phase natural circulation. For example, the effects of MHD load on the two-phase flow characteristics, such as the void fraction and gas-liquid slip ratio are not known well. In the present study, therefore, as the second step of the feasibility study, a series of the experiments were performed to investigate, especially, the effect of MHD load at the single-phase shown-comer flow channel on the characteristics of the two-phase natural circulation. In the first series of the experiments, Woods-metal (Density: 9517 Kg/m 3 ) and nitrogen gas were chosen as the two-phase working fluids. The MHD pressure drop was simulated by the ball valve. The experiments with water and nitrogen gas were also performed to check the effects of the physical properties. From the present experiments, it is found that the average void fraction in the two-phase flow channel is determined by the force balance between the MHD pressure drop, frictional and pressure losses in the tube, and

  12. Safety and licensing analysis of 471 MWt PRISM reactor module

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.

    1991-01-01

    Independent analyses of unscrammed events postulated for the 471 MW t PRISM reactor module, as performed by Brookhaven National Laboratory for the U.S. Nuclear Regulatory Commission, are discussed. Peak fuel, cladding, and sodium temperatures for several postulated events were compared against reference temperatures. In general, BNL could essentially duplicate GE's analytical results, by making GE's assumptions. For the postulated unscrammed loss of flow (ULOF) events, the newly added Gas Expansion Modules (GEMs), have improved the passive shutdown response. With respect to the unscrammed loss of heat sink (ULOHS) event, there was little impact expected from the latest revisions, and little was found. For the unscrammed transient over-power (UTOP) events, uncertainties with respect to the behavior of ternary metal fuel leave unresolved the issue of how large a UTOP initiator (between 20 and 40 cents) may be acceptable. (author)

  13. Progress in liquid metal fast reactor technology. Proceedings of the 28th meeting of the International Working Group on Fast Reactors

    International Nuclear Information System (INIS)

    1996-04-01

    The key objectives and activities of Member State liquid metal fast reactor (LMFR) programmes are: Demonstration of effective designs; demonstration of system safety; demonstration of economic competitiveness with other power generation systems. The International Working Group on Fast Reactors (IWGFR) at its 1995 meeting observed that while some countries (as a result of static or falling power demand) are reducing the research and development programmes or delaying the commercial deployment of fast reactors, other countries are planning to introduce these reactors and are embarking on their own development programmes. In these circumstances the international exchange of information and experience is of increasing importance. These proceedings contain updated information from long standing members of the IWGFR and new information on the status of LMFR research and development from new members of the Group: Brazil, China, Republic of Kazakhstan and the Republic of Korea. Refs, figs, tabs

  14. Progress in liquid metal fast reactor technology. Proceedings of the 28th meeting of the International Working Group on Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-04-01

    The key objectives and activities of Member State liquid metal fast reactor (LMFR) programmes are: Demonstration of effective designs; demonstration of system safety; demonstration of economic competitiveness with other power generation systems. The International Working Group on Fast Reactors (IWGFR) at its 1995 meeting observed that while some countries (as a result of static or falling power demand) are reducing the research and development programmes or delaying the commercial deployment of fast reactors, other countries are planning to introduce these reactors and are embarking on their own development programmes. In these circumstances the international exchange of information and experience is of increasing importance. These proceedings contain updated information from long standing members of the IWGFR and new information on the status of LMFR research and development from new members of the Group: Brazil, China, Republic of Kazakhstan and the Republic of Korea. Refs, figs, tabs.

  15. Evolution of fast reactor core spectra in changing a heavy liquid metal coolant by molten PB-208

    Energy Technology Data Exchange (ETDEWEB)

    Blokhin, D. A.; Mitenkova, E. F. [Nuclear Safety Inst., Russian Academy of Sciences, B. Tulskaya 52, Moscow, 115119 (Russian Federation); Khorasanov, G. L.; Zemskov, E. A.; Blokhin, A. I. [State Scientific Center, Russian Federation, Inst. of Physics and Power Engineering, Bondarenko Square 1, Obninsk, 249033 (Russian Federation)

    2012-07-01

    In the paper neutron spectra of fast reactor cooled with lead-bismuth or lead-208 are given. It is shown that in changing the coolant from lead-bismuth to lead-208 the core neutron spectra of the fast reactor FR RBEC-M are hardening in whole by several percents when a little share of low energy neutrons (5 eV - 50 keV) is slightly increasing. The shift of spectra to higher energies permits to enhance the fuel fission while the increased share of low energy neutrons provides more effective conversion of uranium-238 into plutonium due to peculiarity of {sup 238}U neutron capture cross section. Good neutron and physical features of molten {sup 208}Pb permit to assume it as perspective coolant for fast reactors and accelerator driven systems. The one-group cross sections of neutron radiation capture, {sigma}(n,g), by {sup 208}Pb, {sup 238}U, {sup 99}Tc, mix of lead and bismuth, {sup nat}Pb-Bi, averaged over neutron spectra of the fast reactor RBEC-M are given. It is shown that one-group cross sections of neutron capture by material of the liquid metal coolant consisted from lead enriched with the stable lead isotope, {sup 208}Pb, are by 4-7 times smaller {sigma}(n,g) for the coolant {sup nat}Pb-Bi. The economy of neutrons in the core cooled with {sup 208}Pb can be used for reducing reactor's initial fuel load, increasing fuel breeding and transmutation of long lived fission products, for example {sup 99}Tc. Good neutron and physical features of lead enriched with {sup 208}Pb permit to consider it as a perspective low neutron absorbing coolant for fast reactors and accelerator driven systems. (authors)

  16. The R and D issues necessary to achieve the safety design of commercialized liquid-metal cooled fast reactors

    International Nuclear Information System (INIS)

    Shoji, Kotake; Koji, Dozaki; Shigenobu, Kubo; Yoshio, Shimakawa; Hajime, Niwa; Masakazu, Ichimiya

    2002-01-01

    Within the framework of the feasibility study on commercialized fast reactor cycle systems (hereafter described as F/S), the safety design principle is investigated and several kinds of design studies are now in progress. Among the designs for liquid-metal cooled fast reactor (LMR), the advanced loop type sodium cooled fast reactor (FR) is one of the promising candidate as future commercialized LMR. In this paper, the safety related research and development (R and D) issues necessary to achieve the safety design are described along the defence-in-depth principle, taking account of not only the system characteristics of the advanced loop concepts but also design studies and R and D experiences so far. Safety issues related to the hypothetical core disruptive accidents (CDA) are emphasized both from the prevention and mitigation. A re-criticality free core concept with a special fuel assembly is pursued by performing both analytical and experimental efforts, in order to realize the rational design and to establish easy-to-understand safety logic. Sodium related issues are also given to ensure plant availability and to enhance the acceptability to the public. (authors)

  17. An examination of reliability critical items in liquid metal reactors: An analysis by the Centralized Reliability Data Organization (CREDO)

    International Nuclear Information System (INIS)

    Humphrys, B.L.; Haire, M.J.; Koger, K.H.; Manneschmidt, J.F.; Setoguchi, K.; Nakai, R.; Okubo, Y.

    1987-01-01

    The Centralized Reliability Data Organization (CREDO) is the largest repository of liquid metal reactor (LMR) component reliability data in the world. It is jointly sponsored by the US Department of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. The CREDO data base contains information on a population of more than 21,000 components and approximately 1300 event records. A conservative estimation is that the total component operating hours is approaching 3.5 billion hours. Because data gathering for CREDO concentrates on event (failure) information, the work reported here focuses on the reliability information contained in CREDO and the development of reliability critical items lists. That is, components are ranked in prioritized lists from worst to best performers from a reliability standpoint. For the data contained in the CREDO data base, FFTF and JOYO show reliability growth; EBR-II reveals a slight unreliability growth for those components tracked by CREDO. However, tabulations of events which cause reactor shutdowns decrease with time at each site

  18. Liquid metal purification device

    International Nuclear Information System (INIS)

    Sakai, Takao; Shimoyashiki, Shigehiro.

    1992-01-01

    The device of the present invention concerns a liquid metal purification device for removing and purifying impuries in liquid metal sodium used as coolants of an FBR type reactor. A vessel having a group of pipes made of hydrogen permeable metal at the inside thereof is disposed to the inlet pipeline of a cold trap. The group of hydrogen permeable metal pipes is connected to an exhaust pipe and a vacuum pump, so that the inside of the pipes is exhausted. Liquid metal sodium branched from the main pipeline of a coolant system passes through the outer side of the group of the hydrogen permeable metal pipes. In this cae, hydrogen contained as impurities in the liquid metal sodium diffuses and permeates the hydrogen permeation metal pipes and enters into the pipe group and is discharged out of the system by the vacuum pump. This can mitigate the hydrogen removing burden of the cold trap, to extend the device life time. (I.N.)

  19. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  20. Identification and management of plant aging and life extension issues for a liquid-metal-cooled reactor

    International Nuclear Information System (INIS)

    King, R.W.; Perry, W.H.

    1991-06-01

    Experimental Breeder Reactor 2 (EBR-2) is a pool-type sodium-cooled fast reactor that supports extensive experimental, test and demonstration programs while providing electrical power to the local grid. EBR-2 is a US Department of Energy Facility located at the Idaho National Engineering Laboratory and operated by Argonne National Laboratory (ANL). The current mission of EBR-2 is to serve as the operational prototype for the Integral Fast Reactor demonstration program. This mission and other programs require EBR-2 to operate reliability to a 40-year lifetime, a significant extension beyond the five to ten year life originally planned for the facility. The benefits of operating EBR-2 in the extended-life mode are important for providing long-term operational performance data for a sodium-cooled fast reactor that is not available elsewhere. Identification and preliminary assessment of potential life-limiting factors indicate that, with appropriate consideration given in the design phase, the sodium-cooled plant has potential for a very long operational lifetime. Achievement of a 40-year lifetime with high reliability is important not only for achieving the near-term goals of the EBR-2/IFR programs, but for the advancement of the liquid-metal-cooled reactor concept to the demonstration/commercialization phase. Key features make extended-life operation feasible based on the use of sodium as the primary coolant: low-pressure, high thermal capacity primary system and a low-pressure secondary system requiring no active valves; and limited corrosion of components. 2 refs

  1. Improvements to secondary coolant circuits of a liquid metal cooled nuclear reactor

    International Nuclear Information System (INIS)

    Brachet, Alain.

    1981-01-01

    This invention concerns improvements to secondary coolant-systems for sodium cooled nuclear reactors. It further concerns a protective device for a free level mechanical pump which prevents any gas bubbles due to leaks of the working gas of the pump from entering the secondary system of the nuclear reactor [fr

  2. Liquid metal reactor cover gas purification and analysis in the USA

    Energy Technology Data Exchange (ETDEWEB)

    Allen, K J [Argonne National Laboratory, EBR-II Division, Idaho Falls, ID (United States); Meadows, G E; Schuck, W J [Westinghouse Hanford Company, Richland, WA (United States)

    1987-07-01

    Two sodium cooled reactors are currently being operated In the United States of America for the U.S. Department of Energy. These are Experimental Breeder Reactor II, EBR-ll, and the Fast Flux Test Facility, FFTF. EBR-ll is located near Idaho Falls, Idaho and the FFTF is near Rich land, Washington. These reactors are currently engaged In a wide range of testing including fuels and materials tests, and plant system performance and safety development. The U.S. DOE program also includes designs of a next generation sodium cooled power reactor. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA.

  3. Liquid metal reactor cover gas purification and analysis in the USA

    International Nuclear Information System (INIS)

    Allen, K.J.; Meadows, G.E.; Schuck, W.J.

    1986-09-01

    Two sodium cooled reactors are currently being operated in the United States of America for the US Department of Energy. These are Experimental Breeder Reactor 11, EBR-11, and the Fast Flux Test Facility, FFTF. EBR-11 is located near Idaho Falls, Idaho, and the FFTF is near Richland, Washington. These reactors are currently engaged in a wide range of testing including fuels and materials tests, and plant system performance and safety development. The US DOE program also includes designs of a next generation sodium cooled power reactor. The FFTF and EBR-11 communities are providing input to these designs. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA

  4. Liquid metal reactor cover gas purification and analysis in the USA

    International Nuclear Information System (INIS)

    Allen, K.J.; Meadows, G.E.; Schuck, W.J.

    1987-01-01

    Two sodium cooled reactors are currently being operated In the United States of America for the U.S. Department of Energy. These are Experimental Breeder Reactor II, EBR-ll, and the Fast Flux Test Facility, FFTF. EBR-ll is located near Idaho Falls, Idaho and the FFTF is near Rich land, Washington. These reactors are currently engaged In a wide range of testing including fuels and materials tests, and plant system performance and safety development. The U.S. DOE program also includes designs of a next generation sodium cooled power reactor. This paper discusses the efforts to develop and operate cover gas systems for the sodium cooled nuclear reactor program in the USA

  5. Low-enrichment and long-life Scalable LIquid Metal cooled small Modular (SLIMM-1.2) reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States); Chemical and Biological Engineering Department, University of New Mexico, Albuquerque, NM (United States); Palomino, Luis M.; Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States)

    2017-05-15

    Highlights: • Developed low enrichment and natural circulation cooled SLIMM-1.2 SMR for generating 10–100 MW{sub th}. • Neutronics analyses estimate operation life and temperature reactivity feedback. • At 100 MW{sub th}, SLIMM-1.2 operates for 6.3 FPY without refueling. • SLIMM-1.2 has relatively low power peaking and maximum UN fuel temperature < 1400 K. - Abstract: The Scalable LIquid Metal cooled small Modular (SLIMM-1.0) reactor with uranium nitride fuel enrichment of 17.65% had been developed for generating 10–100 MW{sub th} continuously, without refueling for ∼66 and 5.9 full power years, respectively. Natural circulation of in-vessel liquid sodium (Na) cools the core of this fast energy spectrum reactor during nominal operation and after shutdown, with the aid of a tall chimney and an annular Na/Na heat exchanger (HEX) of concentric helically coiled tubes. The HEX at the top of the downcomer maximizes the static pressure head for natural circulation. In addition to the independent emergency shutdown (RSS) and reactor control (RC), the core negative temperature reactivity feedback safely decreases the reactor thermal power, following modest increases in the temperatures of UN fuel and in-vessel liquid sodium. The decay heat is removed from the core by natural circulation of in-vessel liquid sodium, with aid of the liquid metal heat pipes laid along the reactor vessel wall, and the passive backup cooling system (BCS) using natural circulation of ambient air along the outer surface of the guard vessel wall. This paper investigates modifying the SLIMM-1.0 reactor design to lower the UN fuel enrichment. To arrive at a final reactor design (SLIMM-1.2), the performed neutronics and reactivity depletion analyses examined the effects of various design and material choices on both the cold-clean and the hot-clean excess reactivity, the reactivity shutdown margin, the full power operation life at 100 MW{sub th}, the fissile production and depletion, the

  6. Theoretical investigations of the meltoff and resolidification process of fuel claddings during accidents in liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Angerer, G.

    1978-08-01

    During loss-of-coolant-flow accidents in liquid metal cooled fast breeder reactors with failure to scram the fuel claddings will melt after boiling and evaporation of the coolant. The CMOT model presented here describes the subsequent process of relocation and resolidification of the molten claddings. The basic thermohydrodynamics equations of the two-phase flow of cladding material and sodium vapor are solved numerically by differential approximations in a Eulerian reference net. The results calculated by the model improved the insight into the dynamics of the cladding relocation process. Here are the main results: - Shortly after the onset of cladding relocation large waves of molten cladding material are generated. The motion of these waves contributes considerably to the material transport. - The dynamics of cladding relocation exhibits strong local incoherences. - The formation of cladding blockages observed at the ends of the fuel region is confirmed by the calculations. - In case of incoherent cladding meltoff less cladding material is transported upwards. - Cladding relocation strongly depends on the axial pressure drop and the underlying friction factor correlations. Recalculation of the R5 loss-of-coolant-flow experiment performed in the U.S. TREAT test reactor is in good agreement with the experimental data. (orig./HP) 891 HP [de

  7. Analytical studies on the impact of using repeated-rib roughness in LMR [Liquid Metal Reactor] decay heat removal systems

    International Nuclear Information System (INIS)

    Obot, N.T.; Tessier, J.H.; Pedersen, D.R.

    1988-01-01

    A numerical study was carried out to determine the effects of roughness on the thermal performance of Liquid Metal Reactor (LMR) decay heat removal systems for a range of possible design configurations and operating conditions. The ranges covered for relative rib height (e/D/sub h/), relative pitch (p/e) and flow attack angle were 0.026--0.103, 5--20 and 0--90 degrees, successively. The heat flux was varied between 1.1 and 21.5 kW/m 2 (0.1 and 2.0 kW/ft 2 ). Calculations were made for three cases: smooth duct with no ribs, ribs on both the guard vessel and collector wall, and ribs on the collector wall only. The results indicate that significant benefits, amounting to nearly two-fold reductions in guard vessel and collector wall temperatures, can be realized by placing repeated ribs on both the guard vessel and the collector wall. The magnitudes of the reduction in the reactor vessel temperature are considerably smaller. In general, the level of improvement, be it with respect to temperature or heat flux, is only mildly affected by changes in rib height or pitch but exhibits greater sensitivity to the assumed value for the system form loss. When the ribs are placed only on the collector wall, the heat removal capability is substantially reduced

  8. Report to the Congress: liquid metal fast breeder reactor program--past, present, and future, Energy Research and Development Administration

    International Nuclear Information System (INIS)

    1975-01-01

    The past, present, and future of the liquid metal fast breeder reactor (LMFBR) program, the Nation's highest priority energy program, are studied. ERDA anticipates that the operation of the first large commercial breeder will start in 1987, and that 186 commercial-size breeders will be in operation by the year 2000. The breeder program is made up of six major areas, each dealing with an important element of technology: reactor physics; fuels and materials; fuel recycle; safety; component development; plant experience; and facilities used in the LMFBR program. ERDA is implementing a new system for administering, managing, and controlling the breeder program that will provide increased program visibility and control. Federal funding for breeder development was $168 million in FY 1971, accounting for 40% of the total Federal R and D energy budget; in FY 1976 Federal funding for this program will be $474 million, only 26% of total Federal funding for energy research. Besides Federal funds, over half a billion dollars have been or will be invested by industry over the next 5 to 10 years to develop the breeder and to build a demonstration plant. Five other nations--the United Kingdom, France, Japan, West Germany, and the Soviet Union--have a high priority national energy program for developing the LMFBR. These foreign breeder programs could contribute important data and information to the U.S. program

  9. Inherently safe shutdown of a liquid metal reactor upon a loss of intermediate cooling

    International Nuclear Information System (INIS)

    Feldman, E.E.; Mohr, D.

    1986-01-01

    Two unprotected (i.e., no scram or plant protection system action) loss-of-heat sink transients are scheduled to be performed on the Experimental Breeder Reactor-II in the Spring of 1986. One is to be initiated from full power (about 60 MW) and the other from half power. The loss-of-heat sink in each test is to be accomplished by essentially stopping the secondary-loop sodium coolant flow in about 20 s. Pretest predictions for the two tests, provided herein, show the reactor to passively shut down with average reactor outlet coolant temperatures rising no more that 5/degree/C before dropping and asymptotically approaching a quenching (or ''smothering'') temperature. This mild behavior is seen to be a direct consequence of the substantial negative reactivity feedback of the reactor coupled with the heat capacitance of the massive sodium pool in the primary tank. 14 refs., 11 figs., 1 tab

  10. Consequences of pipe ruptures in metal fueled, liquid metal cooled reactors

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1990-01-01

    The capability to simulate pipe ruptures has recently been added to the SASSYS-1 LMR systems analysis code. Using this capability, the consequences of severe pipe ruptures in both loop-type and pool-type reactors using metal fuel were investigated. With metal fuel, if the control rods scram then either type of reactor can easily survive a complete double-ended break of a single pipe; although, as might be expected, the consequences are less severe for a pool-type reactor. A pool-type reactor can even survive a protected simultaneous breaking of all of its inlet pipes without boiling of the coolant or melting of the fuel or cladding. 2 refs., 16 figs., 1 tab

  11. Status of Liquid Metal Fast Reactor Development in the United States of America, March 1987

    International Nuclear Information System (INIS)

    Horton, K.E.

    1987-01-01

    In order to meet the objective to develop and demonstrate economically competitive reactor designs and associated fuel cycles early in the next century, the U.S. program has become more focused. Two innovative reactor designs supported by the metal-fueled Integral Fast Reactor program are being directed at fulfilling a series of advanced reactor goals. The supporting technology programs and facilities are being refocused to support the overall goals. International collaboration is being broadened to provide the two-way support across the spectrum of plant projects and the fuel cycle. This program is intended to maintain the technology base into the time period (mid-1990s) when a private sector demonstration could be initiated. (author)

  12. LIQUID METAL COMPOSITIONS CONTAINING URANIUM

    Science.gov (United States)

    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.

  13. Calculation of aerodynamics of aerosol filter designs for cleaning of heavy liquid metal cooler reactor gas loops

    International Nuclear Information System (INIS)

    Valery P Melnikov; Pyotr N Martynov; Albert K Papovyants; Ivan V Yagodkin

    2005-01-01

    Full text of publication follows: One of the basic performances of aerosol filters is the aerodynamic resistance to the flow of gaseous medium to be cleaned. Calculation of the aerodynamics of aerosol filters in reference to the gas loops of reactor installations with heavy liquid metal coolant (HLMC) allows the design of the structural components of filters to be optimized to provide minimum initial resistance values. It is established that owing to various factors aerosol particles of different concentration and disperse composition are present always in the gas spaces of heavy liquid metal cooled reactor gas loops. To prevent the negative effect of aerosols on the equipment of the gas loops, it is reasonable to use filters of multistep design with sections of preliminary and fine cleaning to catch micron and submicron particles, respectively. A computer program and technique have been developed to evaluate the aerodynamics of folded aerosol filters for different parameters of their structural components, taking account of the aerosol spectrum and concentration. The algorithm of the calculation is presented by the example of a two-step design assembled in single vessel; the filter dimensions and pattern of the air flow to be cleaned are determined under the given boundary conditions. The evaluation of the aerodynamic resistance of filters was performed with consideration for local resistances and resistances of all the structural components of the filter (sudden constriction, expansion, the flow in air channels, filtering material and so on). Correlations have been derived for the resistance of air channels, filtering materials of preliminary and fine cleaning sections as a function of such parameters as the section depth (50-500 mm), the height of separators (3,5-20 mm), the filtering surface area (1,5-30 m 2 ). Based on the calculation results, the auto-similarity domain was brought out for the minimal values of filter resistances as a function of the ratio of

  14. Technical Meeting on Passive Shutdown Systems for Liquid Metal-Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2015-01-01

    A major focus of the design of modern fast reactor systems is on inherent and passive safety. Specific systems to improve reactor safety performance during accidental transients have been developed in nearly all fast reactor programs, and a large number of proposed systems have reached various stages of maturity. This Technical Meeting on Passive Shutdown Systems for Fast Reactors, which was recommended by the Technical Working Group on Fast Reactors (TWG-FR), addressed Member States’ expressed need for information exchange on projects and programs in the field, as well as for the identification of priorities based on the analysis of technology gaps to be covered through R&D activities. This meeting was limited to shutdown systems only, and did not include other passive features such as natural circulation decay heat removal systems etc.; however the meeting catered to passive shutdown safety devices applicable to all types of fast neutron systems. It was agreed to initiate a new study and produce a Nuclear Energy Series (NES) Technical Report to collect information about the existing operational systems as well as innovative concepts under development. This will be a useful source for member states interested in gaining technical expertise to develop passive shutdown systems as well as to highlight the importance and development in this area

  15. The case for developing the liquid metal fast breeder reactor in France

    International Nuclear Information System (INIS)

    Baumier, J.; Duchatelle, L.

    1984-01-01

    An analysis is presented of the economic aspects of electricity generation by fast breeder reactor (FBR) plants, with particular reference to FBRs planned in France during the 1980s, and in comparison with French pressurized water reactor (PWR) units. Individual sections deal with the following: an outline of the bases of comparison between FBRs and PWRs, describing methods used to calculate costs; a description of the FBRs and PWRs concerned, defining and explaining the concepts of breeding gain and doubling time for FBRs, and calculating the energy extracted from uranium by the two types of reactor; the results of analysis and FBR/PWR comparison for capital, operating, fuel cycle and total generating costs, drawing conclusions for the competitivity of FBRs; general conclusions on the advisability of large scale FBR deployment. (U.K.)

  16. Status of liquid metal reactor development in the United States of America

    International Nuclear Information System (INIS)

    Griffith, J.D.; Horton, K.E.

    1991-01-01

    An existing network of government and industry research facilities and engineering test centers in the United States is currently providing test capabilities and the technical expertise required to conduct an aggressive advanced reactor development program. Subsequent to the directive to shut down the Fast Flux Test Facility in early 1990, a variety of activities were undertaken to provide support for continued operation. The United States has made substantial progress in achieving ALMR program objectives. The metal fuel cycle is designed to recycle and burn its own actiniums, and has the potential to be a very effective burner of actiniums generated in the LWRs. The current emphasis in the IFR Program is on the comprehensive development of the IFR (Integral Fast Reactor) technology, to be followed by a period of technology demonstration which would verify the economic feasibility of the concept. The United States has been active in international cooperative activities in the fast reactor sector since 1969. (author). 11 figs, 1 tab

  17. Selection of engineering materials and fabrication of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Patriarca, P.

    1975-01-01

    Information is presented graphically and pictorially concerning the need for nuclear power; basic nuclear concepts including BWR, PWR, HTGR, and LMFBR; the fissioning process; nuclear reactor fuel; fabrication of reactor vessels for LMFBR's; fabrication of intermediate heat exchangers for LMFBR's; piping fabrication for LMFBR's; transition welds; steam generators for LMFBR demonstration plants worldwide; stress corrosion cracking of steam generator materials and weldments; post--test examination of the Alco/BLH sodium-heated steam generator; alternate steam generator designs; and alternate structural materials. (DCC)

  18. Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2005-09-15

    The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was {approx}50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology.

  19. Assessment of the Dry Processed Oxide Fuel in Liquid Metal Fast Reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2005-09-01

    The neutronic feasibility of the dry process oxide fuel was assessed for the sodium-cooled and lead-cooled fast reactors (SFR and LFR, respectively), which were recommended as Generation-IV (Gen-IV) reactor systems by the Gen-IV international forum. The reactor analysis was performed for the equilibrium fuel cycle of two core configurations: Hybrid BN-600 benchmark core with an enlarged lattice pitch and a modified BN-600 core. The dry process technology assumed in this study is the molten-salt process, which was developed by Russian scientists for recycling oxide fuels. The core calculation was performed by the REBUS-3 code and the reactor characteristics such as the transuranic (TRU) enrichment, breeding ratio, peak linear power, burnup reactivity swing, etc. were calculated for the equilibrium core under a fixed fuel management scheme. The results showed that a fissile self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed. If the design criteria used in this study is proved to be acceptable through a detailed physics design and thermal hydraulic analysis in the future, it is practically possible to construct an equilibrium fuel cycle of the SFR and LFR systems based on the oxide fuel by utilizing the dry process technology

  20. The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

    Science.gov (United States)

    Mosunova, N. A.

    2018-05-01

    The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

  1. Safety of the liquid-metal cooled fast breeder reactor and aspects of its fuel cycle

    International Nuclear Information System (INIS)

    Kessler, G.; Papp, R.; Huebel, D.

    1977-01-01

    Design and construction of the sodium-cooled fast reactors KNK-II (20MW(e)) and SNR-300 (300MW(e)) determine the status of safety engineering and safety R and D of LMFBRs in the Federal Republic of Germany. Both prototype fast power reactors have to go through a civil licensing process similar to that applied to present LWRs. A multilevel safety - or defence in depth - approach is applied to the design and construction of fast power reactors. All design data of the fast reactor plant are confirmed by extensive experimental programmes. Design limits of the plant are thoroughly discussed during the licensing process. Important safety R and D programmes have been and are still being performed. A very conservative safety analysis for hypothetical core and other plant accidents is used for present prototype fast reactors. The paper reviews the future trend of development of theoretical methods for accident analysis and the application of experimental results, especially in view of large commercial-type LMFBRs. The safety approach applied to the LMFBR plant is safe operation under normal operating conditions and safe shutdown under off-normal conditions. The consequences of releases of radioactivity to the environment meet the given standards. No chemical reprocessing plant for fast breeder fuel is in operation in the FRG at present; however, R and D work on investigation of all aspects and problem areas of the fast breeder fuel cycle are under way. Systems studies on safety aspects of the fast breeder fuel cycle (transport, reprocessing, fuel fabrication) and its impact on the environment have been performed and the main consequences of these studies are presented in the paper. (author)

  2. Benchmark exercise for fluid flow simulations in a liquid metal fast reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E., E-mail: emerzari@anl.gov [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Fischer, P. [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Yuan, H. [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL (United States); Van Tichelen, K.; Keijers, S. [SCK-CEN, Boeretang 200, Mol (Belgium); De Ridder, J.; Degroote, J.; Vierendeels, J. [Ghent University, Ghent (Belgium); Doolaard, H.; Gopala, V.R.; Roelofs, F. [NRG, Petten (Netherlands)

    2016-03-15

    Highlights: • A EUROTAM-US INERI consortium has performed a benchmark exercise related to fast reactor assembly simulations. • LES calculations for a wire-wrapped rod bundle are compared with RANS calculations. • Results show good agreement for velocity and cross flows. - Abstract: As part of a U.S. Department of Energy International Nuclear Energy Research Initiative (I-NERI), Argonne National Laboratory (Argonne) is collaborating with the Dutch Nuclear Research and consultancy Group (NRG), the Belgian Nuclear Research Centre (SCK·CEN), and Ghent University (UGent) in Belgium to perform and compare a series of fuel-pin-bundle calculations representative of a fast reactor core. A wire-wrapped fuel bundle is a complex configuration for which little data is available for verification and validation of new simulation tools. UGent and NRG performed their simulations with commercially available computational fluid dynamics (CFD) codes. The high-fidelity Argonne large-eddy simulations were performed with Nek5000, used for CFD in the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite. SHARP is a versatile tool that is being developed to model the core of a wide variety of reactor types under various scenarios. It is intended both to serve as a surrogate for physical experiments and to provide insight into experimental results. Comparison of the results obtained by the different participants with the reference Nek5000 results shows good agreement, especially for the cross-flow data. The comparison also helps highlight issues with current modeling approaches. The results of the study will be valuable in the design and licensing process of MYRRHA, a flexible fast research reactor under design at SCK·CEN that features wire-wrapped fuel bundles cooled by lead-bismuth eutectic.

  3. Italian position paper on the safety analysis of liquid metal fast breeder reactors as related to sodium fires. The PEC reactor

    International Nuclear Information System (INIS)

    Gerosa, A.

    1983-01-01

    To obtain a deep understanding of physical phenomena and engineering problems connected to sodium fires, and to optimize the utilization of human and financial resources available, CNEN (now ENEA) has decided to join the French Commissariat a l'Energie Atomique (CEA) in the realization of a Franco-Italian experimental programme on sodium fires, named ESMERALDA. As for design preventions for PEC reactor (a fast flux, liquid metal cooled, fuel element testing reactor) fundamental choices were made taking into account all available knowledge, but with particular reference to the results of CEA's previous experiments on sodium fires. More detailed design analysis will be possible in the future, based on experimental results coming from the ESMERALDA programme

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  6. Spatial kinetics studies in liquid-metal fast breeder reactor critical assemblies

    International Nuclear Information System (INIS)

    Brumback, S.B.; Goin, R.W.; Carpenter, S.G.

    1988-01-01

    Recent measurements in the zero-power physics reactor have been used to study the effect of spatial decoupling in fast reactor critical assemblies of various sizes and compositions. Flux distributions in these assemblies had varying degrees of sensitivity to perturbations. Decoupling was investigated using rod-drop, boron-oscillator, and noise-coherence techniques, which emphasized different times following perturbations. Equilibrium flux distributions were also measured for subcritical configurations with inserted control rods. For most assemblies, accurate reactivity measurements were obtained by analyzing the power history from a single detector using inverse kinetics methods, assuming an instantaneous efficiency change for the detector. The instantaneous efficiency change assumption broke down, however, in assemblies with zones in which normal plutonium fuel was replaced by /sup 235/U fuel or fuel with a high /sup 240/Pu content. Flux redistributions caused by perturbations in these cores took several minutes to evolve

  7. System for bearing a nuclear reactor vessel cooled by liquid metal

    International Nuclear Information System (INIS)

    Mahe, A.; Jullien, G.

    1976-01-01

    The invention relates to a bearing system for supporting a nuclear reactor vessel of the kind which is suspended from the reactor closure slab. The bearing system comprises a ring connected at one end to a collar and at the other end to two collars. The collar connected to the bottom end of the ring forms the top part of the vessel to be supported while the other two collars fit into the slab at two separate places. The ring and collars are disposed in an annular space formed in the slab and dividing it into two parts, i.e., a central part and a peripheral part surrounding the central part of the slab

  8. Evaluation of the decay heat removal capability using the concept of a thermosyphon in the liquid metal reactor

    International Nuclear Information System (INIS)

    Kim, Y. S.; Sim, Y. S.; Kim, W. K.

    2000-01-01

    A study related to understand the characteristics of the heat pipe and thermosyphon was performed to evaluate their applicabilities to the current PSDRS (Passive Safety Decay heat Removal System) in the KALIMER (Korea Advanced LIquid MEtal Reactor) design. The possible heat transfer rate by the heat pipe and thermosyphon was reviewed to compare the required capability in the PSDRS. A quantitative comparison was done between the current PSDRS and the modified PSDRS with the thermosyphon. The result showed the dominant heat transfer rate in the air channel, e.g. radiation or convection, is different from each other. The total heat transfer rate is not sensitive to the operating temperature of the thermosyphon. The heat removal by the air in the modified case is relatively reduced and the resultant outlet temperature appears less than above 10 .deg. C. A reversal heat transfer between the air and the thermosyphon may exist near the exit of the active heat transfer region. The total heat transfer rate by the modified case showed about 20∼40% increase relative to the reference one

  9. Hydrodynamic problems of heavy liquid metal coolants technology in loop-type and mono-block-type reactor installations

    International Nuclear Information System (INIS)

    Orlov, Yuri I.; Efanov, Alexander D.; Martynov, Pyotr N.; Gulevsky, Valery A.; Papovyants, Albert K.; Levchenko, Yuri D.; Ulyanov, Vladimir V.

    2007-01-01

    In the report, the influence of hydrodynamics of the loop with heavy liquid metal coolants (Pb and Pb-Bi) on the realization methods and efficiency of the coolant technology for the reactor installations of loop, improved loop and mono-block type of design has been studied. The last two types of installations, as a rule, are characterized by the following features: availability of loop sections with low hydraulic head and low coolant velocities, large squares of coolant free surfaces; absence of stop and regulating valve, auxiliary pumps on the coolant pumping-over lines. Because of the different hydrodynamic conditions in the installation types, the tasks of the coolant technology have specific solutions. The description of the following procedures of coolant technology is given in the report: purification by hydrogen (purification using gas mixture containing hydrogen), regulation of dissolved oxygen concentration in coolant, coolant filtrating, control of dissolved oxygen concentration in coolant. It is shown that change of the loop design made with economic purpose and for improvement of the installation safety cause additional requirements to the procedures and apparatuses of the coolant technology realization

  10. Measurement strategy of the water leakage into a low pressure sodium boundary for a liquid metal reactor

    International Nuclear Information System (INIS)

    Hur, S.; Kim, D.H.; Seong, S.H.; Kim, S.O.

    2004-01-01

    This paper deals with the measurement strategy of a water leakage into sodium boundary for a liquid metal reactor. There are several methods including the chemical sensing method, pressure sensing methods, and non-destructive method including the acoustic monitoring technique to measure the leakage. As for the results of the analysis with respect to the event propagation characteristics, it has been recommended that the acoustic method has a capability to detect small and intermediate leaks within required response time of 10 seconds. A leak of one gram/sec could be currently detected within the required response time of 10 seconds with a high reliability. In the case of less than a one gram/sec leakage, the response time could not meet our requirements due to a complicated signal processing logic. Thus, the system configuration for a fast processing of a leak detection has been recommended. It is expected that this configuration of the leak detection system could reduce the response time due to the distributed and parallel processing scheme. (orig.)

  11. A new balance-of-plant model for the SASSYS-1 LMR [liquid metal reactor] systems analysis code

    International Nuclear Information System (INIS)

    Briggs, L.L.

    1989-01-01

    A balance-of-plant (BOP) model has been developed for use within the SASSYS-1 liquid-metal reactor systems analysis code. This model expands the scope of SASSYS-1 so that the code can explicitly model the waterside components of a nuclear power plant; previously, only the water side of the steam generators could be modeled, with the remainder of the water side represented by boundary conditions on the steam generator. The model represents the BOP a set of flow paths and path junctions; the mass and energy equations are solved at the junctions, and the momentum equation is solved along the flow paths. The junctions are thus mass and energy cells, and the paths are momentum cells. The various waterside component models (pumps, valves, etc.) are specialized types of energy or momentum cells, as appropriate. The solution scheme implicitly couples the energy cells through the momentum cells and solves simultaneously for pressures and enthalpies within the energy cells and for flows within the momentum cells

  12. Dynamical analysis on carbon transfer in liquid metal cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Kataoka, Tadayuki; Matsumoto, Keishi

    1979-01-01

    The dynamical analysis was undertaken on the exchange of carbon taking place between the structural steels and sodium for the case of a bi-metallic secondary system constituted of type 304 stainless and 2 1/4Cr-1Mo steels, representing the secondary system of a liquid sodium cooled fast breeder reactor. The analysis brought to light the effects to be expected on the long terms carbon transfer behavior of: (a) the surface areas of structural steels in contact with flowing sodium, (b) the thickness of the sodium-boundary layer, (c) the initial carbon concentration in the sodium, and (d) the rate of carbon contamination of the sodium. (author)

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

    International Nuclear Information System (INIS)

    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

  14. Review of the SIMMER-II analyses of liquid-metal-cooled fast breeder reactor core-disruptive accident fuel escape

    International Nuclear Information System (INIS)

    DeVault, G.P.; Bell, C.R.

    1985-01-01

    Early fuel removal from the active core of a liquid-metal-cooled fast breeder reactor undergoing a core-disruptive accident may reduce the potential for large energetics resulting from recriticalities. This paper presents a review of analyses with the SIMMER-II computer program of the effectiveness of possible fuel escape paths. Where possible, how SIMMER-II compares with or is validated against experiments that simulated the escape paths also is discussed

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

    International Nuclear Information System (INIS)

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

    1995-01-01

    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

  16. Computational Analysis of Supercritical Carbon Dioxide Gas Turbine for Liquid Metal Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Wi S.; Suh, Kune Y. [Seoul National University, Seoul (Korea, Republic of)

    2008-10-15

    Energy demands at a remote site are increased as the world energy requirement diversifies so that they should generate power on their own site. A Small Modular Reactor (SMR) becomes a viable option for these sites. Generally, the economic feasibility of a high power reactor is greater than that for SMR. As a result the supercritical fluid driven Brayton cycle is being considered for a power conversion system to increase economic competitiveness of SMR. The Brayton cycle efficiency is much higher than that for the Rankine cycle. Moreover, the components of the Brayton cycle are smaller than Rankine cycle's due to high heat capacity when a supercritical fluid is adopted. A lead (Pb) cooled SMR, BORIS, and a supercritical fluid driven Brayton cycle, MOBIS, are being developed at the Seoul National University (SNU). Dostal et al. have compared some advanced power cycles and proposed the use of a supercritical carbon dioxide (SCO{sub 2}) driven Brayton cycle. According to their suggestion SCO{sub 2} is adopted as a working fluid for MOBIS. The turbo machineries are most important components for the Brayton cycle. The turbo machineries of Brayton cycle consists of a turbine to convert kinetic energy of the fluid into mechanical energy of the shaft, and a compressor to recompress and recover the driving force of the working fluid. Therefore, turbine performance is one of the pivotal factors in increasing the cycle efficiency. In MOBIS a supercritical gas turbine is designed in the Gas Advanced Turbine Operation (GATO) and analyzed in the Turbine Integrated Numerical Analysis (TINA). A three-dimensional (3D) numerical analysis is employed for more detailed design to account for the partial flow which the one-dimensional (1D) analysis cannot consider.

  17. Computational Analysis of Supercritical Carbon Dioxide Gas Turbine for Liquid Metal Cooled Reactor

    International Nuclear Information System (INIS)

    Jeong, Wi S.; Suh, Kune Y.

    2008-01-01

    Energy demands at a remote site are increased as the world energy requirement diversifies so that they should generate power on their own site. A Small Modular Reactor (SMR) becomes a viable option for these sites. Generally, the economic feasibility of a high power reactor is greater than that for SMR. As a result the supercritical fluid driven Brayton cycle is being considered for a power conversion system to increase economic competitiveness of SMR. The Brayton cycle efficiency is much higher than that for the Rankine cycle. Moreover, the components of the Brayton cycle are smaller than Rankine cycle's due to high heat capacity when a supercritical fluid is adopted. A lead (Pb) cooled SMR, BORIS, and a supercritical fluid driven Brayton cycle, MOBIS, are being developed at the Seoul National University (SNU). Dostal et al. have compared some advanced power cycles and proposed the use of a supercritical carbon dioxide (SCO 2 ) driven Brayton cycle. According to their suggestion SCO 2 is adopted as a working fluid for MOBIS. The turbo machineries are most important components for the Brayton cycle. The turbo machineries of Brayton cycle consists of a turbine to convert kinetic energy of the fluid into mechanical energy of the shaft, and a compressor to recompress and recover the driving force of the working fluid. Therefore, turbine performance is one of the pivotal factors in increasing the cycle efficiency. In MOBIS a supercritical gas turbine is designed in the Gas Advanced Turbine Operation (GATO) and analyzed in the Turbine Integrated Numerical Analysis (TINA). A three-dimensional (3D) numerical analysis is employed for more detailed design to account for the partial flow which the one-dimensional (1D) analysis cannot consider

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  19. Social and ethical aspects of the liquid-metal fast breeder reactor

    International Nuclear Information System (INIS)

    Lovins, A.B.

    1975-01-01

    Development of liquid fast breeder reactors not only indirectly entails (through commitments of time and resources that foreclose other options), but also directly entails large-scale centralized electrification. The massive economic commitments of such a policy, wether or not it is a nuclear policy, demand and cause major social changes, bypass traditional market mechanisms, concentrate political and economic power, persistently distort political structures and social priorities, compromise professional ethics, are probably inimical to greater distributional equity within and among nations, enhance vulnerability and the paramilitarization of civilian life, introduce major economic and social risks, and reinforce current trends toward centrifugal politics. Deployment of fission technology produces further social and ethical problems, since attempts to reduce potential hazards from operating accidents, from escape of nuclear wastes, or from nuclear violence and coercion will have socio-political side-effects even if they succeed, not to mention the side-effects if they fail. These side-effects, many of which would be worse with fast than with thermal reactors, include repressiveness, abrogation of civil liberties, social rigidity and homogeneity, elitist technocracy, dirigiste autarchy, and suppression of ethical objections. The inability of modern political institutions to cope with the persistent hazards of toxic and explosive nuclear materials strains the competence and perceived legitimacy of those institutions as they try to compromise between individual liberties and public safety and to subject to democratic decision technically tinged policy questions that turn largely on unknown or unknowable information. There is no scientific basis for calculating the likelihood on the maximum long-term of nuclear mishaps, nor for guaranteeing that the effects will not exceed a particular level; it is only known that all precautions are, for fundamental reasons

  20. Compensation of equipment housing elements of reactor units with heavy liquid metal coolant vessel temperature deformations

    International Nuclear Information System (INIS)

    Lebedevich, V.; Ahmetshin, M.; Mendes, D.; Kaveshnikov, S.; Vinogradov, A.

    2015-01-01

    In Russia a lot of different versions of fast reactors (FRs) are investigated and one of these is FR cooled by liquid lead and liquid lead-bismuth alloy. In this poster we are interested by FR with concrete vessel; its components are placed in cavities inside the vessel, and connected by a channel system. During the installation the equipment components are placed in several equipment housings. Between these housings there are cavities with coolant. The alignment of the housings should be provided. It can be broken by irregular concrete vessel heating during FR starting or other transition regimes. Our goal is to suggest a list of designing steps to compensate temperature deformations of equipment housing elements. A simplified model of equipment housing was suggested. It consists of two cylinders - tunnels in the concrete vessel, separated by a cavity filled by coolant and inert gas. The bottom part was considered as heated to 420 C. degrees while in the top part temperature decreased to 45 C. degrees (on the concrete surface). According to this data, results show that temperature gradient leads to a concrete layer dislocation of about 12.5 mm, which can lead to damage and breaking alignment. We propose the following solution to compensate for temperature deformation: -) to chisel out part of the upper top of the insulating concrete; -) to install an adequate misalignment of equipment housing elements preliminary; and -) to use a torsion system like a piston-type device for providing additional strength in order to compensate deformation and vibrations

  1. Liquid metal reactor development -Studies on safety measure of LMR coolant

    International Nuclear Information System (INIS)

    Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung

    1995-07-01

    A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author)

  2. Liquid metal reactor development -Studies on safety measure of LMR coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Sung Tae; Choi, Yoon Dong; Park, Jin Hoh; Kwon, Sun Kil; Choi, Jong Hyun; Cho, Byung Ryul; Kim, Tae Joon; Kwon, Sang Woon; Jung, Kyung Chae; Kim, Byung Hoh; Hong, Soon Bok; Jung, Ji Yung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    A study on the safety measures of LMR coolant showed the results as follows; 1. LMR coolant safety measure. A. Analysis and improvement of sodium fire code. B. Analysis of sodium fire phenomena. 2. Sodium fire aerosol characteristics. It was carried out conceptual design and basic design for sodium fire facility of medium size composed of sodium supply tank, sodium reactor vessel, sodium fire aerosol filter system and scrubbing column, and drain tank etc. 3. Sodium purification technology. A. Construction of calibration loop. (1) Design of sodium loop for the calibration of the equipment. (2) Construction of sodium loop including test equipments and other components. B. Na-analysis technology. (1) Oxygen concentration determination by the wet method. (2) Cover gas purification preliminary experiment. 4. The characteristics of sodium-water reaction. A. Analysis of the micro and small leak phenomena. (1) Manufacture of the micro-leak test apparatus. B. Analysis of large leak events. (1) Development of preliminary code for analysis of initial spike pressure. (2) Sample calculation and comparison with previous works. C. Development of test facility for large leak event evaluation. (1) Conceptional and basic design for the water and sodium-water test facility. D. Technology development for water leak detection system. (1) Investigations for the characteristics of active acoustic detection system. (2) Testing of the characteristics of hydrogen leak detection system. 171 figs, 29 tabs, 3 refs. (Author).

  3. Sodium/water reactions in steam generators of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Hori, M.

    1980-01-01

    The status of the research and development on sodium/water reactions resulting from the leakage of water into sodium in LMFBR steam generators is reviewed. The importance of sodium/water reaction phenomena in the design and operation of steam generators is discussed. The effects of sodium/water reactions are evaluated and methods of protection against these phenomena are surveyed. The products of chemical reactions between sodium and water under steam generator conditions are H 2 , NaOH, Na 2 O and NaH. Together with the temperature rise due to the associated exothermic reaction, these reaction products cause effects such as self-wastage, single- and multi-target wastage, and rapid pressure increase, depending on the size of the leak hole or the magnitude of leak rate. As for the wastage phenomena of small leaks, the effects of various factors have been studied and experimental correlations, as well as some theoretical work, have been performed. To investigate the pressure phenomena of a large leak, large-scale tests have been conducted by various organizations, and the computer codes to analyse these phenomena have been developed and verified by experiments. In the design of steam generators, an initial failure up to a hypothetical double-ended guillotine rupture of a single heat transfer tube is widely used as the design basis leak. Protection systems for LMFBR plants consist of leak detection devices, leak termination devices, and reaction pressure relief devices. From analyses based on research and development activities, as well as from experience with leaks in steam generator test loops and reactor plants, it has been confirmed that protection systems can satisfactorily be designed to accommodate leak incidents in LMFBR plants. (author)

  4. Actively convected liquid metal divertor

    International Nuclear Information System (INIS)

    Shimada, Michiya; Hirooka, Yoshi

    2014-01-01

    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. (letter)

  5. An innovative liquid metal design with worldwide application potential

    International Nuclear Information System (INIS)

    Quinn, J.E.; Berglund, R.C.

    1989-01-01

    This paper reports that the United States nuclear program has been faced with major political, economic and technical challenges in recent years. One US program element, the Liquid Metal Reactor, has addressed these challenges in a systematic, focused manner. The result is an innovative modular design incorporating safety features that utilize inherent characteristics. This Advanced Liquid Metal Reactor (ALMR) is based on the PRISM concept, originated by the General Electric Company in 1981. This design should also be attractive in other developed countries that have deployed, and/or are deploying, nuclear power. The design's safety features can achieve neutronic shutdown and decay heat removal without relying on operator action or engineered active safety features. The ALMR utilizes many innovations including: a passive reactor vessel air cooling system for decay heat removal; the use of a sealed reactor assembly; seismic isolation; electromagnetic primary pumps; and an in-vessel fuel transfer machine. The US ALMR design incorporates a metal fuel core as its reference, however, the required safety performance can also be achieved with an oxide core having similar safety features. This flexibility is particularly important when addressing world wide ALMR applications. The reference ALMR reactor module, of which there are nine in a typical 1395 MW e plant, has a -6 meter by -20 meter vessel and a 471 MW thermal output, with a reactor outlet temperature of 485 degrees C and an overall conversion efficiency of 33%. This plant uses a saturated steam cycle and a non-safety grade secondary sodium system

  6. Multipurpose sampler device for liquid metal

    International Nuclear Information System (INIS)

    Nelson, P.A.; Kolba, V.M.; Holmes, J.T.

    1975-01-01

    A device for collecting samples or examining a flow of liquid metal is provided for use with such as a liquid-metal-cooled nuclear reactor. The sampler device includes a casing surrounded by an external heater for establishing an upper isothermal zone and a lower zone for heating the entering liquid metal. One of various inserts is suspended into the isothermal zone where it is surrounded by a shroud tube for directing liquid-metal flow from the heating zone into the top of the insert. Discharge flow from the insert gravitates through a helically wound tube in heat exchange contact with entering liquid-metal flow within the heating zone. The inserts comprise an overflow cup with upper and lower freeze seals, a filter for removing particulate matter, and a fixture for maintaining various sample materials in equilibrium with liquid-metal flow. (U.S.)

  7. Liquid metal heat transfer issues

    International Nuclear Information System (INIS)

    Hoffman, H.W.; Yoder, G.L.

    1984-01-01

    An alkali liquid metal cooled nuclear reactor coupled with an alkali metal Rankine cycle provides a practicable option for space systems/missions requiring power in the 1 to 100 MW(e) range. Thermal issues relative to the use of alkali liquid metals for this purpose are identified as these result from the nature of the alkali metal fluid itself, from uncertainties in the available heat transfer correlations, and from design and performance requirements for system components operating in the earth orbital microgravity environment. It is noted that, while these issues require further attention to achieve optimum system performance, none are of such magnitude as to invalidate this particular space power concept

  8. Liquid metal cold trap

    International Nuclear Information System (INIS)

    Hundal, R.

    1976-01-01

    A cold trap assembly for removing impurities from a liquid metal is described. A hole between the incoming impure liquid metal and purified outgoing liquid metal acts as a continuous bleed means and thus prevents the accumulation of cover gases within the cold trap assembly

  9. FLIT: Flowing LIquid metal Torus

    Science.gov (United States)

    Kolemen, Egemen; Majeski, Richard; Maingi, Rajesh; Hvasta, Michael

    2017-10-01

    The design and construction of FLIT, Flowing LIquid Torus, at PPPL is presented. FLIT focuses on a liquid metal divertor system suitable for implementation and testing in present-day fusion systems, such as NSTX-U. It is designed as a proof-of-concept fast-flowing liquid metal divertor that can handle heat flux of 10 MW/m2 without an additional cooling system. The 72 cm wide by 107 cm tall torus system consisting of 12 rectangular coils that give 1 Tesla magnetic field in the center and it can operate for greater than 10 seconds at this field. Initially, 30 gallons Galinstan (Ga-In-Sn) will be recirculated using 6 jxB pumps and flow velocities of up to 10 m/s will be achieved on the fully annular divertor plate. FLIT is designed as a flexible machine that will allow experimental testing of various liquid metal injection techniques, study of flow instabilities, and their control in order to prove the feasibility of liquid metal divertor concept for fusion reactors. FLIT: Flowing LIquid metal Torus. This work is supported by the US DOE Contract No. DE-AC02-09CH11466.

  10. CONTAIN LMR/1B-Mod.1, A computer code for containment analysis of accidents in liquid-metal-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Murata, K.K.; Carroll, D.E.; Bergeron, K.D.; Valdez, G.D.

    1993-01-01

    The CONTAIN computer code is a best-estimate, integrated analysis tool for predicting the physical, chemical, and radiological conditions inside a nuclear reactor containment building following the release of core material from the primary system. CONTAIN is supported primarily by the U. S. Nuclear Regulatory Commission (USNRC), and the official code versions produced with this support are intended primarily for the analysis of light water reactors (LWR). The present manual describes CONTAIN LMR/1B-Mod. 1, a code version designed for the analysis of reactors with liquid metal coolant. It is a variant of the official CONTAIN 1.11 LWR code version. Some of the features of CONTAIN-LMR for treating the behavior of liquid metal coolant are in fact present in the LWR code versions but are discussed here rather than in the User's Manual for the LWR versions. These features include models for sodium pool and spray fires. In addition to these models, new or substantially improved models have been installed in CONTAIN-LMR. The latter include models for treating two condensables (sodium and water) simultaneously, sodium atmosphere and pool chemistry, sodium condensation on aerosols, heat transfer from core-debris beds and to sodium pools, and sodium-concrete interactions. A detailed description of each of the above models is given, along with the code input requirements

  11. Liquid metal fast breeder reactor steam generator survey of the consequences of large scale sodium water reaction

    International Nuclear Information System (INIS)

    Vambenepe, G.

    1978-01-01

    The ''Retona'' three-dimensional hydrodynamic computing code is being developed by Electricity de France to survey the consequences, on the very plant, of a large scale sodium water reaction in liquid metal steam generators. In this communication, the heat-exchanger geometry is schematized and the problem solving process briefly described under assumed simplifying hypotheses. The application of the results to the Creusot-Loire steam generator selected for Super-Phenix are given as an example. (author)

  12. Liquid metal reactors. Hearing before the Subcommittee on Energy and Power of the Committee on Energy and Commerce, House of Representatives, One Hundred Third Congress, First Session, June 9, 1993

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    This hearing reviews the Liquid Metal Reactor Program of the US DOE, asking the question of whether this program is necessary to the future of nuclear power and whether it has any serious application to the disposal of high-level radioactive waste to justify the probable expense of the program. The liquid metal reactor would rely on plutonium as a fuel, produced by reprocessing of spent fuel wastes or by using the reactor as a breeder. Also discussed are the proliferation risks of a Liquid Metal Reactor. Principal testifiers include: David, Institute for Science and International Security; Barrett and Brolin, DOE; Coops, American Nuclear Society; Kratzer, independent consultant; Lancelot, National Taxpayers Union; Larsen, Oil, Chemical and Atomic Workers Union; Till, ANL; von Hippel, Princeton University

  13. Scoping calculations for design and analysis of large reactor vessels for liquid-metal fast breeder reactor (LMFBR) plants

    International Nuclear Information System (INIS)

    Fiala, C.; Kulak, R.F.; Ma, D.C.; Pan, Y.C.; Seidensticker, R.W.; Wang, C.Y.; Zeuch, W.R.

    1982-01-01

    Reactor vessels for commercial-sized LMFBR plants are quite large - ranging 40 to 70 ft in diameter and 50 to 70 ft in overall depth. These stainless steel vessels contain liquid sodium at relatively low pressures, but at high temperatures. The resulting thin-walled vessels present the structural designer and analyst with special problems, particularly in providing a balanced design to accommodate seismic loads, design basis accident loads, and thermal loadings. A comprehensive set of scoping calculations - though preliminary in detail and depth of design - provides substantial guidance to the vessel designer for subsequent design iterations. Emphasis is placed on the analysis of the large-diameter top closure of the vessel - the deck structure

  14. The PRISM reactor as a possible option to deal with the british civilian plutonium stockpile

    Energy Technology Data Exchange (ETDEWEB)

    Fichtlscherer, Christopher [IANUS, TU Darmstadt (Germany); Friess, Friederike [IANUS, TU Darmstadt (Germany); ISR, Universitaet fuer Bodenkultur Wien (Boku) (Austria)

    2017-07-01

    Dealing with stocks of separated weapon-usable plutonium is a big challenge for our modern society. This work focuses on the British civil plutonium stockpiles, which amount to 103.3 tons. One option is seen in irradiating the plutonium in a fast reactor under development, namely the GE PRISM reactor. The PRISM reactor is a small modular, fast reactor which has a thermal power of 840 MW and an electrical output of 311 MW. It is intended to use MOX fuel and proponents claim, that it thus would be possible to produce clean energy, while making the plutonium proliferation resistant. A MCNP model of the reactor is built and depletion calculations with different target burnups of the fuel were conducted to check whether the burned material would fulfil the Spent-Fuel Standard. Particularly it was checked whether the spent fuel is self protecting, meaning that the dose rate does not fall below a limit of 1 Sv/h in 1 meter distance after a cooling period of 30 years. Based on the reactor model calculations the irradiation time to fulfill this limit for the spent fuel is calculated. Based on the needed target burnup, it can be verified, whether it is possible for the PRISM reactor to render the civil plutonium proliferation resistant in only 20 years as is is claimed by its proponents.

  15. Liquid metal steam generator

    International Nuclear Information System (INIS)

    Wolowodiuk, W.

    1975-01-01

    A liquid metal heated steam generator is described which in the event of a tube failure quickly exhausts out of the steam generator the products of the reaction between the water and the liquid metal. The steam is generated in a plurality of bayonet tubes which are heated by liquid metal flowing over them between an inner cylinder and an outer cylinder. The inner cylinder extends above the level of liquid metal but below the main tube sheet. A central pipe extends down into the inner cylinder with a centrifugal separator between it and the inner cylinder at its lower end and an involute deflector plate above the separator so that the products of a reaction between the liquid metal and the water will be deflected downwardly by the deflector plate and through the separator so that the liquid metal will flow outwardly and away from the central pipe through which the steam and gaseous reaction products are exhausted. (U.S.)

  16. Liquid Metal Transformers

    OpenAIRE

    Sheng, Lei; Zhang, Jie; Liu, Jing

    2014-01-01

    The room temperature liquid metal is quickly emerging as an important functional material in a variety of areas like chip cooling, 3D printing or printed electronics etc. With diverse capabilities in electrical, thermal and flowing behaviors, such fluid owns many intriguing properties that had never been anticipated before. Here, we show a group of unconventional phenomena occurring on the liquid metal objects. Through applying electrical field on the liquid metals immersed in water, a series...

  17. Liquid metal thermal-hydraulics

    International Nuclear Information System (INIS)

    Kottowski-Duemenil, H.M.

    1994-01-01

    This textbook is a report of the 26 years activity of the Liquid Metal Boiling Working Group (LMBWG). It summarizes the state of the art of liquid metal thermo-hydraulics achieved through the collaboration of scientists concerned with the development of the Fast Breeder Reactor. The first chapter entitled ''Liquid Metal Boiling Behaviour'', presents the background and boiling mechanisms. This section gives the reader a brief but thorough survey on the superheat phenomena in liquid metals. The second chapter of the text, ''A Review of Single and Two-Phase Flow Pressure Drop Studies and Application to Flow Stability Analysis of Boiling Liquid Metal Systems'' summarizes the difficulty of pressure drop simulation of boiling sodium in core bundles. The third chapter ''Liquid Metal Dry-Out Data for Flow in Tubes and Bundles'' describes the conditions of critical heat flux which limits the coolability of the reactor core. The fourth chapter dealing with the LMFBR specific topic of ''Natural Convection Cooling of Liquid Metal Systems''. This chapter gives a review of both plant experiments and out-of-pile experiments and shows the advances in the development of computing power over the past decade of mathematical modelling ''Subassembly Blockages Suties'' are discussed in chapter five. Chapter six is entitled ''A Review of the Methods and Codes Available for the Calculation on Thermal-Hydraulics in Rod-Cluster and other Geometries, Steady state and Transient Boiling Flow Regimes, and the Validation achieves''. Codes available for the calculation of thermal-hydraulics in rod-clusters and other geometries are reviewed. Chapter seven, ''Comparative Studies of Thermohydraulic Computer Code Simulations of Sodium Boiling under Loss of Flow Conditions'', represents one of the key activities of the LMBWG. Several benchmark exercises were performed with the aim of transient sodium boiling simulation in single channels and bundle blockages under steady state conditions and loss of

  18. Liquid metal corrosion considerations in alloy development

    International Nuclear Information System (INIS)

    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

  19. Liquid metals pumping

    International Nuclear Information System (INIS)

    Le Frere, J.P.

    1984-01-01

    Pumps used to pump liquid metals depend on the liquid metal and on the type of application concerned. One deals more particularly with electromagnetic pumps, the main pumps used with mechanical pumps. To pump sodium in the nuclear field, these two types of pumps are used; the pumps of different circuits of Super Phenix are presented and described [fr

  20. Liquid metal monitor

    International Nuclear Information System (INIS)

    Caldwell-Nichols, C.J.; Roach, P.F.

    1982-01-01

    A liquid metal monitor of the by-pass plugging meter kind described in British Patent 1,308,466, is further provided with a pump arranged to oppose flow through a by-pass thereby to provide a constant pressure difference across an orifice and improve the sensitivity of the instrument. The monitor estimates the impurity content in a liquid metal stream. (author)

  1. Transient performance of S-prism

    International Nuclear Information System (INIS)

    Dubberley, A.E.; Boardman, C.E.; Gamble, R.E.; Hiu, M.M.; Lipps, A.J.; Wu, T.

    2001-01-01

    S-PRISM is an advanced Fast Reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test of a single Nuclear Steam Supply System (NSSS) for design certification at minimum cost and risk. Based on the success of the previous DOE sponsored Advanced Liquid Metal Reactor (ALMR) program GE has continued to develop and assess the technical viability and economic potential of an up-rated plant called SuperPRISM (S-PRISM). This paper presents the results of transient analyses performed to assess the ability of S-PRISM to accommodate severe accident initiator events. A unique safety capability of S-PRISM is accommodation of the ''higher probability'' accident initiators that led to core melt accidents in prior large LMRs. These events, the Anticipated Transients Without Scram (ATWS) events, are thus the focus of passive safety confirmation analyses. The events included in this assessment are: Unprotected Loss of Flow, Unprotected Loss of Heat Sink, Unprotected Loss of Flow and Heat sink, Unprotected Transient Overpower and Unprotected Safe Shutdown Earthquake. (author)

  2. Assessment of prism responses to loss of flow events

    International Nuclear Information System (INIS)

    Slovik, G.C.; Van Tuyle, G.J.; Sands, S.

    1992-01-01

    The Nuclear Regulatory Commission (NRC), with Brookhaven national Laboratory providing technical support, is continuing a preapplication review of the 471 MWt, advanced liquid metal reactor (ALMR), PRISM by General Electric. The revised design has been evaluated using the SSC code, for a series of loss of flow events (LOF) with and without gas expansion modules (GEMs). These devices have a net worth of 69 cents and have reduced the seriousness of the LOF in PRISM. However, it was found that the extremely low probability case of an instantaneous loss of 4 EM pumps without scram could lead to sodium boiling even with the GEMs

  3. Data management and communication networks for Man-Machine Interface System in Korea Advanced Liquid MEtal Reactor : its functionality and design requirements

    International Nuclear Information System (INIS)

    Cha, Kyung Ho; Park, Gun Ok; Suh, Sang Moon; Kim, Jang Yeol; Kwon, Kee Choon

    1998-01-01

    The DAta management and Communication NETworks(DACONET), which it is designed as a subsystem for Man-Machine Interface System of Korea Advanced LIquid MEtal Reactor(KALIMER MMIS) and advanced design concept is approached, is described. The DACONET has its roles of providing the real-time data transmission and communication paths between MMIS systems, providing the quality data for protection, monitoring and control of KALIMER and logging the static and dynamic behavioral data during KALIMER operation. The DACONET is characterized as the distributed real-time system architecture with high performance. Future direction, in which advanced technology is being continually applied to Man-Machine Interface System development and communication networks of KALIMER MMIS

  4. Data management and communication networks for Man-Machine Interface System in Korea Advanced Liquid MEtal Reactor : its functionality and design requirements

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Kyung Ho; Park, Gun Ok; Suh, Sang Moon; Kim, Jang Yeol; Kwon, Kee Choon [KAERI, Taejon (Korea, Republic of)

    1998-05-01

    The DAta management and Communication NETworks(DACONET), which it is designed as a subsystem for Man-Machine Interface System of Korea Advanced LIquid MEtal Reactor(KALIMER MMIS) and advanced design concept is approached, is described. The DACONET has its roles of providing the real-time data transmission and communication paths between MMIS systems, providing the quality data for protection, monitoring and control of KALIMER and logging the static and dynamic behavioral data during KALIMER operation. The DACONET is characterized as the distributed real-time system architecture with high performance. Future direction, in which advanced technology is being continually applied to Man-Machine Interface System development and communication networks of KALIMER MMIS.

  5. Data management and communication networks for man-machine interface system in Korea Advanced LIquid MEtal Reactor : Its functionality and design requirements

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Kyung Ho; Park, Gun Ok; Suh, Sang Moon; Kim, Jang Yeol; Kwon, Kee Choon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    The DAta management and COmmunication NETworks(DACONET), which it is designed as a subsystem for Man-Machine Interface System of Korea Advanced LIquid MEtal Reactor (KALIMER MMIS) and advanced design concept is approached, is described. The DACONET has its roles of providing the real-time data transmission and communication paths between MMIS systems, providing the quality data for protection, monitoring and control of KALIMER and logging the static and dynamic behavioral data during KALIMER operation. The DACONET is characterized as the distributed real-time system architecture with high performance. Future direction, in which advanced technology is being continually applied to Man-Machine Interface System development of Nuclear Power Plants, will be considered for designing data management and communication networks of KALIMER MMIS. 9 refs., 1 fig. (Author)

  6. Data management and communication networks for man-machine interface system in Korea Advanced LIquid MEtal Reactor : Its functionality and design requirements

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Kyung Ho; Park, Gun Ok; Suh, Sang Moon; Kim, Jang Yeol; Kwon, Kee Choon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    The DAta management and COmmunication NETworks(DACONET), which it is designed as a subsystem for Man-Machine Interface System of Korea Advanced LIquid MEtal Reactor (KALIMER MMIS) and advanced design concept is approached, is described. The DACONET has its roles of providing the real-time data transmission and communication paths between MMIS systems, providing the quality data for protection, monitoring and control of KALIMER and logging the static and dynamic behavioral data during KALIMER operation. The DACONET is characterized as the distributed real-time system architecture with high performance. Future direction, in which advanced technology is being continually applied to Man-Machine Interface System development of Nuclear Power Plants, will be considered for designing data management and communication networks of KALIMER MMIS. 9 refs., 1 fig. (Author)

  7. Effect of fluid-to-structure heat transfer on the structural damage potential to a liquid-metal fast breeder reactor

    International Nuclear Information System (INIS)

    Hakim, S.J.; Abramson, P.B.

    1979-01-01

    Deterministic calculations simulating a hypothetical accident in a liquid-metal fast breeder reactor that leads to a hydrodynamic disassembly of the core have been carried out to estimate the system's damage potential due to the vapor-pressure-driven expansion of molten core material and its dependency on the heat transfer to the remaining structure. These calculations ignored the effect on the work potential of sodium left in the core during the disassembly. Results indicate that steel cladding in the upper axial blankets and fission gas plenum acts as a thermodynamic energy sink that could reduce the total thermodynamic work energy by between one and two orders of magnitude, provided little or no sodium is present in the core at the time of interaction. These results have been found to be insensitive to the rate of heat transferred from the molten fuel to the molten steel that comprises the molten core material

  8. Simplified analysis of PRISM RVACS [Reactor Vessel Auxiliary Cooling System] performance without liner spill-over

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.

    1990-01-01

    Simplified analysis of the performance of the PRISM RVACS decay heat removal system under off-normal conditions, i.e., without the liner spill-over, is described. Without the spilling of hot-pool sodium over the liner and the resultant down-flow along the inside of the reactor vessel wall, the RVACS system performance becomes dominated by the radial heat condition and radiation. Simple estimates of the resulting heat conduction and radiation processes support GE's contention that the RVACS performance is not severely impacted by the absence of spillover, and can improve significantly if sodium has leaked into the region between the reactor and containment vessels. 7 refs

  9. Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

    International Nuclear Information System (INIS)

    2005-01-01

    This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

  10. Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

  11. Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors. Report of the collaborative project COOL of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-05-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000, based on a resolution by the IAEA General Conference (GC(44)/RES/21). INPRO aims at helping to ensure that nuclear energy is available in the twenty-first century in a sustainable manner, and seeks to bring together all interested Member States, both technology holders and technology users, to jointly consider actions to achieve desired innovations. INPRO is taking care of the specific needs of developing countries. One of the aims of INPRO is to develop options for enhanced sustainability through promotion of technical and institutional innovations in nuclear energy technology through collaborative projects among IAEA Member States. Collaboration among INPRO members is fostered on selected innovative nuclear technologies to bridge technology gaps. Collaborative projects have been selected so that they complement other national and international R and D activities. The INPRO Collaborative Project COOL on Investigation of Technological Challenges Related to the Removal of Heat by Liquid Metal and Molten Salt Coolants from Reactor Cores Operating at High Temperatures investigated the technological challenges of cooling reactor cores that operate at high temperatures in advanced fast reactors, high temperature reactors and accelerator driven systems by using liquid metals and molten salts as coolants. The project was initiated in 2008 and was led by India; experts from Brazil, China, Germany, India, Italy and the Republic of Korea participated and provided chapters of this report. The INPRO Collaborative Project COOL addressed the following fields of research regarding liquid metal and molten salt coolants: (i) survey of thermophysical properties; (ii) experimental investigations and computational fluid dynamics studies on thermohydraulics, specifically pressure drop and heat transfer under different operating conditions; (iii) monitoring and control of coolant

  12. Containment performance of S-prism under severe BDB conditions

    International Nuclear Information System (INIS)

    Boardman, C.E.; Dubberley, A.E.; Hui, M.; Iwashige, K.

    2001-01-01

    S-PRISM is an advanced Fast Reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test of a single Nuclear Steam Supply System (NSSS) for design certification at minimum cost and risk. Based on the success of the previous DOE sponsored Advanced Liquid Metal Reactor (ALMR) program GE has continued to develop and assess the technical viability and economic potential of an up-rated modular Fast Reactor called Super PRISM (S-PRISM). S-PRISM retains all of the key ALMR design features including passive reactor shutdown, passive shutdown heat removal, and passive reactor cavity cooling that were developed under an earlier DOE program. An additional feature of S-PRISM involves the use an innovative containment system that reduces the required design basis containment pressure by a factor of two through the use of a controlled venting system. The performance of this innovative containment system is evaluated and described in this paper. (author)

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

    International Nuclear Information System (INIS)

    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

  14. Experimental study of the large-scale axially heterogeneous liquid-metal fast breeder reactor at the fast critical assembly: Power distribution measurements and their analyses

    International Nuclear Information System (INIS)

    Iijima, S.; Obu, M.; Hayase, T.; Ohno, A.; Nemoto, T.; Okajima, S.

    1988-01-01

    Power distributions of the large-scale axially heterogeneous liquid-metal fast breeder reactor were studied by using the experiment results of fast critical assemblies XI, XII, and XIII and the results of their analyses. The power distributions were examined by the gamma-scanning method and fission rate measurements using /sup 239/Pu and /sup 238/U fission counters and the foil irradiation method. In addition to the measurements in the reference core, the power distributions were measured in the core with a control rod inserted and in a modified core where the shape of the internal blanket was determined by the radial boundary. The calculation was made by using JENDL-2 and the Japan Atomic Energy Research Institute's standard calculation system for fast reactor neutronics. The power flattening trend, caused by the decrease of the fast neutron flux, was observed in the axial and radial power distributions. The effect of the radial boundary shape of the internal blanket on the power distribution was determined in the core. The thickness of the internal blanket was reduced at its radial boundary. The influence of the internal blanket was observed in the power distributions in the core with a control rod inserted. The calculation predicted the neutron spectrum harder in the internal blanket. In the radial distributions of /sup 239/Pu fission rates, the space dependency of the calculated-to-experiment values was found at the active core close to the internal blanket

  15. Decay heat removal analyses on the heavy liquid metal cooled fast breeding reactor. Comparisons of the decay heat removal characteristics on lead, lead-bismuth and sodium cooled reactors

    International Nuclear Information System (INIS)

    Sakai, Takaaki; Ohshima, Hiroyuki; Yamaguchi, Akira

    2000-04-01

    The feasibility study on several concepts for the commercial fast breeder reactor(FBR) in future has been conducted in JNC for the kinds of possible coolants and fuel types to confirm the direction of the FBR developments in Japan. In this report, Lead and Lead-Bismuth eutectic coolants were estimated for the decay heat removal characteristics by the comparison with sodium coolant that has excellent features for the heat transfer and heat transport performance. Heavy liquid metal coolants, such as Lead and Lead-Bismuth, have desirable chemical inertness for water and atmosphere. Therefore, there are many economical plant proposals without an intermediate heat transport system that prevents the direct effect on a reactor core by the chemical reaction between water and the liquid metal coolant at the hypocritical tube failure accidents in a steam generator. In this study, transient analyses on the thermal-hydraulics have been performed for the decay heat removal events in Equivalent plant' with the Lead, Lead-Bismuth and Sodium coolant by using Super-COPD code. And a resulted optimized lead cooled plant in feasibility study was also analyzed for the comparison. In conclusion, it is become clear that the natural circulation performance, that has an important roll in passive safety characteristic of the reactor, is more excellent in heavy liquid metals than sodium coolant during the decay heat removal transients. However, we need to confirm the heat transfer reduction by the oxidized film or the corrosion products expected to appear on the heat transfer surface in the Lead and Lead-Bismuth circumstance. (author)

  16. Decay heat removal analyses in heavy-liquid-metal-cooled fast breeding reactors. Development of the thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakai, Takaaki; Enuma, Yasuhiro [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Iwasaki, Takashi [Nuclear Energy System Inc., Tokyo (Japan); Ohyama, Kazuhiro [Advanced Reactor Technology Co., Ltd., Tokyo (Japan)

    2001-05-01

    The feasibility study on future commercial fast breeder reactors in Japan has been conducted at JNC, in which various plant design options with all the possible coolant and fuel types are investigated to determine the conditions for the future detailed study. Lead-bismuth eutectic coolant has been selected as one of the possible coolant options. During the phase-I activity of the feasibility study in FY1999 and FY2000, several plant concepts, which were cooled by the heavy liquid metal coolant, were examined to evaluate the feasibility mainly with respect to economical competitiveness with other coolant reactors. A medium-scale (300 - 550 MWe) plant, cooled by a lead-bismuth natural circulation flow in a pool type vessel, was selected as the most possible plant concept for the heavy liquid metal coolant. Thus, a conceptual design study for a lead-bismuth-cooled, natural-circulation reactor of 400 MWe has been performed at JNC to identify remaining difficulties in technological aspect and its construction cost evaluation. In this report, a thermal-hydraulic analysis method for lead-bismuth-cooled, natural-circulation reactors is described. A Multi-dimensional Steam Generator analysis code (MSG) was applied to evaluate the natural circulation plant by combination with a flow-network-type, plant dynamics code (Super-COPD). By using this combined multi-dimensional plant dynamics code, decay heat removals, ULOHS and UTOP accidents were evaluated for the 100 MWe STAR-LM concept designed by ANL. In addition, decay heat removal by the Primary Reactor Auxiliary Cooling System (PRACS) in the 400 MWe lead-bismuth-cooled, natural-circulation reactor, being studied at JNC, was analyzed. In conclusion, it becomes clear that the combined multi-dimensional plant dynamics code is suitably applicable to analyses of lead-bismuth-cooled, natural-circulation reactors to evaluate thermal-hydraulic phenomena during steady-state and transient conditions. (author)

  17. Design, calculation and experimental studies for liquid metal system main parameters in support of the liquid lithium fusion reactor

    International Nuclear Information System (INIS)

    Evtikhin, V.A.; Lyublinski, I.E.; Vertkov, A.V.

    2001-01-01

    A new concept of a Liquid Lithium Fusion Reactor and the first experimental results were presented at the 16th IAEA Conference on Fusion Energy. During the past two years theoretical estimations have been made, and calculated and experimental results have been obtained in confirmation of this concept and supporting its progress. The main results of this work are given in the paper. (author)

  18. Design, calculation and experimental studies for liquid metal system main parameters in support of the liquid lithium fusion reactor

    International Nuclear Information System (INIS)

    Evtikhin, V.A.; Lyublinski, I.E.; Vertkov, A.V.

    1999-01-01

    A new concept of a Liquid Lithium Fusion Reactor and the first experimental results were presented at the 16th IAEA Conference on Fusion Energy. During the past two years theoretical estimations have been made, and calculated and experimental results have been obtained in confirmation of this concept and supporting its progress. The main results of this work are given in the paper. (author)

  19. Liquid metal pump

    Science.gov (United States)

    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.

  20. Liquid metal pump

    International Nuclear Information System (INIS)

    Pennell, W.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

  1. Liquid metal pump

    International Nuclear Information System (INIS)

    Pennell, W.E.

    1981-01-01

    A liquid metal pump comprising a shaft support structure which is isolated from the pump housing for better preservation of alignment of shaft bearings. The shaft carries an impeller and the support structure carries an impeller cage which is slidably disposed in a diffuser so as to allow complete removal of pump internals for inspection and repair. The diffuser is concentrically supported in the pump housing which also takes up all reaction forces generated by the discharge of the liquid metal from the diffuser, with floating seals arranged between impeller cage and the diffuser. The space between the diffuser and the pump housing permits the incoming liquid to essentially surround the diffuser. (author)

  2. Evaluation of cobalt and nickel base materials for sliding and static contact applications in a liquid metal fast breeder reactor

    International Nuclear Information System (INIS)

    Hoffman, N.J.; Droher, J.J.; Chang, J.Y.; Galioto, T.A.; Miller, R.L.; Schrock, S.L.; Whitlow, G.A.; Wilson, W.L.; Johnson, R.N.

    1976-01-01

    The paper covers pertinent metallurgical and tribological aspects of three alloys that are being considered for surfaces that must rub while immersed in liquid sodium coolant within a fast breeder reactor system. The alloys are cobalt-base hardfacing alloy type 6, Tribaloy 700, and Inconel 718. Topics discussed include chemistry and microstructure, hardness, and behavior in high-temperature sodium with respect to dynamic friction, diffusion bonding, and corrosion

  3. Brazed thermocouple pass-through for sodium service in a liquid-metal-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Walker, D.E.

    1975-10-01

    Sensors installed in special fuel elements for the EBR-II reactor had 30-ft-long leads that would pass from the sodium environment through a sealed bulkhead. A hydrogen-atmosphere, induction-heated brazing furnace was constructed to simultaneously braze 20-26 separate sensor leads at one time. The brazed seals were leak-tight, and the sheath wall has less than 10 percent interaction with the braze alloy

  4. Comparative analysis of quality assurance systems which effectively control, review and verify the quality of components manufactured for liquid metal cooled fast breeder reactors within the EEC

    International Nuclear Information System (INIS)

    Benn, L.A.

    1985-01-01

    Comparative analyses are made of Quality Assurance Systems, by techniques and the methodology used, for the manufacture of component parts for the Liquid Metal Cooled Fast Breeder Reactor (LMFBR) within the EEC. Two differing alternative systems are presented in the analysis. First, a tabulated analytical treatment which analyses 14 codes and standards relating to Quality Assurance which can be applied to LMFBR's. The comparison equates equivalent clauses between codes and standards followed by an analysis of individual clauses in tabular form, the International Standard ISO 6215. A statistical summary and recommendations conclude this analysis. The second alternative system used in the comparison is a descriptive analytical method applied to 9 selected codes and standards relating to Quality Assurance based on the 13 criteria of the International IAEA Code of Practice no. 50 C.QA entitled ''Quality Assurance for Safety in Nuclear Power Plants''. An investigation is then made of the state of the art on the subject of classification of component parts bearing generally on Quality Assurance. The method of classification is segregated into General, Safety and Inspection categories. A summary of destructive and non destructive controls that may be applied during the manufacture of LMFBR components is given, together with tests that may be applied to selected components, namely Primary Tank, Secondary Sodium Pump and the Primary Cold Trap allocated to Safety Classes, 1, 2 and 3 respectively. The report concludes with a summary of typical records produced at the delivery of a component

  5. Method of charging instruments into liquid metal coolant

    International Nuclear Information System (INIS)

    Yamazaki, Hiroshi

    1980-01-01

    Purpose: To alleviate the thermal shock of a reactor charging machine when charging the machine into liquid metal coolant after the machine is preheated in cover gas. Method: When a reactor fueling machine reaches at the lowermost portion the position immediately above liquid metal coolant surface level, the machine is stopped moving down. The reactor fueling machine is heated at the lowermost portion by thermal radiation from the surface of the liquid metal coolant. After the machine is thus preheated in cover gas, it is again steadily moved down by a winch and charged into the liquid metal coolant. Therefore, the thermal shock of the machine becomes low when charging the machine into the liquid metal coolant to eliminate the damage and deformation at the machine. (Yoshihara, H.)

  6. Properties of structural materials in liquid metal environment. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    Borgstedt, H U [ed.

    1991-12-15

    The International Working Group on Fast Reactors (IWGFR) Specialists Meeting on Properties of Structural Materials in Liquid Metal Environment was held during June 18 to June 20, 1991, at the Nuclear Research Centre (Kernforschungszentrum) in Karlsruhe, Germany. The Specialists Meeting was divided into five technical sessions which addressed topics as follows: Creep-Rupture Behaviour of Structural Materials in Liquid Metal Environment; Behaviour of Materials in Liquid Metal Environments under Off-Normal Conditions;Fatigue and Creep-Fatigue of Structural Materials in Liquid Metal Environment; Crack Propagation in Liquid Sodium; and Conclusions and recommendations. Individual papers have been cataloged separately.

  7. Collection and transmission of milli-volt-level data on a liquid-metal fast-breeder reactor

    International Nuclear Information System (INIS)

    Robinson, C.; Thoburn, R.

    1978-01-01

    The paper describes a high integrity data multiplexing system which caters for up to 252 input elements with remote measuring facilities, using a 12-core flexible cable. Tests carried out on a prototype unit and recommendations for effecting improvements are described. The system was developed following the identification of the problem of having to handle a large number of cables on the top of the Commercial Demonstration Fast Reactor. It has high reliability, a relatively small and conventional cable installation and is applicable to situations outside the nuclear field. (author)

  8. A simple reactivity feedback model accounting for radial core expansion effects in the liquid metal fast reactor

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Yong Bum; Chang, Won Pyo; Haha, Do Hee

    2002-01-01

    The radial core expansion due to the structure temperature rise is one of major negative reactivity insertion mechanisms in metallic fueled reactor. Thermal expansion is a result of both the laws of nature and the particular core design and it causes negative reactivity feedback by the combination of increased core volume captures and increased core surface leakage. The simple radial core expansion reactivity feedback model developed for the SSC-K code was evaluated by the code-to-code comparison analysis. From the comparison results, it can be stated that the radial core expansion reactivity feedback model employed into the SSC-K code may be reasonably accurate in the UTOP analysis

  9. A simple reactivity feedback model accounting for radial core expansion effects in the liquid metal fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min; Lee, Yong Bum; Chang, Won Pyo; Haha, Do Hee [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    The radial core expansion due to the structure temperature rise is one of major negative reactivity insertion mechanisms in metallic fueled reactor. Thermal expansion is a result of both the laws of nature and the particular core design and it causes negative reactivity feedback by the combination of increased core volume captures and increased core surface leakage. The simple radial core expansion reactivity feedback model developed for the SSC-K code was evaluated by the code-to-code comparison analysis. From the comparison results, it can be stated that the radial core expansion reactivity feedback model employed into the SSC-K code may be reasonably accurate in the UTOP analysis.

  10. Detection of gas entrainment into liquid metals

    Energy Technology Data Exchange (ETDEWEB)

    Vogt, T., E-mail: t.vogt@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Fluid Dynamics, 01328 Dresden (Germany); Boden, S. [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Fluid Dynamics, 01328 Dresden (Germany); Andruszkiewicz, A. [Faculty of Mechanical and Power Engineering, Wroclaw University of Technology (Poland); Eckert, K. [Technische Universität Dresden, Institute of Fluid Mechanics, 01062 Dresden (Germany); Eckert, S.; Gerbeth, G. [Helmholtz-Zentrum Dresden-Rossendorf (HZDR), Institute of Fluid Dynamics, 01328 Dresden (Germany)

    2015-12-01

    Highlights: • We present liquid metal experiments dedicated to gas entrainment on the free surface. • Ultrasonic and X-ray attenuation techniques have been used to study the mechanisms of gas entrainment. • A comparison between bubbly flow in water and GaInSn showed substantial differences. • Our results emphasize the importance of liquid metal experiments which are able to provide a suitable data base for numerical code validation. - Abstract: Entrainment of cover gas into the liquid metal coolant is one of the principal safety issues in the design of innovative liquid metal-cooled fast reactors. We present generic experimental studies of this phenomenon in low-melting metals. Ultrasonic and X-ray diagnostic tools were considered for a visualization of gas entrainment at the free surface of the melt. Laboratory experiments were conducted using the eutectic alloy GaInSn, which is liquid at room temperature. Vortex-activated entrainment of air at the free surface of a rotating flow was revealed by ultrasonic techniques. X-ray radioscopy was used to visualize the behavior of argon bubbles inside a slit geometry. The measurements reveal distinct differences between water and GaInSn, especially with respect to the process of bubble formation and the coalescence and breakup of bubbles. Our results emphasize the importance of liquid metal experiments which are able to provide a suitable data base for numerical code validation.

  11. UK status report on detection and localisation of leaks in steam generators of liquid metal fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Smedley, J A [Dounreay Nuclear Power Development Establishment, Caithness, Scotland (United Kingdom)

    1978-10-01

    The development of detection and location techniques applied to defects in sodium heated steam generators in Uk has been associated primarily with the PFR commissioning program. The UK position on leak detection and location studies for LMFBR steam generators may be briefly summarised as follows: a) The Initial concept of' detection using hydrogen concentration measuring equipment in secondary circuit sodium and in steam generator gas spaces has proved sound. The system used on the PFR has been shown to work well under plant conditions in both the under sodium and gas phase modes. b) Experience with small leaks in tube to tube plate welds in the PFR steam generators has indicated the majority of leaks would be detected by the installed equipment. There is, however, a very small possibility an under sodium leak in. a ferritic tube which will quickly and easily block may not be detected before it erupts as a relatively significant defect. Methods are being sought to protect against this unlikely event, although it is recognised there is no reactor hazard but there may be a plant availability problem. c) It is concluded the single barrier concept is still valid, based on the experience in the PFR steam generator units. For future units it is intended to use ferritic boiler tube materials rather than austenitic materials. In general, however, it is concluded the material choice has been satisfactory and manufacturing methods are basically sound. Improved quality assurance methods are being sought with the aim of making future steam generators more reliable than those presently in service. d) It is recognised steam generators are still in a development regime in all countries of the world working in the LMFBR field, but the time is anticipated when a utility can order an LMFBR as a commercial proposition. An attempt has been made to quantify the information necessary to achieve this state.

  12. Liquid metal magnetohydrodynamic convertor

    International Nuclear Information System (INIS)

    Aladiev, I.T.; Dzhamardzhashvili, V.A.

    1981-01-01

    This invention relates to the generation of electrical energy by direct conversion from thermal or electrical energy and notably to liquid metal magnetohydrodynamic convertors. The convertor described in this invention can be successfully used as a source of electrical energy for space vessels, for underwater vessels, for aeronautics and for the generation of electrical energy in thermal or atomic power plants. This liquid metal convertor consists of a heat source, a two phase nozzle, a separator, a steam diffuser and a condenser. These elements are connected together hydraulically in series. The condenser is connected hydraulically to a heat source, a liquid diffuser and a magnetohydrodynamic generator. These elements are interconnected hydraulically to the separator and heat source [fr

  13. Experimental studies and computational benchmark on heavy liquid metal natural circulation in a full height-scale test loop for small modular reactors

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Yong-Hoon, E-mail: chaotics@snu.ac.kr [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of); Cho, Jaehyun [Korea Atomic Energy Research Institute, 111 Daedeok-daero, 989 Beon-gil, Yuseong-gu, Daejeon 34057 (Korea, Republic of); Lee, Jueun; Ju, Heejae; Sohn, Sungjune; Kim, Yeji; Noh, Hyunyub; Hwang, Il Soon [Department of Energy Systems Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 08826 (Korea, Republic of)

    2017-05-15

    Highlights: • Experimental studies on natural circulation for lead-bismuth eutectic were conducted. • Adiabatic wall boundaries conditions were established by compensating heat loss. • Computational benchmark with a system thermal-hydraulics code was performed. • Numerical simulation and experiment showed good agreement in mass flow rate. • An empirical relation was formulated for mass flow rate with experimental data. - Abstract: In order to test the enhanced safety of small lead-cooled fast reactors, lead-bismuth eutectic (LBE) natural circulation characteristics have been studied. We present results of experiments with LBE non-isothermal natural circulation in a full-height scale test loop, HELIOS (heavy eutectic liquid metal loop for integral test of operability and safety of PEACER), and the validation of a system thermal-hydraulics code. The experimental studies on LBE were conducted under steady state as a function of core power conditions from 9.8 kW to 33.6 kW. Local surface heaters on the main loop were activated and finely tuned by trial-and-error approach to make adiabatic wall boundary conditions. A thermal-hydraulic system code MARS-LBE was validated by using the well-defined benchmark data. It was found that the predictions were mostly in good agreement with the experimental data in terms of mass flow rate and temperature difference that were both within 7%, respectively. With experiment results, an empirical relation predicting mass flow rate at a non-isothermal, adiabatic condition in HELIOS was derived.

  14. Liquid metal MHD generator systems

    International Nuclear Information System (INIS)

    Satyamurthy, P.; Dixit, N.S.; Venkataramani, N.; Rohatgi, V.K.

    1985-01-01

    Liquid Metal MHD (LMMHD) Generator Systems are becoming increasingly important in space and terrestrial applications due to their compactness and versatility. This report gives the current status and economic viability of LMMHD generators coupled to solar collectors, fast breeder reactors, low grade heat sources and conventional high grade heat sources. The various thermodynamic cycles in the temperatures range of 100degC-2000degC have been examined. The report also discusses the present understanding of various loss mechanisms inherent in LMMHD systems and the techniques for overcoming these losses. A small mercury-air LMMHD experimental facility being set up in Plasma Physics Division along with proposals for future development of this new technology is also presented in this report. (author)

  15. Steam generation device with heat exchange between a liquid metal coolant and the feedwater

    International Nuclear Information System (INIS)

    Malaval, C.

    1983-01-01

    The invention is particularly applicable to a liquid metal fast breeder reactor plant, the liquid metal being sodium. The steam generation device is described in detail, it allows to get an upper liquid metal level without turbulence and an easier passage for the shock wave towards the steam generator up to the liquid metal level without being laterally reflected back to the intermediate heat exchangers [fr

  16. The SAFR liquid metal concept

    International Nuclear Information System (INIS)

    Baumeister, E.B.

    1987-01-01

    The Sodium Advanced Fast Reactor (SAFR) modular reactor concept is being developed by the team of Rockwell International, Combustion Engineering, and Bechtel under the U.S. Department of Energy's (DOE's) Advanced Liquid Metal Reactor (LMR) program. The SAFR plant would provide a viable alternate to light water reactors, especially for applications favoring small incremental capacity additions. SAFR is also a logical step to facilitate the later transition to LMFBRs. The SAFR plant concept employs multiple 350-MWe LMR Power Pak modules. Each Power Pak is a standardized, shop-fabricated unit that can be barge-shipped to the plant site for installation. The 350-MWe size allows SAFR to capitalize on all the inherent safety features provided by small reactors and factory fabrication, while still preserving some economy of scale. Shop fabrication minimizes nuclear-grade field fabrication and minimizes the overall plant construction schedule and capital cost. Each Power Pak consists of one reactor assembly and associated heat transfer equipment coupled to a single turbine generator. The reactor core employs mixed uranium-plutonium zirconium alloy metal fuel. The metal-alloy fuel (which has been used in EBR-II) has cost, safety, and safeguard advantages. The intrinsic properties of the sodium coolant (e.g., high boiling point, low vapor pressure, and strong natural convection), blended together with the pool-type LMR concept and the metal fuel, result in an inherently safe plant. Passive inherent features provide both public safety and plant investment protection. Refueling is carried out annually on each Power Pak, replacing one-fourth of the core over a 6-day refueling outage. A colocated pyroprocessing fuel cycle facility can be accommodated at the site such that no off-site shipments are required. (J.P.N.)

  17. Technical committee meeting on Liquid Metal Fast Reactor (LMFR) developments. 33rd annual meeting of the International Working Group on Fast Reactors (IWG-FR). Working material

    International Nuclear Information System (INIS)

    2000-01-01

    Over the past 33 years, the IAEA has actively encouraged and advocated international cooperation in fast reactor technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1999/2000, as reported at the 33. annual meeting of the International Working Group on Fast Reactors. It is intended to provide information regarding the current status of LMFR development in IAEA Member States

  18. Archive-cup insert for liquid-metal sampling

    International Nuclear Information System (INIS)

    Nelson, P.A.; Kolba, V.M.; Filewicz, E.C.; Holmes, J.T.

    1975-01-01

    An insert for collecting liquid-metal samples within a vertical casing including an elongated housing with an upper and a lower overflow seal of annular shape is described. The lower seal includes a centrally located pedestal on which a sample cup is disposed. Liquid metal enters the annulus of the upper seal and overflows into the cup which fills and overflows into the lower seal. Liquid-metal overflow from the lower seal is discharged from the insert. On cooling, the liquid metal trapped within the seals solidifies to hermetically isolate the metal sample within the cup. The device is particularly applicable for use with sampling systems on liquid metal-cooled reactors. (U.S.)

  19. Natural convection and boiling heat transfer of a liquid metal in a magnetic field

    International Nuclear Information System (INIS)

    Seki, Masahiro; Kawamura, Hiroshi

    1983-02-01

    A liquid metal is often assumed as a coolant and a breeding material of a Tokamak fusion reactor. However, many problems on the thermo-hydraulics of a liquid metal in a magnetic field are still remained to be studied. In the present report, natural convection and boiling of a liquid metal in a strong magnetic field are studied to examine a fundamental feasibility of a fusion reactor cooled by a liquid metal. In the experimental study of the natural convection, the circulation of a liquid metal was found to be surpressed even by a magnetic field parallel to the gravity. A numerical study has confirmed the conclusion drawn by the experiment. In the study of boiling heat transfer, stable boiling of a liquid metal has been found also in a strong magnetic field. The burnout heat flux hardly affected by the magnetic field. These indicate a fundamental feasibility of the liquid-metal cooling for a Tokamak fusion reactor. (author)

  20. Container for liquid metal

    International Nuclear Information System (INIS)

    Abe, Yoshihito; Imazu, Takayuki; Ueda, Sabuo; Ueya, Katsumi.

    1980-01-01

    Purpose: To arrange a vapor trapping member of a specific structure at the inlet part of a cylindrical gap formed by the inner peripheral surface of the circular opening of a container and the outer peripheral surface of a rotary plug thereby to prevent ingress of vapor in the upper part of the cylindrical gap for a long period of time. Constitution: A sealing material receiving tray is fitted to the container side of the inlet part of a cylindrical gap, and a partition plate is fitted to the rotary plug side. The tray is filled with a sealing material consisting of a large number of steel balls, mesh wire gages and the like, and the partition plate is placed in the tray thereby to carry out sealing of the container. Liquid metal vapor evaporating from the liquid level of the liquid metal adheres to the sealing material to fill the gap, and therefore ingress of vapor to the upper part of the cylindrical gap is prevented, and there is no possibility of causing seal cutting due to the use for a long period. (Sekiya, K.)

  1. Analysis of unscrammed loss of flow and heat sink for PRISM with GEM

    International Nuclear Information System (INIS)

    Slovik, G.C.; Van Tuyle, G.J.; Kennett, R.J.

    1991-01-01

    The US Department of Energy is sponsoring an advanced liquid-metal reactor design by General Electric Company (GE) called PRISM. The intent is to design a reactor with passively safe responses to many operational and severe accidents. PRISM is under review at the US Nuclear Regulatory Commission for licensability with Brookhaven National Laboratory providing technical assistance. Recently, the PRISM design has been modified to include three gas expansion modules (GEMs) on its core periphery. The GEMs were added to quickly reduce the power (by inserting negative reactivity) during loss-of-flow events to curtail peak fuel and clad temperatures predicted in the previous design. The GEM prototypes have been tested at the Fast Flux Test Facility. The significance of the GEMs in PRISM is discussed in this paper through the evaluation of the unprotected loss of flow (ULOF) and loss of heat sink using the SSC code. It has been demonstrated in the past that SSC predicts results similar to GE and other liquid-metal reactor codes

  2. Evaluations of 1990 PRISM design revisions

    International Nuclear Information System (INIS)

    Van Tuyle, G.J.; Slovik, G.C.; Chan, B.C.; Aronson, A.L.; Kennett, R.J.

    1992-03-01

    Analyses of the 1990 version of the PRISM Advanced Liquid Metal Reactor (ALMR) design are presented and discussed. Most of the calculations were performed using BNL computer codes, particularly SSC and MINET. In many cases, independent BNL calculations were compared against analyses presented by General Electric when they submitted the PRISM design revisions for evaluation by the Nuclear Regulatory Commission (NRC). The current PRISM design utilizes the metallic fuel developed by Argonne National Laboratory (ANL) which facilitates the passive/''inherent'' shutdown mechanism that acts to shut down reactor power production whenever the system overheats. There are a few vulnerabilities in the passive shutdown, with the most worrisome being the positive feedback from sodium density decreases or sodium voiding. Various postulated unscrammed events were examined by GE and/or BNL, and much of the analysis discussed in this report is focused on this category of events. For the most part, the BNL evaluations confirm the information submitted by General Electric. The principal areas of concern are related to the performance of the ternary metal fuel, and may be resolved as ANL continues with its fuel development and testing program

  3. PFR liquid metals disposal at Dounreay

    International Nuclear Information System (INIS)

    McIntyre, A.W.

    1997-01-01

    When the Prototype Fast Reactor (PFR) at Dounreay was shut down in 1994, the UKAEA commissioned a series of studies to determine the least cost, lowest risk option for dealing with the liquid metal coolants, i.e. the sodium from the primary and secondary circuits and the NaK from the decay heat removal system. The studies concluded that leaving the liquid metals in situ was not a viable option. Removing the liquid metals had three options, provision of long term external storage facilities, re-use in other projects or treatment for final disposal. The UKAEA invited companies to bid for the challenging task of disposing of more than 1500 t of liquid metals. In 1995 UKAEA awarded NNC Ltd. one of the largest decommissioning projects ever to be let competitively in the UK. During the first year of the contract, the challenges have focused on solving design problems and a number of innovative solutions have been developed by NNC and its subcontractors. From January 1997 the focus has moved to construction on site at Dounreay, and the manufacturing and installation of the mechanical components of the plant

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

    Energy Technology Data Exchange (ETDEWEB)

    Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Groth, Katrina M. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Cardoni, Jeffrey N. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Wheeler, Timothy A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    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.

  5. Status of liquid metal fast reactor development. Proceedings of the 27. meeting of the International Working Group on Fast Reactors held in Vienna, 17-19 May 1994

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-03-01

    These proceedings contain updated and new information on the status of fast reactor development and on activities in the field of advanced nuclear power technology during 1993, as reported at the 27th meeting of the IWGFR held in Vienna, from 17 to 19 May 1994. Refs, figs and tabs.

  6. Status of liquid metal fast reactor development. Proceedings of the 27. meeting of the International Working Group on Fast Reactors held in Vienna, 17-19 May 1994

    International Nuclear Information System (INIS)

    1995-03-01

    These proceedings contain updated and new information on the status of fast reactor development and on activities in the field of advanced nuclear power technology during 1993, as reported at the 27th meeting of the IWGFR held in Vienna, from 17 to 19 May 1994. Refs, figs and tabs

  7. Progress of liquid metal technology and application in energy industries

    International Nuclear Information System (INIS)

    Miyazaki, Keiji; Kamei, Mitsuru; Nei, Hiromichi.

    1990-01-01

    Liquid metals are excellent energy transport media, and recently remarkable development has been observed in the technology of handling sodium and the machinery and equipment. In nuclear fusion, the development of the use of lithium as the coolant is advanced. For space technology, attention has been paid from the early stage to various liquid metals. For general industries, liquid metals have been used for high temperature heat pipes and the utilization of solar heat, and mercury vapor turbines were manufactured for trial. Besides, attention is paid anew to liquid metal MHD electric power generation. The development of the NaS batteries for electric cars and electric power storage and the interchange of liquid metal technology with the fields of iron and steel, metallurgy and so on advance. It is expected that liquid metal technology bears future advanced energy engineering while deepening the interchange with other advanced fields also in order to reactivate atomic energy technology. Liquid metals have the features of high electric and thermal conductivities, chemical activity and opaque property as metals, and fluidity and relatively high boiling point and melting point as liquids. FBRs, fusion reactors and the power sources for space use are described. (K.I.)

  8. Evaluation of postulated LOF [loss-of-flow] events in PRISM and SAFR

    International Nuclear Information System (INIS)

    Chan, B.C.; Van Tuyle, G.J.; Slovik, G.C.; Aronson, A.L.

    1987-01-01

    The PRISM and SAFR designs, as currently proposed by DOE, are designed for ''inherent'', as opposed to ''engineered'', safety. Brookhaven National Laboratory is supporting the initial NRC review of these advanced LMR concepts. A loss-of-flow (LOF) accident coupled with a failure of the reactor shutdown system is one of the major safety concerns in the advanced liquid metal reactor (LMR) evaluation effort. The analysis discussed here covers: (1) primary pipe break without pump trip, (2) primary coolant pump seizure, and (3) primary coolant pump coastdown. The analytical modelling and the calculated thermal and hydraulic behavior are described in detail

  9. Delayed neutrons in liquid metal spallation targets

    International Nuclear Information System (INIS)

    Ridikas, D.; Bokov, P.; David, J.C.; Dore, D.; Giacri, M.L.; Van Lauwe, A.; Plukiene, R.; Plukis, A.; Ignatiev, S.; Pankratov, D.

    2003-01-01

    ) spallation target at PSI (Switzerland), where CEA Saclay is already involved in the development of neutron flux detectors, namely microscopic fission chambers. In addition, in this context other potential applications are: Spallation Neutron Source (USA), European Spallation Source (EU), hybrid reactors for nuclear waste transmutation, EURISOL Radioactive Ion Beam facility (EU), all of them based on liquid metal target technologies. The above scientific project involves 3 countries, namely France, Lithuania and Russia, and is realized in the frame of the ECO-NET program, financed in part by the French Ministry of Foreign Affairs. (authors)

  10. Economic assessment of S-prism including development and generating costs

    Energy Technology Data Exchange (ETDEWEB)

    Boardman, Ch.E. [GE Nuclear Energy San Jose (United States)

    2001-07-01

    S-PRISM is an advanced Fast Reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test of a single Nuclear Steam Supply System (NSSS) for design certification at minimum cost and risk. S-PRISM retains all of the key ALMR (advanced liquid metal reactor) design features including passive reactor shutdown, passive shutdown heat removal, and passive reactor cavity cooling that were developed under an earlier DOE program. Key factors that make S-PRISM competitive include: 1) The use of passive safety systems that eliminate the need for diesel generators and hardened active heat sinks to assure that sufficient heat is removed from the core, reactor, and containment systems following design and beyond design basis events. 2) A seven point advantage in the plant capacity factor (93 versus 86%) over a single large plant. 3) A much shorter construction schedule (45%) made possible by a modular design that allows near parallel (sequenced) construction of three relatively small, simple factory fabricated NSSSs instead of one large complex NSSS. This paper describes the approach, methods, and results of an in-depth economic assessment of S-PRISM. The assessment found that the generation cost from an NOAK plant would be less than 3 cents/kW-hr and that a design certification could be obtained in less than 15 years at a cost of 2.1 billion dollars. (authors)

  11. Economic assessment of S-prism including development and generating costs

    International Nuclear Information System (INIS)

    Boardman, Ch.E.

    2001-01-01

    S-PRISM is an advanced Fast Reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test of a single Nuclear Steam Supply System (NSSS) for design certification at minimum cost and risk. S-PRISM retains all of the key ALMR (advanced liquid metal reactor) design features including passive reactor shutdown, passive shutdown heat removal, and passive reactor cavity cooling that were developed under an earlier DOE program. Key factors that make S-PRISM competitive include: 1) The use of passive safety systems that eliminate the need for diesel generators and hardened active heat sinks to assure that sufficient heat is removed from the core, reactor, and containment systems following design and beyond design basis events. 2) A seven point advantage in the plant capacity factor (93 versus 86%) over a single large plant. 3) A much shorter construction schedule (45%) made possible by a modular design that allows near parallel (sequenced) construction of three relatively small, simple factory fabricated NSSSs instead of one large complex NSSS. This paper describes the approach, methods, and results of an in-depth economic assessment of S-PRISM. The assessment found that the generation cost from an NOAK plant would be less than 3 cents/kW-hr and that a design certification could be obtained in less than 15 years at a cost of 2.1 billion dollars. (authors)

  12. Liquid-metal aspects of HYLIFE

    International Nuclear Information System (INIS)

    Meier, W.R.; Hoffman, N.J.; McDowell, M.W.

    1980-01-01

    The High Yield Lithium Injection Fusion Energy (HYLIFE) converter is a reactor concept for an inertial fusion electric power plant. In this concept, flowing molten lithium protects the structures of the fusion chamber from the deleterious effects of deuterium-tritium (DT) fusion reactions and converts the pulsed fusion energy into steay thermal power. Lithium is circulated as the primary coolant to transfer heat to an intermediate sodium loop which drives a superheated steam cycle. Lithium is also the source of the tritium fuel which is recovered via a molten-salt extraction process. The liquid-metal aspects of the HYLIFE plant with particular emphasis on the lithium systems

  13. Development of oxygen sensors for use in liquid metal

    International Nuclear Information System (INIS)

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

    2015-01-01

    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)

  14. Development of oxygen sensors for use in liquid metal

    Energy Technology Data Exchange (ETDEWEB)

    Van Nieuwenhove, Rudi [Institutt for Energiteknikk, Halden, (Norway); Ejenstam, Jesper; Szakalos, Peter [KTH Royal Institute of Technology, Division of Surface and Corrosion Science, Stockholm, (Sweden)

    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)

  15. Review of the critical heat flux correlations for liquid metals

    International Nuclear Information System (INIS)

    Lee, Yong Bum; Han, H. D.; Chang, W. P.; Kwon, Y. M.

    1999-09-01

    The CHF phenomenon in the two-phase convective flows has been an important issue in the fields of design and safety analysis of light water reactor (LWR) as well as sodium cooled liquid metal reactor (LMR). Especially in the LWR application, many physical aspects of the CHF phenomenon are understood and reliable correlations and mechanistic models to predict the CHF condition have been proposed over the past three decades. Most of the existing CHF correlations have been developed for light water reactor core applications. Compared with water, liquid metals show a divergent picture of boiling pattern. This can be attributed to the consequence that special CHF conditions obtained from investigations with water cannot be applied to liquid metals. Numerous liquid metal boiling heat transfer and two-phase flow studies have put emphasis on development of models and understanding of the mechanism for improving the CHF predictions. Thus far, no overall analytical solution method has been obtained and the reliable prediction method has remained empirical. The principal objectives of the present report are to review the state of the art in connection with liquid metal critical heat flux under low pressure and low flow conditions and to discuss the basic mechanisms. (author)

  16. The modular ALMR (PRISM) fuel cycle

    International Nuclear Information System (INIS)

    Thompson, M.L.

    1990-01-01

    The modular reactor concept, PRISM (power reactor, innovative, small module), originated by General Electric in conjunction with the integral fast reactor (IFR) metal fuel being developed by Argonne National Laboratory (ANL), is the reference US Department of Energy advanced liquid-metal reactor (ALMR). The reference ALMR is unique in several ways; for example, it can produce (or breed) substantially more fissile material than it consumes. It is also unique in that it has the capability to utilize as fuel the long-life radioactive actinides (primarily plutonium, and the minor actinides, neptunium, americium, and curium) present as waste in light water reactor (LWR) spent fuels. This capability provides a means for converting long-life actinide radioactive wastes to elements whose lifetimes and thus storage needs are much shorter, namely, hundreds of years. This could clearly focus and potentially alleviate a controversial aspect (waste disposal) of the nuclear option. While it does not change the need for, or timing of, an initial high-level waste (HLW) repository, the conversion of actinides could change in a dramatic way the time period required for safe storage of nuclear waste and potentially the number and criteria for future repositories. This work considers the potential for utilizing LWR actinides in the ALMR fuel cycle

  17. Pair potentials in liquid metals

    International Nuclear Information System (INIS)

    Faber, T.E.

    1980-01-01

    The argument which justifies the use of a pair potential to describe the structure-dependent term in the energy of liquid metals is briefly reviewed. Because there is an additional term in the energy which depends upon volume rather than structure, and because the pair potential itself is volume-dependent, the relationship between pair potential and observable properties such as pressure, bulk modulus and pair distribution function is more complicated for liquid metals than it is for molecular liquids. Perhaps for this reason, the agreement between pair potentials inferred from observable properties and pair potentials calculated by means of pseudo-potential theory is still far from complete. The pair potential concept is applicable only to simple liquid metals, in which the electron-ion interaction is weak. No attempt is made to discuss liquid transition and rare-earth metals, which are not simple in this sense. (author)

  18. Vapor trap for liquid metal

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, T

    1968-05-22

    In a pipe system which transfers liquid metal, inert gas (cover gas) is packed above the surface of the liquid metal to prevent oxidization of the liquid. If the metal vapor is contained in such cover gas, the circulating system of the cover gas is blocked due to condensation of liquid metal inside the system. The present invention relates to an improvement in vapor trap to remove the metal vapor from the cover gas. The trap consists of a cylindrical outer body, an inlet nozzle which is deeply inserted inside the outer body and has a number of holes to inject the cove gas into the body, metal mesh or steel wool which covers the exterior of the nozzle and on which the condensation of the metal gas takes place, and a heater wire hich is wound around the nozzle to prevent condensation of the metal vapor at the inner peripheral side of the mesh.

  19. Forces in Liquid Metal Contacts

    DEFF Research Database (Denmark)

    Duggen, Lars; Mátéfi-Tempfli, Stefan

    2014-01-01

    Using rather well known theory about capillary bridges between two electrodes we calculate the tensile force that can be applied to liquid metal contacts in the micrometer regime. Assuming circular symmetry, full wetting of the electrodes, and neglecting gravity, we present a brief review of the ...... of the necessary theory and find numerically the forces to be in the 100μN range for liquid metals as mercury and liquid Gallium suspended between electrodes of 20μm radius.......Using rather well known theory about capillary bridges between two electrodes we calculate the tensile force that can be applied to liquid metal contacts in the micrometer regime. Assuming circular symmetry, full wetting of the electrodes, and neglecting gravity, we present a brief review...

  20. Report on the specialists' meeting on passive and active safety features of liquid-metal fast breeder reactors organized by the international atomic energy agency at Oarai Engineering Centre of power reactor and nuclear development corporation, Japan, November 5-7, 1991

    International Nuclear Information System (INIS)

    Paranjpe, S.R.

    1992-01-01

    As recommended by the International Working Group on Fast Reactors (IWGFR), the International Atomic Energy Agency organized a specialists' meeting on passive and active safety features of liquid-metal fast breeder reactors. Specialists from all member countries of IWGFR-China, France, Germany, India, Italy, Japan, Russia, the United Kingdom, and the United States-participated in the meeting and made presentations as listed in Table 1. The Commission of European Communities also sent representatives to the meeting. Table 2 contains a list of participants. The meeting consisted of five sessions: (1) an overview, (2) safety characteristics of decay heal removal systems, (3) safely characteristics of reactor protection systems and reactor shutdown systems, (4) safely characteristics of reactor cores, and (5) general discussions antiformulation of recommendations

  1. Liquid metal cooling concepts in solar power application

    International Nuclear Information System (INIS)

    Deegan, P.B.; Mangus, J.D.; Whitlow, G.A.

    1978-01-01

    The thermodynamic and thermophysical properties and proven technology of a liquid sodium heat transport system provide numerous advantages and benefits for application in a central receiver solar thermal power plant concept. The major advantages of utilizing liquid sodium are: attainment of high thermodynamic cycle efficiencies, reduced relative costs, and achievement of these goals by the mid-1980's through the utilization of proven liquid metal technology developed in the power industry, without the need for extensive development programs. The utilization of liquid sodium reduces the complexity of the design of these systems, thus providing confidence in system reliability. The implementation of the proven technology in liquid metal systems also provides assurance of reliability. In addition, the ease of transition from liquid metal breeder reactor systems to solar application provides immediate availability of this technology

  2. Liquid metal flow control by DC electromagnetic pumps

    International Nuclear Information System (INIS)

    Borges, Eduardo Madeira; Braz Filho, Francisco Antonio; Guimaraes, Lamartine Nogueira Frutuoso

    2006-01-01

    The cooling system of high-density thermal power requires fluids of high thermal conductivity, such as liquid metals. Electromagnetic pumps can be used to liquid metal fluid flow control in cooling circuits. The operation of electromagnetic pumps used to flow control is based on Lorentz force. This force can be achieved by magnetic field and electric current interaction, controlled by external independent power supplies. This work presents the electromagnetic pump operational principles, the IEAv development scheme and the BEMC-1 simulation code. The theoretical results of BEMC-1 simulation are compared to electromagnetic pump operation experimental data, validating the BEMC-1 code. This code is used to evaluate the DC electromagnetic pump performance applied to Mercury flow control and others liquid metal such as Sodium, Lead and Bismuth, used in nuclear fast reactors. (author)

  3. Liquid Metal Engineering and Technology. Volume 1

    International Nuclear Information System (INIS)

    1988-01-01

    These proceedings of the fourth international conference on liquid metal engineering and technology volume 1, are devided into 3 sections bearing on: - Apparatus and components for liquid metal (29 papers) - Liquid metal leaks, fires and fumes (10 papers) - Cleaning, decontamination, waste disposal (14 papers) [fr

  4. Chemistry of liquid metal coolants and sensors

    International Nuclear Information System (INIS)

    Gnanasekaran, T.

    2015-01-01

    Liquid sodium is the coolant of choice for the current generation fast breeder reactors. When sodium contains low levels of dissolved non-metallic impurities, it is highly compatible with structural steels. When the dissolved oxygen level is high, corrosion and mass transfer in sodium-steel circuits are enhanced and this involves formation of NaxMyOz type of species (M = alloying components in steels). Experience has shown that this enhancement of corrosion in a sodium circuit with all austenitic steel structural materials would not be encountered if oxygen level in sodium is below ~ 5ppm. For understanding this observation, a complete knowledge on the phase diagrams of Na-M-O systems and the thermochemical data of all relevant NaxMyOz compounds is essential. This presentation would highlight the work carried out at IGCAR on the chemistry of liquid sodium and heavy liquid metal coolants. Work carried out on various sensors for their use in these liquid metal circuits would be described and their current status would be discussed

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

    International Nuclear Information System (INIS)

    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. Nuclear reactor

    International Nuclear Information System (INIS)

    Garabedian, G.

    1988-01-01

    A liquid reactor is described comprising: (a) a reactor vessel having a core; (b) one or more satellite tanks; (c) pump means in the satellite tank; (d) heat exchanger means in the satellite tank; (e) an upper liquid metal conduit extending between the reactor vessel and the satellite tank; (f) a lower liquid metal duct extending between the reactor vessel and satellite tanks the upper liquid metal conduit and the lower liquid metal duct being arranged to permit free circulation of liquid metal between the reactor vessel core and the satellite tank by convective flow of liquid metal; (g) a separate sealed common containment vessel around the reactor vessel, conduits and satellite tanks; (h) the satellite tank having space for a volume of liquid metal that is sufficient to dampen temperature transients resulting from abnormal operating conditions

  7. Compact, Lightweight Electromagnetic Pump for Liquid Metal

    Science.gov (United States)

    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.

  8. Characterization and management of radioactive sodium and other reactor components as input data for the decommissioning of liquid metal-cooled fast reactors. A compilation of data produced of data produced by members of the IAEA technical working group on fast reactors (TWG-FR) at two consultancies and one technical committee meeting. Working material

    International Nuclear Information System (INIS)

    2002-01-01

    A number of liquid metal cooled fast reactors (LMFRs) are in operation and, some have already been shut down; other reactors will reach the end of their design lifetime in a few years and become candidates for decommissioning. It is unfortunate that little consideration was devoted to decommissioning of reactors at the plant design and construction stage. It is with this focus that the Technical Working Group on Fast Reactors (TWGFR) recommended that the IAEA organize the exchange of information on LMFRs decommissioning technology. It was pointed out that the decommissioning of small sodium-cooled reactors has shown that there are two basic differences between thermal and fast reactors decommissioning: on the one side, the treatment and disposal of radioactive sodium coolant, and on the other side, the management of reactor components, for which the structural materials are activated in depth by fast neutrons. To this end, a Technical Committee Meeting on Sodium Removal and Disposal from LMFRs in Normal Operation and in the framework of Decommissioning (Aix-en-Provence, France, November 1997) and two Consultancies on Decommissioning of the Kazakh BN-350 LMFR (Vienna, Austria, October 1996; Obninsk, Russian Federation, February 1998) were convened by the IAEA. These Meetings brought together a group of experts from France, Russia, Kazakhstan, the UK, and the USA to exchange information on, and to review current technical knowledge and experience in the management of radioactive coolant and reactor components following closing of LMFRs, as well as their design features and operating experience relevant for decommissioning procedures. The report provides general and detailed information on activation characteristics of the primary coolant; treatment and disposal of the spent sodium; removal of the residual sodium deposits and decontamination; the activation characteristics of the reactor components and the management of the latter. The recurring theme is finding

  9. Liquid Metal Cooled Reactor Core Design and Structural Material Research by the Institute for Energy and Transport at the European Commission's Joint Research Centre

    International Nuclear Information System (INIS)

    Nilsson, K.-F.; D’Agata, E.; Fütterer, M.A.; Lázaro, A.; Grah, A.; Tsige-Tamirat, H.; Ammirabile, L.; Flores y Flores, A.; Hähner, P.

    2013-01-01

    JRC supports, in accordance with the SET Plan, nuclear energy as one of the options in a future sustainable energy mix for Europe. JRC supports the policy instruments for coordination and implementation of research, financing, demonstration and deployment of next generation nuclear energy systems, incl. ESNII prototypes and demonstrators. JRC performs independent and contributes to collaborative R&D on important safety issues for next generation reactors

  10. Liquid-metal dip seal with pneumatic spring

    International Nuclear Information System (INIS)

    Poindexter, A.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

  11. Experience on the removal of impurities from liquid metal systems by cold-trapping

    Energy Technology Data Exchange (ETDEWEB)

    Bray, J. A.

    1963-10-15

    Experience in impurity removal by cold-trapping, which was obtained on DFR and its associated liquid metal rigs, is reviewed. The development of the present DFR cold-trapping system is outlined, and the operation of the additional pumped loops, which were required in order to control the reactor impurity levels, are described. Operation of the liquid metal rigs ancillary to the reactor project is discussed with particular reference to the control of impurity levels. (auth)

  12. Nuclear fast neutron reactor cooled by a liquid metal and of which internal structures are equipped with a thermal protection device

    International Nuclear Information System (INIS)

    Lemercier, G.; Lions, N.

    1986-01-01

    The internal structures of a nuclear fast neutron reactor are covered at least partially, on the most hot side, by a thermal protection device. This device comprises modular plates arranged end to end with a certain play between themselves and taking approximately the shape of the internal structures. Each plate is fixed in its center on the internal structures by a stud. A small plate fixed at one of the corners of each plate and covering partially the adjacent plates ensures the safety fixing of these ones [fr

  13. Analyses of deformation and thermal-hydraulics within a wire-wrapped fuel subassembly in a liquid metal fast reactor by the coupled code system

    Energy Technology Data Exchange (ETDEWEB)

    Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp; Ohshima, Hiroyuki; Ito, Masahiro

    2017-06-15

    Highlights: • The coupled computational code system allowed for mechanical and thermal-hydraulic analyses in a fast reactor fuel subassembly. • In this system interactive calculations between flow area deformations and coolant temperature changes are repeated to their convergence state. • Effects on bundle-duct interaction on coolant temperature distributions were investigated by using the code system. - Abstract: The coupled numerical analysis of mechanical and thermal-hydraulic behaviors was performed for a wire-wrapped fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal-hydraulic analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that the radial distribution of coolant temperature in the subassembly tended to flatten as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such flattening of temperature distribution was slightly observed as a result of fuel pin bowings due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal-hydraulics was also investigated in this study.

  14. Evaluation for the effects of a ring plate device to eliminate free surface gradients in liquid metal fast breeder reactor vessel using multi-dimensional thermohydraulics computer code

    International Nuclear Information System (INIS)

    Gao Ming Qing.

    1997-02-01

    There is a free surface at the upper plenum in a reactor vessel of LMFBR. The free surface has spatial gradient caused by the internal coolant flow. This is a disadvantageous factor to engineering from the view point of gas entrainment into coolant. To eliminate the free surface gradients, ring plates about 20 cm wide are fitted at about 1 meter under the free surface. They interfere fluid flow, and decrease the component velocity in vertical direction. To investigate the efficiency of the ring plates, analyses with the AQUA-VOF code were carried out. For contrast, three conditions were given: Case-1: Without ring plates. Case-2: Ring plates, fitted at 1.125 m under the free surface. Case-3: Ring plates, fitted at 1.5 m under the free surface. The results shown that the ring plates have a sufficiently high potential to eliminate the free surface gradients due to disperse the momentum along reactor vessel axis to radial direction. In the calculations with ring plate (Cases-2 and -3), the maximum free surface height differences and the maximum gradients of free surface were decreased to less than 15% and 64% compared with the case without ring plates, respectively. (author)

  15. Liquid metals replace water steam

    Energy Technology Data Exchange (ETDEWEB)

    Kozlov, V

    1976-12-01

    The techniques are described of power generation with regard to their effectiveness which depends on the efficiency of the conversion of thermal energy into electric energy. The magnetohydrodynamic conversion of energy is based on the use of induced electromotive force which results from the movement of the conductor in the magnetic field. The use of liquid metal as the working medium makes it possible to increase the initial temperature of the magnetohydrodynamic cycle to the limit of the highest technically attainable temperatures. The total efficiency of energy conversion in magnetohydrodynamic converters is 2 to 6%.

  16. Liquid metals replace water steam

    International Nuclear Information System (INIS)

    Kozlov, V.

    1976-01-01

    The techniques are described of power generation with regard to their effectiveness which depends on the efficiency of the conversion of thermal energy into electric energy. The magnetohydrodynamic conversion of energy is based on the use of induced electromotive force which results from the movement of the conductor in the magnetic field. The use of liquid metal as the working medium makes it possible to increase the initial temperature of the magnetohydrodynamic cycle to the limit of the highest technically attainable temperatures. The total efficiency of energy conversion in magnetohydrodynamic converters is 2 to 6%. (J.B.)

  17. Bearing for liquid metal pump

    International Nuclear Information System (INIS)

    Dickinson, R.J.; Pennell, W.E.; Wasko, J.

    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

  18. Development of a Two-dimensional Thermohydraulic Hot Pool Model and ITS Effects on Reactivity Feedback during a UTOP in Liquid Metal Reactors

    International Nuclear Information System (INIS)

    Lee, Yong Bum; Jeong, Hae Yong; Cho, Chung Ho; Kwon, Young Min; Ha, Kwi Seok; Chang, Won Pyo; Suk, Soo Dong; Hahn, Do Hee

    2009-01-01

    The existence of a large sodium pool in the KALIMER, a pool-type LMR developed by the Korea Atomic Energy Research Institute, plays an important role in reactor safety and operability because it determines the grace time for operators to cope with an abnormal event and to terminate a transient before reactor enters into an accident condition. A two-dimensional hot pool model has been developed and implemented in the SSC-K code, and has been successfully applied for the assessment of safety issues in the conceptual design of KALIMER and for the analysis of anticipated system transients. The other important models of the SSC-K code include a three-dimensional core thermal-hydraulic model, a reactivity model, a passive decay heat removal system model, and an intermediate heat transport system and steam generation system model. The capability of the developed two-dimensional hot pool model was evaluated with a comparison of the temperature distribution calculated with the CFX code. The predicted hot pool coolant temperature distributions obtained with the two-dimensional hot pool model agreed well with those predicted with the CFX code. Variations in the temperature distribution of the hot pool affect the reactivity feedback due to an expansion of the control rod drive line (CRDL) immersed in the pool. The existing CRDL reactivity model of the SSC-K code has been modified based on the detailed hot pool temperature distribution obtained with the two-dimensional pool model. An analysis of an unprotected transient over power with the modified reactivity model showed an improved negative reactivity feedback effect

  19. The CRDL model of SSC-K code for the safety improvement of a pool-type liquid metal-cooled reactor

    International Nuclear Information System (INIS)

    Jung, H. Y.; Ha, K. S.; Jang, W. P.; Hu, S.; Lee, Y. B.

    2004-01-01

    With the increased thermal power of KALIMER-600, it becomes important to model accurately the reactivity feedback effects due to the thermal expansion of a fuel rod and internal structure during a transient. In KALIMER design, the fuel axial expansion, core radial expansion, and the control rod drive line/reactor vessel (CRDL/RV) thermal expansion are the important reactivity feedback mechanisms. It is required to develop a more detailed CRDL/RV model for the accurate analysis of the KALIMER-600 transient because the control rod drive line of 9.5 m is immersed in the hot pool. For this a new CRDL/RV model was developed to model the effect of expansion of CRDL utilizing the temperature distribution obtained with the hot-pool 2-D model of SSC-K code. It is estimated that the developed model describes more realistically the negative reactivity insertion effect due to the initial temperature change during the UTOP transient of KALIMER-150

  20. The fifth research coordination meeting (RCM) on 'Updated codes and methods to reduce the calculational uncertainties of liquid metal fast reactors reactivity effects'. Working material

    International Nuclear Information System (INIS)

    2004-01-01

    The general objective of the CRP is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. The objectives of the fifth RCM were: to review the progress achieved since the 4th RCM; to review and finalize the draft synthesis report on BN-600 MOX Fueled Core Benchmark Analysis (Phase 4); to compare the results of Phase 5 (BFS Benchmark Analysis); to agree on the work scope of Phase 6 (BN-Full MOX Minor Actinide Core Benchmark); to discuss the preparation of the final report. In this context, review and related discussions were made on the following items: summary review of Actions and results since the 4th RCM; finalization of the draft synthesis report on BN-600 full MOX-fueled core benchmark analysis (Phase 4); presentation of individual results for Phase 5 by Member States; preliminary inter-comparison analysis of the results for Phase 5; definition of the benchmark model and work scope to be performed for Phase 6; details of the work scope and future CRP timetable for preparing a final report

  1. Liquid metals fire control engineering handbook

    International Nuclear Information System (INIS)

    Ballif, J.L.

    1979-02-01

    This handbook reviews the basic requirements of the use of liquid metals with emphasis on sodium which has the greatest current usage. It delineates the concepts necessary to design facilities both radioactive and nonradioactive for use with liquid metals. It further reviews the state-of-the-art in fire extinguishers and leak detection equipment and comments on their application and sensitivity. It also provides details on some engineering features of value to the designer of liquid metal facilities

  2. The concepts of liquid metal of IV generation

    International Nuclear Information System (INIS)

    Carbonnier, J. L.

    2005-01-01

    The concepts of liquid metals, due to their large spectrum, show important possibility of sustainable development: two concepts of liquid metal (Sodium and Lead) were engaged in the frame of the IV generation. The reactors with sodium benefit from considerable background of experience and of important work on projects to aim at the price diminution and the increase of safety (EFR, JSFR). The commitment of Japan as a leader of this concept and the support by France allow to contemplate an industrial deployment from 2015. The lead reactors offer some advantages in the domain of safety but otherwise require a highly important research and development binded to the control of the corrosion, the perspective of deployment of this concept are more hypothetical

  3. Liquid metal Flow Meter - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. Technique for detecting liquid metal leaks

    International Nuclear Information System (INIS)

    Bauerle, J.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

  5. The fourth research co-ordination meeting (RCM) on 'Updated codes and methods to reduce the calculational uncertainties of liquid metal fast reactors reactivity effects'. Working material

    International Nuclear Information System (INIS)

    2003-01-01

    The fourth Research Co-ordination Meeting (RCM) of the Co-ordinated Research Project (CRP) on 'Updated Codes and Methods to Reduce the Calculational Uncertainties of the LMFR Reactivity Effect' was held during 19-23 May, 2003 in Obninsk, Russian Federation. The general objective of the CRP is to validate, verify and improve methodologies and computer codes used for the calculation of reactivity coefficients in fast reactors aiming at enhancing the utilization of plutonium and minor actinides. The first RCM took place in Vienna on 24 - 26 November 1999. The meeting was attended by 19 participants from 7 Member States and one from an international organization (France, Germany, India, Japan, Rep. of Korea, Russian Federation, the United Kingdom, and IAEA). The participants from two Member States (China and the U.S.A.) provided their results and presentation materials even though being absent at the meeting. The results for several relevant reactivity parameters obtained by the participants with their own state-of-the-art basic data and codes, were compared in terms of calculational uncertainty, and their effects on the ULOF transient behavior of the hybrid BN- 600 core were evaluated. Contributions of the participants in the benchmark analyses is shown. This report first addresses the benchmark definitions and specifications given for each Phase and briefly introduces the basic data, computer codes, and methodologies applied to the benchmark analyses by various participants. Then, the results obtained by the participants in terms of calculational uncertainty and their effect on the core transient behavior are intercompared. Finally it addresses some conclusions drawn in the benchmarks

  6. Generation and characterization of gas bubbles in liquid metals

    International Nuclear Information System (INIS)

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

    1996-01-01

    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

  7. Rapid-action valve especially for liquid metal

    International Nuclear Information System (INIS)

    Velek, V.; Cejka, J.; Jakl, J.

    1976-01-01

    A rapid-action large-diameter valve and control fitting was designed for cooling circuits of nuclear power plants with fast liquid metal cooled reactors, namely sodium cooled. The design meets the requirements for axial symmetry about the flow axis and secures the equilibrium of hydraulic and hydrodynamic forces observing the uniformity of the thickness of the fitting walls which is a pre-requisite for good resistance to thermal shock. (F.M.)

  8. Numerical simulation on single bubble rising behavior in liquid metal using moving particle semi-implicit method

    International Nuclear Information System (INIS)

    Zuo Juanli; Tian Wenxi; Qiu Suizheng; Chen Ronghua; Su Guanghui

    2011-01-01

    The gas-lift pump in liquid metal cooling fast reactor (LMFR) is an innovational conceptual design to enhance the natural circulation ability of reactor core. The two-phase flow character of gas-liquid metal makes significant improvement of the natural circulation capacity and reactor safety. In present basic study, the rising behavior of a single nitrogen bubble in five kinds of liquid metals (lead bismuth alloy, liquid kalium, sodium, potassium sodium alloy and lithium lead alloy) was numerically simulated using moving particle semi-implicit (MPS) method. The whole growing process of single nitrogen bubble in liquid metal was captured. The bubble shape and rising speed of single nitrogen bubble in each liquid metal were compared. The comparison between simulation results using MPS method and Grace graphical correlation shows a good agreement. (authors)

  9. Specificity in liquid metal induced embrittlement

    CSIR Research Space (South Africa)

    Fernandes, PJL

    1996-12-01

    Full Text Available One of the most intriguing features of liquid metal induced embrittlement (LMIE) is the observation that some liquid metal-solid metal couples are susceptible to embrittlement, while others appear to be immune. This is referred to as the specificity...

  10. Liquid metals: fundamentals and applications in chemistry.

    Science.gov (United States)

    Daeneke, T; Khoshmanesh, K; Mahmood, N; de Castro, I A; Esrafilzadeh, D; Barrow, S J; Dickey, M D; Kalantar-Zadeh, K

    2018-04-03

    Post-transition elements, together with zinc-group metals and their alloys belong to an emerging class of materials with fascinating characteristics originating from their simultaneous metallic and liquid natures. These metals and alloys are characterised by having low melting points (i.e. between room temperature and 300 °C), making their liquid state accessible to practical applications in various fields of physical chemistry and synthesis. These materials can offer extraordinary capabilities in the synthesis of new materials, catalysis and can also enable novel applications including microfluidics, flexible electronics and drug delivery. However, surprisingly liquid metals have been somewhat neglected by the wider research community. In this review, we provide a comprehensive overview of the fundamentals underlying liquid metal research, including liquid metal synthesis, surface functionalisation and liquid metal enabled chemistry. Furthermore, we discuss phenomena that warrant further investigations in relevant fields and outline how liquid metals can contribute to exciting future applications.

  11. Experiments with Liquid Metal Walls: Status of the Lithium Tokamak Experiment

    OpenAIRE

    Boyle, Dennis; Gray, Timothy; Granstedt, Erik; Kozub, Thomas; Berzak, Laura; Hammett, Gregory; Kugel, Henry; Leblanc, Benoit; Logan, Nicholas; Jacobson, Craig M.; Lucia, Matthew; Jones, Andrew; Lundberg, Daniel; Timberlake, John; Majeski, Richard

    2010-01-01

    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 wa...

  12. The PFR challenge: how to dispose of 1500t of liquid metal

    International Nuclear Information System (INIS)

    Anon.

    1996-01-01

    The decommissioning of liquid metal cooled FBRs presents many interesting challenges requiring innovative engineering solutions. In advance of the Prototype Fast Reactor (PFR) at Dounreay in Scotland being shut down in 1994 the owners, UKAEA, commissioned a series of risk and cost assessment studies to determine the least risk, lowest cost long term strategy. This concluded that a priority was safe disposal of the large inventories of liquid metal coolant. UK companies were invited to take up the challenge. (author)

  13. Liquid Metal Fast Breeder Reactors: a bibliography

    International Nuclear Information System (INIS)

    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

  14. Liquid Metal Fast Breeder Reactors: a bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Raleigh, H.D. (ed.)

    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.

  15. LMR [liquid metal reactor] centrifugal pump coastdowns

    International Nuclear Information System (INIS)

    Dunn, F.E.; Malloy, D.J.

    1987-01-01

    A centrifugal pump model which describes the interrelationships of the pump discharge flowrate, pump speed, shaft torque and dynamic head has been implemented based upon existing models. Specifically, the pump model is based upon the dimensionless-homologous pump theory of Wylie and Streeter. Given data from a representative pump, homologous theory allows one to predict the transient characteristics of similarly sized pumps. This homologous pump model has been implemented into both the one-dimensional SASSYS-1 systems analysis code and the three-dimensional COMMIX-1A code. Comparisons have been made both against other pump models (CRBR) and actual pump coastdown data (EBR-II and FFTF). Agreement with this homologous pump model has been excellent. Additionally, these comparisons indicate the validity of applying the medium size pump data of Wylie and Streeter to a range of typical LMR centrifugal pumps

  16. Liquid Metal Fast Breeder Reactors: a bibliography

    Energy Technology Data Exchange (ETDEWEB)

    Raleigh, H.D. (ed.)

    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 Fast Breeder Reactors: a bibliography

    International Nuclear Information System (INIS)

    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

  18. Turbulent magnetohydrodynamics in liquid metals

    International Nuclear Information System (INIS)

    Berhanu, Michael

    2008-01-01

    In electrically conducting fluids, the electromagnetic field is coupled with the fluid motion by induction effects. We studied different magnetohydrodynamic phenomena, using two experiments involving turbulent flows of liquid metal. The first mid-sized uses gallium. The second, using sodium, is conducted within the VKS (Von Karman Sodium) collaboration. It has led to the observation of the dynamo effect, namely converting a part of the kinetic energy of the fluid into magnetic energy. We have shown that, depending on forcing conditions, a statistically stationary dynamo, or dynamical regimes of magnetic field can be generated. In particular, polarity reversals similar to those of Earth's magnetic field were observed. Meanwhile, experiment with Gallium has been developed to study the effects of electromagnetic induction by turbulent flows in a more homogeneous and isotropic configuration than in the VKS experiment. Using data from these two experiments, we studied the advection of magnetic field by a turbulent flow and the induced fluctuations. The development of probes measuring electrical potential difference allowed us to further highlight the magnetic braking of a turbulent flow of Gallium by Lorentz force. This mechanism is involved in the saturation of the dynamo instability. (author) [fr

  19. Surface studies of liquid metals and alloys

    International Nuclear Information System (INIS)

    Bastasz, Robert

    2003-01-01

    Liquid metals and alloys have been proposed for use in nuclear fusion reactors to serve as replaceable plasma-facing surfaces that remove particles and heat from reacting plasmas. Several materials are being considered for this purpose including lithium, gallium, and tin as well as some of the alloys made from these elements. In order to better understand the properties of liquid surfaces, the technique of low-energy ion scattering was used to examine the surface composition of several of these materials in vacuum as a function of temperature. Oxygen is found to rapidly segregate to the surface of several metallic liquids. The segregation process can be interpreted using a simple thermodynamic model based on Gibbs theory. In the case of an alloy of Sn and Li, Li also segregates to the liquid surface. This provides a means to produce a surface enriched in Li, which is more plasma compatible than Sn, without the need to handle large quantities of liquid Li. (author)

  20. Liquid metal coolant flow rate regulation

    International Nuclear Information System (INIS)

    Vitkovskij, I.V.; Glukhikh, V.A.; Kirillov, I.R.; Smirnov, A.M.

    1981-01-01

    Some aspects of fast reactor and experimental bench operation related to liquid metal flow rate regulation are considered. Requirements to the devices for the flow rate regulation are formulated. A new type of these devices namely magnetohydrodynamic (MHD) throttles is described. Structural peculiarities of MHD throttles of different types are described as well. It is noted that the MHD throttles with a screw channel have the best energy mass indices. On the basis of the comparison of the MHD throttles with mechanical valves it is concluded that the MHD throttles described are useful for regulating the flow rates of any working media. Smoothness and accuracy of the flow rate regulation by the throttles are determined by the electric control circuit and may be practically anyone. The total coefficient of hydraulic losses in the throttle channel in the absence of a magnetic field is ten and more times lesser than in completely open mechanical valve. Electromagnetic time constant of the MHD throttles does not exceed several tenths of a second [ru

  1. MHD power conversion employing liquid metals

    International Nuclear Information System (INIS)

    Houben, J.W.M.A.; Massee, P.

    1969-02-01

    The work performed in the field of MHD generation of electricity by means of liquid metals is described. It is shown that the study of two-phase flows is essential in this topic of research; two-phase flows are therefore described. Two types of generators which can be utilized with liquid metals have been studied. The results of this study are described. A short survey of the prospects of other liquid metal systems which emerge from a study of the literature is given. Finally, conclusions are drawn concerning possibilities for further investigation

  2. Problems of hydrogen - water vapor - inert gas mixture use in heavy liquid metal coolant technology

    International Nuclear Information System (INIS)

    Ul'yanov, V.V.; Martynov, P.N.; Gulevskij, V.A.; Teplyakov, Yu.A.; Fomin, A.S.

    2014-01-01

    The reasons of slag deposit formation in circulation circuits with heavy liquid metal coolants, which can cause reactor core blockage, are considered. To prevent formation of deposits hydrogen purification of coolant and surfaces of circulation circuit is used. It consists in introduction of gaseous mixtures hydrogen - water vapor - rare gas (argon or helium) directly into coolant flow. The principle scheme of hydrogen purification and the processes occurring during it are under consideration. Measures which make it completely impossible to overlap of the flow cross section of reactor core, steam generators, pumps and other equipment by lead oxides in reactor facilities with heavy liquid metal coolants are listed [ru

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

    Energy Technology Data Exchange (ETDEWEB)

    Mirhoseini, S. M. H.; Volpe, F. A., E-mail: fvolpe@columbia.edu [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

    2016-11-15

    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.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

    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. Liquid metal cooled divertor for ARIES

    International Nuclear Information System (INIS)

    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 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

  6. Mechanisms of liquid-metal embrittlement

    International Nuclear Information System (INIS)

    Popovich, V.V.

    1979-01-01

    The mechanism of the embrittlement of metals and alloys during deformation in contact with liquid metals are discussed. With 20Kh13 steel in a Pb-Sn melt and polycrystalline Al in the presence of various mercury solutions a.s examples, considered are the three main processes - adsorption, corrosion (dissolution), formation of new phases which cause the disintegration of materials under the action of liquid-metallic media. Presented are data on plastic ductile and strength properties of the above materials in the presence of liquid-metallic media. A model is described that takes into account the effect of the medium upon the plastic deformation and the part the medium plays in liquid-metallic embrittlement

  7. Equipment for liquid metal pressure measurement

    International Nuclear Information System (INIS)

    Jung, J.

    1977-01-01

    Equipment is proposed for measuring liquid metal pressure in piping or a tank. An auxiliary piping is connected to the piping or tank at the measuring point. The auxiliary piping transports liquid metal to a container by means of an electromagnetic pump. The piping also houses an electromagnetic flow ratemeter connected to an electric comparator. The comparator and the electromagnetic pump are connected to the pump output generator. (Z.M.)

  8. Overview of liquid-metal MHD

    International Nuclear Information System (INIS)

    Dunn, P.F.

    1978-01-01

    The basic features of the two-phase liquid-metal MHD energy conversion under development at Argonne National Laboratory are presented. The results of system studies on the Rankine-cycle and the open-cycle coal-fired cycle options are discussed. The liquid-metal MHD experimental facilities are described in addition to the system's major components, the generator, mixer and nozzle-separator-diffuser

  9. Tokamak with liquid metal toroidal field coil

    International Nuclear Information System (INIS)

    Ohkawa, T.; Schaffer, M.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

  10. Liquid metal cooling of synchrotron optics

    International Nuclear Information System (INIS)

    Smither, R.K.

    1993-01-01

    The installation of insertion devices at existing synchrotron facilities around the world has stimulated the development of new ways to cool the optical elements in the associated x-ray beamlines. Argonne has been a leader in the development of liquid metal cooling for high heat load x-ray optics for the next generation of synchrotron facilities. The high thermal conductivity, high volume specific heat, low kinematic viscosity, and large working temperature range make liquid metals a very efficient heat transfer fluid. A wide range of liquid metals were considered in the initial phase of this work. The most promising liquid metal cooling fluid identified to date is liquid gallium, which appears to have all the desired properties and the fewest number of undesired features of the liquid metals examined. Besides the special features of liquid metals that make them good heat transfer fluids, the very low vapor pressure over a large working temperature range make liquid gallium an ideal cooling fluid for use in a high vacuum environment. A leak of the liquid gallium into the high vacuum and even into very high vacuum areas will not result in any detectable vapor pressure and may even improve the vacuum environment as the liquid gallium combines with any water vapor or oxygen present in the system. The practical use of a liquid metal for cooling silicon crystals and other high heat load applications depends on having a convenient and efficient delivery system. The requirements for a typical cooling system for a silicon crystal used in a monochromator are pumping speeds of 2 to 5 gpm (120 cc per sec to 600 cc per sec) at pressures up to 100 psi. No liquid metal pump with these capabilities was available commercially when this project was started, so it was necessary to develop a suitable pump in house

  11. Intercomparison of liquid metal fast reactor seismic analysis codes. V. 3: Comparison of observed effects with computer simulated effects on reactor cores from seismic disturbances. Proceedings of a final research co-ordination meeting

    International Nuclear Information System (INIS)

    1996-05-01

    This publication contains the final papers summarizing the validation of the codes on the basis of comparison of observed effects with computer simulated effects on reactor cores from seismic disturbances. Refs, figs tabs

  12. Liquid metal technology in fusion

    International Nuclear Information System (INIS)

    Torre Cabezas, M. de la; Martin Espigares, M.; Lapena, J.

    1985-01-01

    Lithium (or Li-Pb) is one of the several possible coolants being considered for the blanket of magnetic toroidal fusion reactor, not only because of its good thermal and neutron properties, but also because the tritium required to fuel the reactor can be produced by neutron reactions in the lithium. In this paper two main technology tasks to be proposed in our fusion programme have been identified: 1) the development of impurity monitoring devices for use in lithium and Li-Pb environments; 2) effects of Li and Li-Pb environments on the low cycle fatigue properties of different steels. (author)

  13. Improvement to liquid metal pumps

    International Nuclear Information System (INIS)

    Pennell, W.E.

    1981-01-01

    This invention concerns the coolant pumps of nuclear reactors. It resolves the problems of structures which have to withstand high temperatures, the difficulties in keeping the multiple bearings of the shaft aligned, the excessive fluid flows, the risks of scoring and seizing-up by self welding, the need for narrow machining tolerances and the difficulties of access for inspection and repairs [fr

  14. Liquid metal flow measurement by neutron radiography

    International Nuclear Information System (INIS)

    Takenaka, N.; Ono, A.; Matsubayashi, M.; Tsuruno, A.

    1996-01-01

    Visualization of a liquid metal flow and image processing methods to measure the vector field are carried out by real-time neutron radiography. The JRR-3M real-time thermal neutron radiography facility in the Japan Atomic Energy Research Institute was used. Lead-bismuth eutectic was used as a working fluid. Particles made from a gold-cadmium intermetallic compound (AuCd 3 ) were used as the tracer for the visualization. The flow vector field was obtained by image processing methods. It was shown that the liquid metal flow vector field was obtainable by real-time neutron radiography when the attenuation of neutron rays due to the liquid metal was less than l/e and the particle size of the tracer was larger than one image element size digitized for the image processing. (orig.)

  15. Intercomparison of liquid metal fast reactor seismic analysis codes. V.1: Validation of seismic analysis codes using reactor core experiments. Proceedings of a research co-ordination meeting held in Vienna, 16-17 November 1993

    International Nuclear Information System (INIS)

    1995-05-01

    The Research Co-ordination Meeting held in Vienna, 16-17 November 1993, was attended by participants from France, India, Italy, Japan and the Russian Federation. The meeting was held to discuss and compare the results obtained by various organizations for the analysis of Italian tests on PEC mock-up. The background paper by A. Martelli, et al., Italy, entitled Fluid-Structure Interaction Experiments of PEC Core Mock-ups and Numerical Analysis Performed by ENEA presented details on the Italian PEC (Prova Elementi di Combustibile, i.e. Fuel Element Test Facility) test data for the benchmark. Several papers were presented on the analytical investigations of the PEC reactor core experiments. The paper by M. Morishita, Japan, entitled Seismic Response Analysis of PEC Reactor Core Mock-up, gives a brief review of the Japanese data on the Monju mock-up core experiment which had been distributed to the participating countries through the IAEA. Refs, figs and tabs

  16. Liquid metal blanket module testing and design for ITER/TIBER II

    International Nuclear Information System (INIS)

    Mattas, R.F.; Cha, Y.; Finn, P.A.; Majumdar, S.; Picologlou, B.; Stevens, H.; Turner, L.

    1988-05-01

    A major goal for ITER is the testing of nuclear components to demonstrate the integrated performance of the most attractive concepts that can lead to a commercial fusion reactor. As part of the ITER/TIBER II study, the test program and design of test models were examined for a number of blanket concepts. The work at Argonne National Laboratory focused on self-cooled liquid metal blankets. A test program for liquid metal blankets was developed based upon the ITER/TIBER II operating schedule and the specific data needs to resolve the key issues for liquid metals. Testing can begin early in reactor operation with liquid metal MHD tests to confirm predictive capability. Combined heat transfer/MHD tests can be performed during initial plasma operation. After acceptable heat transfer performance is verified, tests to determine the integrated high temperature performance in a neutron environment can begin. During the high availability phase operation, long term performance and reliability tests will be performed. It is envisioned that a companion test program will be conducted outside ITER to determine behavior under severe accident conditions and upper performance limits. A detailed design of a liquid metal test module and auxiliary equipment was also developed. The module followed the design of the TPSS blanket. Detailed analysis of the heat transfer and tritium systems were performed, and the overall layout of the systems was determined. In general, the blanket module appears to be capable of addressing most of the testing needs. 8 refs., 27 figs., 11 tabs

  17. Liquid metal-to-gas leak-detection instruments

    International Nuclear Information System (INIS)

    Matlin, E.; Witherspoon, J.E.; Johnson, J.L.

    1982-01-01

    It is desirable for liquid-metal-cooled reactors that small liquid metal-to-gas leaks be reliably detected. Testing has been performed on a number of detection systems to evaluate their sensitivity, response time, and performance characteristics. This testing has been scheduled in three phases. The first phase was aimed at screening out the least suitable detectors and optimizing the performance of the most promising. In the second phase, candidates were tested in a 1500 ft 3 walk-in type enclosure in which leaks were simulated on 24-in. and 3-in. piping. In the third phase of testing, selected type detectors were tested in the 1500-ft 3 enclosure with Clinch River Breeder Reactor Plant (CRBRP) pipe insulation configurations and detector tubing configuration with cell gas recirculation simulated. Endurance testing of detection equipment was also performed as part of this effort. Test results have been shown that aerosol-type detectors will reliably detect leaks as small as a few grams per hour when sampling pipe insulation annuli

  18. Applicability of PRISM PRA Methodology to the Level II Probabilistic Safety Analysis of KALIMER-600 (I) (Core Damage Event Tree Analysis Part)

    International Nuclear Information System (INIS)

    Park, S. Y.; Kim, T. W.; Ha, K. S.; Lee, B. Y.

    2009-03-01

    The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the PSA domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of KALIMER-600 reactor. An applicability of the PSA of the PRISM plant to the KALIMER-600 has been studied. The study is confined to a core damage event tree analysis which is a part of a level 2 PSA. Assuming that the accident types, which can be developed from level 1 PSA, are same as the PRISM PRA, core damage categories are defined and core damage event trees are developed for the KALIMER-600 reactor. Fission product release fractions of the core damage categories and branch probabilities of the core damage event trees are referred from the PRISM PRA temporarily. Plant specific data will be used during the detail analysis

  19. Stretchable and Soft Electronics using Liquid Metals.

    Science.gov (United States)

    Dickey, Michael D

    2017-07-01

    The use of liquid metals based on gallium for soft and stretchable electronics is discussed. This emerging class of electronics is motivated, in part, by the new opportunities that arise from devices that have mechanical properties similar to those encountered in the human experience, such as skin, tissue, textiles, and clothing. These types of electronics (e.g., wearable or implantable electronics, sensors for soft robotics, e-skin) must operate during deformation. Liquid metals are compelling materials for these applications because, in principle, they are infinitely deformable while retaining metallic conductivity. Liquid metals have been used for stretchable wires and interconnects, reconfigurable antennas, soft sensors, self-healing circuits, and conformal electrodes. In contrast to Hg, liquid metals based on gallium have low toxicity and essentially no vapor pressure and are therefore considered safe to handle. Whereas most liquids bead up to minimize surface energy, the presence of a surface oxide on these metals makes it possible to pattern them into useful shapes using a variety of techniques, including fluidic injection and 3D printing. In addition to forming excellent conductors, these metals can be used actively to form memory devices, sensors, and diodes that are completely built from soft materials. The properties of these materials, their applications within soft and stretchable electronics, and future opportunities and challenges are considered. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  20. Liquid metals. Coexistence line, critical parameters, compressibility

    International Nuclear Information System (INIS)

    Filippov, L.P.

    1986-01-01

    Formulae to calculate four characteristic parameters of liquid metals (density, compressibility, critical temperature and individual parameter) according to four initial data are obtained: two values of vapor density and two values of vapor pressure. Comparison between experimental and calculation results are presented for liquid Cs, Na, Li, K, Rb

  1. Clad buffer rod sensors for liquid metals

    International Nuclear Information System (INIS)

    Jen, C.-K.; Ihara, I.

    1999-01-01

    Clad buffer rods, consisting of a core and a cladding, have been developed for ultrasonic monitoring of liquid metal processing. The cores of these rods are made of low ultrasonic-loss materials and the claddings are fabricated by thermal spray techniques. The clad geometry ensures proper ultrasonic guidance. The lengths of these rods ranges from tens of centimeters to 1m. On-line ultrasonic level measurements in liquid metals such as magnesium at 700 deg C and aluminum at 960 deg C are presented to demonstrate their operation at high temperature and their high ultrasonic performance. A spherical concave lens is machined at the rod end for improving the spatial resolution. High quality ultrasonic images have been obtained in the liquid zinc at 600 deg C. High spatial resolution is needed for the detection of inclusions in liquid metals during processing. We also show that the elastic properties such as density, longitudinal and shear wave velocities of liquid metals can be measured using a transducer which generates and receives both longitudinal and shear waves and is mounted at the end of a clad buffer rod. (author)

  2. 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

    International Nuclear Information System (INIS)

    Moisseytsev, A.; Sienicki, J.J.

    2012-01-01

    Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO 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 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 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 2 heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO 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 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 calculations reveal that the

  3. Heaters to simulate fuel pins for heat transfer tests in single-phase liquid-metal-flow

    International Nuclear Information System (INIS)

    Casal, V.; Graf, E.; Hartmann, W.

    1976-09-01

    The development of heaters for thermal simulation of the fuel elements of liquid metal cooled fast breeder reactors (SNR) is reported. Beginning with the experimental demands various heating methods are discussed for thermodynamic investigations of the heat transfer in liquid metals. Then a preferred heater rod is derived to simulate the fuel pins of a SNR. Finally it is reported on the fabrication and the operation practice. (orig.) [de

  4. High-power spallation target using a heavy liquid metal free surface flow

    International Nuclear Information System (INIS)

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

    2015-01-01

    A prototype of a heavy liquid metal free surface target as proposed for the multi-purpose hybrid research reactor for high-tech applications in Mol, Belgium, has been set up and experimentally investigated at the Karlsruhe Liquid Metal Laboratory. A stable operation was demonstrated in a wide range of operating conditions and the surface shape was detected and compared with numerical pre-calculations employing Star-CD. Results show a very good agreement of experiment and numerical predictions which is an essential input for other windowless target designs like the META:LIC target for the European Spallation Source. (author)

  5. Analysis of unscrammed events in PRISM

    International Nuclear Information System (INIS)

    Slovik, G.C.; Van Tuyle, G.J.

    1990-01-01

    The PRISM reactor is presently under pre-application licensing review by the NRC, with Brookhaven National Laboratory (BNL) providing technical assistance. The purpose of this paper is to review the current PRISM design and describe the results from the SSC Code calculations performed at BNL, for a series of unscrammed accidents. 3 refs., 5 figs

  6. Compact device to heat up a liquid metal

    International Nuclear Information System (INIS)

    Blanc, R.; Pelloux, L.

    1981-01-01

    Device for heating a liquid metal, sodium for instance, this device being in one piece and capable of being introduced in one go into the tank containing the liquid metal and comprising heating rods and an electromagnetic pump [fr

  7. Fundamental study on cavitation erosion in liquid metal. Effect of liquid parameter on cavitation erosion in liquid metals (Joint research)

    International Nuclear Information System (INIS)

    Hattori, Shuji; Kurachi, Hiroaki; Inoue, Fumitaka; Watashi, Katsumi; Tsukimori, Kazuyuki; Yada, Hiroki; Hashimoto, Takashi

    2009-02-01

    Cavitation erosion, which possibly occurs on the surfaces of fluid machineries and components contacting flowing liquid and causes sponge-like damage on the material surface, is important problem, since it may become the cause of performance deduction, life shortening, noise, vibration of mechanical components and moreover failure of machine. Research on cavitation erosion in liquid metal is very important to confirm the safety of fast breeder reactor using sodium coolant and to avoid serious damage of the target vessel of spallation neutron source containing liquid-mercury. But the research on cavitation erosion in liquid metal has been hardly performed because of its specially in comparison with that in water. In this study, a cavitation erosion test apparatus was developed to carry out the erosion tests in low-temperature liquid metals. Cavitation erosion tests were carried out in liquid lead-bismuth alloy and in deionized water. We discuss the effect of liquid parameters and temperature effects on the erosion rate. We reach to the following conclusions. The erosion rate was evaluated in terms of a relative temperature which was defind as the percentage between freezing and boiling points. At 14degC relative temperature, the erosion rate is 10 times in lead-bismuth alloy, and 2 to 5 times in sodium, compared with that in deionized water. At 14degC relative temperature, the erosion rate can be evaluated in terms of the following parameter. 1 / (1/ρ L /C L +1/ρ S C S )√ρ L . Where ρ is the material density and c is the velocity of sound, L and S denote liquid and solid. In the relative temperature between 14 and 30degC, the temperature dependence on the erosion rate is due to the increase in vapor pressure. (author)

  8. Measurement of Liquid-Metal Two-Phase Flow with a Dynamic Neutron Radiography

    International Nuclear Information System (INIS)

    Cha, J. E.; Lim, I. C.; Kim, H. R.; Kim, C. M.; Nam, H. Y.; Saito, Y.

    2005-01-01

    The dynamic neutron radiography(DNR) has complementary characteristics to X-ray radiography and is suitable to visualization and measurement of a multi-phase flow research in a metallic duct and liquid metal flow. The flow-field information of liquid metal system is very important for the safety analysis of fast breeder reactor and the design of the spallation target of accelerator driven system. A DNR technique was applied to visualize the flow field in the gas-liquid metal two-phase flow with the HANARO-beam facility. The lead bismuth eutectic and the nitrogen gas were used to construct the two-phase flow field in the natural circulation U-channel. The two-phase flow images in the riser were taken at various combinations of the liquid flow and gas flow with high frame-rate neutron radiography at 1000 fps

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

    Science.gov (United States)

    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.

  10. Properties of structural materials in liquid metal environment

    International Nuclear Information System (INIS)

    Borgstedt, H.U.

    1991-12-01

    The proceedings contain 16 contributions to the following topics: 1. Creep-Rupture Behaviour of Structural Materials in Liquid Metal Environment; 2. Behaviour of Materials in Liquid Metal Environment under Off-Normal Conditions; 3. Fatigue and Creep-Fatigue of Structural Materials in Liquid Metal Environment; and 4. Crack Propagation in Liquid Sodium. (MM)

  11. The Dounreay PFR Liquid-Metal Disposal Project

    International Nuclear Information System (INIS)

    Sherwood, D.V.; Comline, A.; Small, J.; Blyth, J.

    2005-01-01

    The UKAEA Prototype Fast Reactor at Dounreay had a liquid sodium-cooled core. Following its shutdown in 1994, the liquid metal is being removed from the reactor and other vessels by means of specialized equipment and reacted with an aqueous solution of sodium hydroxide in a special vessel. The reaction products are neutralized with hydrochloric acid to produce a saline solution.The reactor sodium delivery and processing equipment is all of novel design. As sodium has been withdrawn from the vessel, it has been necessary to switch off the primary sodium pumps (used to heat the sodium), and the reactor is now kept at temperature by a purpose-designed electric heater and a NaK loop heater.A primary sodium extract pump has currently removed [approximately]450 tonnes of primary sodium from the reactor. As the level falls special equipment will be used to punch a hole in the primary circuit pipe work and to drill the strongback to allow trapped sodium to drain for extraction

  12. Oxidation-Mediated Fingering in Liquid Metals

    Science.gov (United States)

    Eaker, Collin B.; Hight, David C.; O'Regan, John D.; Dickey, Michael D.; Daniels, Karen E.

    2017-10-01

    We identify and characterize a new class of fingering instabilities in liquid metals; these instabilities are unexpected due to the large interfacial tension of metals. Electrochemical oxidation lowers the effective interfacial tension of a gallium-based liquid metal alloy to values approaching zero, thereby inducing drastic shape changes, including the formation of fractals. The measured fractal dimension (D =1.3 ±0.05 ) places the instability in a different universality class than other fingering instabilities. By characterizing changes in morphology and dynamics as a function of droplet volume and applied electric potential, we identify the three main forces involved in this process: interfacial tension, gravity, and oxidative stress. Importantly, we find that electrochemical oxidation can generate compressive interfacial forces that oppose the tensile forces at a liquid interface. The surface oxide layer ultimately provides a physical and electrochemical barrier that halts the instabilities at larger positive potentials. Controlling the competition between interfacial tension and oxidative (compressive) stresses at the interface is important for the development of reconfigurable electronic, electromagnetic, and optical devices that take advantage of the metallic properties of liquid metals.

  13. Some Issues in Liquid Metals Research

    Directory of Open Access Journals (Sweden)

    Maria José Caturla

    2015-11-01

    Full Text Available The ten articles [1–10] included in this Special Issue on “Liquid Metals” do not intend to comprehensively cover this extensive field, but, rather, to highlight recent discoveries that have greatly broadened the scope of technological applications of these materials. Improvements in understanding the physics of liquid metals are, to a large extent, due to the powerful theoretical tools in the hands of scientists, either semi-empirical [1,5,6] or ab initio (molecular dynamics, see [7]. Surface tension and wetting at metal/ceramic interfaces is an everlasting field of fundamental research with important technological implications. The review of [2] is broad enough, as the work carried out at Grenoble covers almost all interesting matters in the field. Some issues of interest in geophysics and astrophysics are discussed in [3]. The recently discovered liquid–liquid transition in several metals is dealt with in [4]. The fifth contribution [5] discusses the role of icosahedral superclusters in crystallization. In [6], thermodynamic calculations are carried out to identify the regions of the ternary phase diagram of Al-Cu-Y, where the formation of amorphous alloys is most probable. Experimental data and ab initio calculations are presented in [7] to show that an optimal microstructure is obtained if Mg is added to the Al-Si melt before than the modifier AlP alloy. Shock-induced melting of metals by means of laser driven compression is discussed in [8]. With respect to recent discoveries, one of the most outstanding developments is that of gallium alloys that are liquid at room temperature [9], and that, due to the oxide layer that readily cover their surface, maintain some “stiffness”. This has opened the possibility of 3D printing with liquid metals. The last article in this Special Issue [10] describes nano-liquid metals, a suspension of liquid metal and its alloy containing nanometer-sized particles. A room-temperature nano-liquid metal

  14. Self-cooled liquid-metal blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Arheidt, K.; Barleon, L.

    1988-01-01

    A blanket concept for the Next European Torus (NET) where 83Pb-17Li serves both as breeder material and as coolant is described. The concept is based on the use of novel flow channel inserts for a decisive reduction of the magnetohydrodynamic (MHD) pressure drop and employs beryllium as neutron multiplier in order to avoid the need for breeding blankets at the inboard side of the torus. This study includes the design, neutronics, thermal hydraulics, stresses, MHDs, corrosion, tritium recovery, and safety of a self-cooled liquid-metal blanket. The results of the investigations indicate that the self-cooled blanket is an attractive alternative to other driver blanket concepts for NET and that it can be extrapolated to the conditions of a DEMO reactor

  15. Liquid metal mist cooling and MHD Ericsson cycle for fusion energy conversion

    International Nuclear Information System (INIS)

    Greenspan, E.

    1989-01-01

    The combination of liquid metal mist coolant and a liquid metal MHD (LMMHD) energy conversion system (ECS) based on the Ericsson cycle is being proposed for high temperature fusion reactors. It is shown that the two technologies are highly matchable, both thermodynamically and physically. Thermodynamically, the author enables delivering the fusion energy to the cycle with probably the highest practical average temperature commensurate with a given maximum reactor design constraint. Physically, the mist cooling and LMMHD ECSs can be coupled directly, thus eliminating the need for primary heat exchangers and reheaters. The net result is expected to be a high efficiency, simple and reliable heat transport and ECS. It is concluded that the proposed match could increase the economic viability of fusion reactors, so that a thorough study of the two complementary technologies is recommended. 11 refs., 3 figs

  16. Evaluation of liquid metal embrittlement of SS304 by Cd and Cd-Al solutions

    International Nuclear Information System (INIS)

    Thomas, J.K.; Iyer, N.C.; Begley, J.A.

    1992-01-01

    The susceptibility of stainless steel 304 to liquid metal embrittlement (LME) by cadmium (Cd) and cadmium-aluminum (Cd-Al) solutions was examined as part of a failure evaluation for SS304-clad cadmium reactor safety rods which had been exposed to elevated temperatures. The active, or cadmium (Cd) bearing, portion of the safety rod consists of a 0.756 in. diameter aluminum allow (Al-6061) core, a 0.05 in. thick Cd layer, and a 0.042 in. thick Type 304 stainless steel cladding. The safety rod thermal tests were conducted as part of a program to define the response of reactor core components to a hypothetical LOCA for the Savannah River Site (SRS) production reactor. LME was considered as a potential failure mechanism based on the nature of the failure and susceptibility of austenitic stainless steels to embrittlement by other liquid metals

  17. Overview of fast reactor safety research and development in the USA

    International Nuclear Information System (INIS)

    Neuhold, R.J.; Avery, R.; Marchaterre, J.F.

    1986-01-01

    The liquid metal reactor (LMR) safety R and D program in the U.S. is presently focused on support of two modular innovative reactor concepts: PRISM - the General Electric Power Reactor Inherently Safe Module and SAFR - the Rockwell International Sodium Advanced Fast Reactor. These reactor plant concepts accommodate the use of either oxide fuel or the metal fuel which is under development in the Argonne National Laboratory (ANL) Integral Fast Reactor (IFR) program. Both concepts emphasize prevention of accidents through enhancement of inherent and passive safety characteristics. Enhancement of these characteristics is expected to be a major factor in establishing new and improved safety criteria and licensing arrangements with regulatory authorities for advanced reactors. Limited work is also continuing on the Large Scale Prototype Breeder (LSPB), a large pool plant design. Major elements of the current and restructured safety program are discussed. (author)

  18. Liquid metal reactor/Pressurized water reactor plant comparison study

    International Nuclear Information System (INIS)

    Halverson, T.G.

    1986-01-01

    The selection between alternative electric power generating technologies is mainly based on their overall economics. Capital costs account for over 60% of the total busbar cost of nuclear plants. Estimates reported in the literature have shown capital cost ratios of LMRs to PWRs ranging from less than 1 to as high as 1.8. To reduce this range of uncertainty, the study selected a method for cataloging plant hardware and then performed comparisons using engineering judgment as to the anticipated and reasonable cost differences. The paper summarizes the resulting one-on-one comparisons of components, systems, and buildings and identifies the LMR-PWR similarities and differences which influence costs. The study leads to the conclusion that the capital cost of the most up-to-date large LMR design would be very close to that of the latest PWRs

  19. Design analysis of liquid metal pipe supports

    International Nuclear Information System (INIS)

    Margolin, L.L.; LaSalle, F.R.

    1979-02-01

    Design guidelines pertinent to liquid metal pipe supports are presented. The numerous complex conditions affecting the support stiffness and strength are addressed in detail. Topics covered include modeling of supports for natural frequency and stiffness calculations, support hardware components, formulas for deflection due to torsion, plate bending, and out-of-plane flexibility. A sample analysis and a discussion on stress analysis of supports are included. Also presented are recommendations for design improvements for increasing the stiffness of pipe supports and which were utilized in the FFTF system

  20. Liquid metal degassing in electromagnetic mixing

    Energy Technology Data Exchange (ETDEWEB)

    Pakhomov, A I; EHL' -FAVAKHRI, KAMAL' -ABD-RABU MOKHAMED [LENINGRADSKIJ POLITEKHNICHESKIJ INST. (USSR)

    1977-01-01

    Experimental results for laboratory and industrial conditions are presented showing the favourable effect of electromagnetic mixing on hot metal degassing process. It has been found that the intensity and duration of the mixing process increase with the degree of iron and steel degassing. Initiation of cavitation phenomena during hot metal electromagnetic mixing is intensified because of the presence of alien inclusions in the metal reducing the tensile strength of the liquid metal. This is the most substantial factor contributing to the gas content in the process of electromagnetic mixing.

  1. Sorption of radiostrontium from liquid metallic rubidium

    International Nuclear Information System (INIS)

    Zhujkov, B.L.; Kokhnyuk, V.M.; Vincent, J.S.

    2008-01-01

    One studied the radiostrontium (in particular, 82 Sr) sorption from the liquid metallic Rb in various inorganic sorbents (both metals and oxides). One studied the temperature dependence and the dynamics of the adsorption and put forward an interpretation of the observed mechanisms. The elaborated approach enables to extract 82 Sr efficiently from the Rb targets upon their irradiation by the accelerated protons. One studies various procedures and the applicability of the process to ensure elaboration of a new highly efficient process to produce 82 Sr, in particular, with the targets of the circulating Rb [ru

  2. Development of thermohydraulic codes for modeling liquid metal boiling in LMR fuel subassemblies

    International Nuclear Information System (INIS)

    Sorokin, G.A.; Avdeev, E.F.; Zhukov, A.V.; Bogoslovskaya, G.P.; Sorokin, A.P.

    2000-01-01

    An investigation into the reactor core accident cooling, which are associated with the power grow up or switch off circulation pumps in the event of the protective equipment comes into action, results in the problem of liquid metal boiling heat transfer. Considerable study has been given over the last 30 years to alkaline metal boiling including researches of heat transfer, boiling patterns, hydraulic resistance, crisis of heat transfer, initial heating up, boiling onset and instability of boiling. The results of these investigations have shown that the process of liquid metal boiling has substantial features in comparison with water boiling. Mathematical modeling of two phase flows in fast reactor fuel subassemblies have been developed intensively. Significant success has been achieved in formulation of two phase flow through the pin bundle and in their numerical realization. Currently a set of codes for thermohydraulic analysis of two phase flows in fast reactor subassembly have been developed with 3D macrotransfer governing equations. These codes are used for analysis of boiling onset and liquid metals boiling in fuel subassemblies during loss-of-coolant accidents, of warming up of reactor core, of blockage of some part of flow cross section in fuel subassembly. (author)

  3. Design analyses of self-cooled liquid metal blankets

    International Nuclear Information System (INIS)

    Gohar, Y.

    1986-12-01

    A trade-off study of liquid metal self-cooled blankets was carried out to define the performance of these blankets and to determine the potential to operate at the maximum possible values of the performance parameters. The main parameters considered during the course of the study were the tritium breeding ratio (TBR), the blanket energy multiplication factor, the energy fraction lost to the shield, the lithium-6 enrichment in the breeder material, the total blanket thickness, the reflector material selection, and the compositions of the different blanket zones. Also, a study was carried out to assess the impact of different reactor design choices on the reactor performance parameters. The design choices include the impurity control system (limiter or divertor), the material choice for the limiter, the elimination of tritium breeding from the inboard section of tokamak reactors, and the coolant choice for the nonbreeding inboard blanket. In addition, tritium breeding benchmark calculations were performed using different transport codes and nuclear data libraries. The importance of the TBR in the blanket design motivated the benchmark calculations

  4. An inherently safe power reactor module

    International Nuclear Information System (INIS)

    Salerno, L.N.

    1985-01-01

    General Electric's long participation in liquid metal reactor technology has led to a Power Reactor Inherently Safe Module (PRISM) concept supported by DOE contract DE-AC06-85NE37937. The reactor module is sized to maximize inherent safety features. The small size allows factory fabrication, reducing field construction and field QA/QC labor, and allows safety to be demonstrated in full scale, to support a pre-licensed standard commercial product. The module is small enough to be placed underground, and can be combined with steam and electrical generating equipment to provide a complete electrical power producing plant in the range of 400-1200 MWe. Initial assessments are that the concept has the potential to be economically competitive with existing methods of power production used by the utility industry

  5. Experimental testing facilities for ultrasonic measurements in heavy liquid metal

    International Nuclear Information System (INIS)

    Cojocaru, V.; Ionescu, V.; Nicolescu, D.; Nitu, A.

    2016-01-01

    The thermo-physical properties of Heavy Liquid Metals (HLM), like lead or its alloy, Lead Bismuth Eutectic (LBE), makes them attractive as coolant candidates in advanced nuclear systems. The opaqueness, that is common to all liquid metals, disables all optical methods. For this reason ultrasound waves are used in different applications in heavy liquid metal technology, for example for flow and velocity measurements and for inspection techniques. The practical use of ultrasound in heavy liquid metals still needs to be demonstrated by experiments. This goal requires heavy liquid metal technology facility especially adapted to this task. In this paper is presented an experimental testing facility for investigations of Heavy Liquid Metals acoustic properties, designed and constructed in RATEN ICN. (authors)

  6. Oscillatory Convection in Rotating Liquid Metals

    Science.gov (United States)

    Bertin, Vincent; Grannan, Alex; Aurnou, Jonathan

    2016-11-01

    We have performed laboratory experiments in a aspect ratio Γ = 2 cylinder using liquid gallium (Pr = 0 . 023) as the working fluid. The Ekman number varies from E = 4 ×10-5 to 4 ×10-6 and the Rayleigh number varies from Ra = 3 ×105 to 2 ×107 . Using heat transfer and temperature measurements within the fluid, we characterize the different styles of low Pr rotating convective flow. The convection threshold is first overcome in the form of a container scale inertial oscillatory mode. At stronger forcing, wall-localized modes develop, coexisting with the inertial oscillatory modes in the bulk. When the strength of the buoyancy increases further, the bulk flow becomes turbulent while the wall modes remain. Our results imply that rotating convective flows in liquid metals do not develop in the form of quasi-steady columns, as in Pr = 1 planetary and stellar dynamo models, but in the form of oscillatory motions. Therefore, convection driven dynamo action in low Pr fluids can differ substantively than that occurring in typical Pr = 1 numerical models. Our results also suggest that low wavenumber, wall modes may be dynamically and observationally important in liquid metal dynamo systems. We thank the NSF Geophysics Program for support of this project.

  7. Turbulent convection in liquid metal with and without rotation

    OpenAIRE

    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 bu...

  8. CFD analysis of liquid metal cooled rod assembly

    International Nuclear Information System (INIS)

    Son, H.M.; Suh, K.Y.

    2007-01-01

    The model subassembly of the BREST-type reactor core is a pin bundle of square arrangement. In this bundle there are two zones which differ with respect to pin diameters and level of heat production. The model pin bundle contains one spacer grid which is located near the midplane of the rod bundle geometry. The coolant consists of a eutectic alloy of 22% sodium (Na) plus 78% potassium (K). Experiments were performed in order to observe the thermal hydraulic behavior of the liquid metal coolant in the BREST core simulator. Results were obtained for the coolant exit temperatures, central measuring pin simulator external surface temperatures, and coolant velocities at the perimeter of the measuring pin simulator. A computational fluid dynamics (CFD) code is used to simulate the liquid metal flows in subchannels. Semi-fine mesh structures were used to model the flow with reasonable accuracy and speed once rigorous node resolution dependency had been tested. A subchannel analysis code was used to investigate the flows as well. Since the subchannel analysis code is based on a lumped parameter model, it only calculates the subchannel averaged velocity values. The CFD code results were averaged on the subchannel basis to be comparable with the results from the subchannel code. The mixing vane is not considered for the time being so as to simplify the problem and to reduce the computational cost. The two codes showed similar results. The difference between the experimental and computational results is considered to mainly originate from the existence of the mixing vane. (authors)

  9. CFD analysis of liquid metal cooled rod assembly

    Energy Technology Data Exchange (ETDEWEB)

    Son, H.M.; Suh, K.Y. [Seoul National Univ. (Korea, Republic of)

    2007-07-01

    The model subassembly of the BREST-type reactor core is a pin bundle of square arrangement. In this bundle there are two zones which differ with respect to pin diameters and level of heat production. The model pin bundle contains one spacer grid which is located near the midplane of the rod bundle geometry. The coolant consists of a eutectic alloy of 22% sodium (Na) plus 78% potassium (K). Experiments were performed in order to observe the thermal hydraulic behavior of the liquid metal coolant in the BREST core simulator. Results were obtained for the coolant exit temperatures, central measuring pin simulator external surface temperatures, and coolant velocities at the perimeter of the measuring pin simulator. A computational fluid dynamics (CFD) code is used to simulate the liquid metal flows in subchannels. Semi-fine mesh structures were used to model the flow with reasonable accuracy and speed once rigorous node resolution dependency had been tested. A subchannel analysis code was used to investigate the flows as well. Since the subchannel analysis code is based on a lumped parameter model, it only calculates the subchannel averaged velocity values. The CFD code results were averaged on the subchannel basis to be comparable with the results from the subchannel code. The mixing vane is not considered for the time being so as to simplify the problem and to reduce the computational cost. The two codes showed similar results. The difference between the experimental and computational results is considered to mainly originate from the existence of the mixing vane. (authors)

  10. Numerical simulation of turbulent forced convection in liquid metals

    International Nuclear Information System (INIS)

    Vodret, S; Di Maio, D Vitale; Caruso, G

    2014-01-01

    In the frame of the future generation of nuclear reactors, liquid metals are foreseen to be used as a primary coolant. Liquid metals are characterized by a very low Prandtl number due to their very high heat diffusivity. As such, they do not meet the so-called Reynolds analogy which assumes a complete similarity between the momentum and the thermal boundary layers via the use of the turbulent Prandtl number. Particularly, in the case of industrial fluid-dynamic calculations where a resolved computation near walls could be extremely time consuming and could need very large computational resources, the use of the classical wall function approach could lead to an inaccurate description of the temperature profile close to the wall. The first aim of the present study is to investigate the ability of a well- established commercial code (ANSYS FLUENT v.14) to deal with this issue, validating a suitable expression for the turbulent Prandtl number. Moreover, a thermal wall-function developed at Universite Catholique de Louvain has been implemented in FLUENT and validated, overcoming the limits of the solver to define it directly. Both the resolved and unresolved approaches have been carried out for a channel flow case and assessed against available direct numerical and large eddy simulations. A comparison between the numerically evaluated Nusselt number and the main correlations available in the literature has been also carried out. Finally, an application of the proposed methodology to a typical sub-channel case has been performed, comparing the results with literature correlations for tube banks

  11. Numerical modelling of inert gas bubble rising in liquid metal pool

    International Nuclear Information System (INIS)

    Pradeep, Arjun; Sharma, Anil Kumar; Ponraju, D.; Nashine, B K.

    2016-01-01

    Two-phase flow finds several applications in safe operation of Sodium-cooled Fast Reactor (SFR). Numerical modelling of bubble rise dynamics in liquid metal pool of SFR is essential for the evaluation of residence time and shape changes, which are of utmost importance for simulating associated heat and mass transfer processes involved in reactor safety. A numerical model has been developed based on OpenFOAM for the evaluation of two-dimensional inert gas bubble rise dynamics in stagnant liquid metal pool. The governing model equations are discretized and solved using the Volume of Fluid based solver available in OpenFOAM with appropriate initial and boundary conditions. The model has been validated with available numerical benchmark results for laminar transient two-phase flow. The model has been used to evaluate velocity and rise trajectory of argon gas bubble with different diameters through a pool of liquid sodium. (author)

  12. Interfacial transport phenomena and stability in liquid-metal/water systems: scaling considerations

    International Nuclear Information System (INIS)

    Abdulla, S.; Liu, X.; Anderson, M.; Bonazza, R.; Corradini, M.; Cho, D.

    2001-01-01

    One concept being considered for steam generation in innovative nuclear reactor applications, involves water coming into direct contact with a circulating molten metal. The vigorous agitation of the two fluids, the direct liquid-liquid contact and the consequent large interfacial area give rise to very high heat transfer coefficients and rapid steam generation. For an optimum design of such direct contact heat exchange and vaporization systems, detailed knowledge is necessary of the various flow regimes, interfacial transport phenomena, heat transfer and operational stability. In this paper we describe current results from the first year of this research that studies the transport phenomena involved with the injection of water into molten metals (e.g., lead alloys). In particular, this work discusses scaling considerations related to direct contact heat exchange, our experimental plans for investigation and a test plan for the important experimental parameters; i.e., the water and liquid metal mass flow rates, the liquid metal pool temperature and the ambient pressure of the direct contact heat exchanger. Past experimental work and initial scaling results suggest that our experiments can directly represent the proper liquid metal pool temperature and the water subcooling. The experimental variation in water and liquid metal flow rates and system pressure (1-10 bar), although smaller than the current conceptual system designs, is sufficient to verify the expected scale effects to demonstrate the phenomena. (authors)

  13. Nitrogen injection in stagnant liquid metal. Eulerian-Eulerian and VOF calculations by fluent

    International Nuclear Information System (INIS)

    Pena, A.; Esteban, G.A.

    2004-01-01

    High power spallation sources are devices that can be very useful in different fields, as medicine, material science, and also in the Accelerator Driven Systems (ADS). This devices use Heavy Liquid Metals (HLM) as the spallation target. Furthermore, HLM are thought to be the coolant of those big energy sources produced by the process. Fast breeder reactors, advanced nuclear reactors, as well as the future designs of fusion reactors, also consider HLM as targets or coolants. Gas injection in liquid metal flows allows the enhancement of this coolant circulation. The difference in densities between the gas and the liquid metal is a big challenge for the multiphase models implemented in the Computational Fluid Dynamics (CFD) codes. Also the changing shape of the bubbles involves extra difficulties in the calculations. A N 2 flow in stagnant Lead-Bismuth eutectic (Pb-Bi), experiment available at Forschungszentrum Rossendorf e.V (FZR) in Germany, was used in one of the work-packages of the ASCHLIM project (EU contract number FIKW-CT-2001-80121). In this paper, calculations made by the UPV/EHU (University of the Basque Country) show measuring data compared with numerical results using the CFD (Computational Fluid Dynamics) code FLUENT and two multiphase models: the Eulerian-Eulerian and the Volume of Fluid (VOF). The interpretation of the experimental resulting velocities was difficult, because some parameters were not known, bubble trajectory and bubble shape, for example, as direct optical methods cannot be used, like it is done with water experiments. (author)

  14. Equilibrium of current driven rotating liquid metal

    International Nuclear Information System (INIS)

    Velikhov, E.P.; Ivanov, A.A.; Zakharov, S.V.; Zakharov, V.S.; Livadny, A.O.; Serebrennikov, K.S.

    2006-01-01

    In view of great importance of magneto-rotational instability (MRI) as a fundamental mechanism for angular momentum transfer in magnetized stellar accretion disks, several research centers are involved in experimental study of MRI under laboratory conditions. The idea of the experiment is to investigate the rotation dynamics of well conducting liquid (liquid metal) between two cylinders in axial magnetic field. In this Letter, an experimental scheme with immovable cylinders and fluid rotation driven by radial current is considered. The analytical solution of a stationary flow was found taking into account the external current. Results of axially symmetric numerical simulations of current driven fluid dynamics in experimental setup geometry are presented. The analytical solution and numerical simulations show that the current driven fluid rotation in axial magnetic field provides the axially homogeneous velocity profile suitable for MRI study in classical statement

  15. Generation of electricity using liquid metal magnetohydrodynamics

    International Nuclear Information System (INIS)

    Goodwin, F.E.

    1992-01-01

    With liquid metal magnetohydrodynamics, a column of molten lead is passed through a magnetic field, thereby generating a voltage potential according to Faraday's law. The molten lead is propelled through a closed loop by steam from water injected just above where the lead is heated at the bottom of the loop. This water in turn boils explosively, propelling the lead upward through the loop and past the point where the steam escapes through a separator. Electricity can be generated more efficiently from steam with LMMHD than with conventional turbines. With the DC current generated by LMMHD, industriell cogeneration is seen as the most likely application, where the byproduct steam still has enough pressure to also power other steam-driven machinery. Furthermore, the byproduct steam is essentially lead-free since the operating temperature of the LMMHD generator is well below the temperature where lead could dissolve into the steam. (orig.) [de

  16. Flow balancing in liquid metal blankets

    International Nuclear Information System (INIS)

    Tillack, M.S.; Morley, N.B.

    1995-01-01

    Non-uniform flow distribution between parallel channels is one of the most serious concerns for self-cooled liquid metal blankets with electrically insulated walls. We show that uncertainties in flow distribution can be dramatically reduced by relatively simple design modifications. Several design features which impose flow uniformity by electrically coupling parallel channels are surveyed. Basic mechanisms for ''flow balancing'' are described, and a particular self-regulating concept using discrete passive electrodes is proposed for the US ITER advanced blanket concept. Scoping calculations suggest that this simple technique can be very powerful in equalizing the flow, even with massive insulator failures in individual channels. More detailed analyses and experimental verification will be required to demonstrate this concept for ITER. (orig.)

  17. Study of liquid metal mixed convection in cavities

    International Nuclear Information System (INIS)

    Abadie, Philippe.

    1979-10-01

    This study has enabled some results to be obtained on the flow of liquid metals in cavities. The effects of different adimensional parameters characteristic of mixed convection flows were experimentally demonstrated. In the case of a roof heated cavity, three zones were distinguished: the mixing zone at the channel exit, a quasi constant temperature recirculation zone and a stratified zone at the top of the cavity. The thickness of this last region depends on natural convection effects: it disappears completely in a pure forced convection regime. A simple model using a critical Richardson number concept was developed in order to be able to predict the thickness of this region. Heat transfer correlation formulas were established both for the heated roof and forward direction heated wall cases. Some data was also obtained on temperature fluctuations for both cases. The different topics investigated are useful for defining heat transfers in certain regions of fast neutron sodium cooled reactors. A more extensive program is currently being developed in order to be able to investigate a wider range of variations in the above mentioned parameters and to more closely approximate reactor vessels [fr

  18. Effects of Liquid Metal Fin on Critical Heat Flux under IVR-ERVC Condition

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-05-15

    The molten fuel is relocated to bottom of reactor vessel after core is damaged and not cooled continuously. In-vessel retention through external reactor vessel cooling (IVR-ERVC) is presented to terminate the progression of accidents by removing the decay heat. IVR-ERVC is suitable for small size reactors like AR-600, AP-1000. There is uncertainty for high power reactor like APR-1400 and CAP-1400. This uncertainty originates from the thermal margin between the CHF value and real heat flux on the reactor vessel under severe accidents. The main mechanism of heat removal on IVR-ERVC strategy is boiling on the outer wall of reactor vessel. The boiling heat transfer is limited due to the CHF phenomenon. There should be an enough margin for preventing the CHF in boiling heat transfer systems. The CHF tests for IVR-ERVC system were conducted to confirm or increase the thermal margin. The design of thermal insulator was changed to vent the vapor smoothly. Forming the coating layer on the vessel surface was proposed to enhance the CHF margin. The liquid metal was designed to flood the space around the reactor vessel. The liquid metal has high boiling point and superb thermal conductivity in comparison with the coolant. In this work, experimental tests were conducted to validate the CFD results about the IVR-ERVC system with liquid metal. The behavior of vapor was observed to predict the tendency of CHF increase with small-scaled facility to simulate the IVR-ERVC system.

  19. Liquid metals as electrodes in polymer light emitting diodes

    NARCIS (Netherlands)

    Andersson, G.G.; Gommans, H.H.P.; Denier van der Gon, A.W.; Brongersma, H.H.

    2003-01-01

    We demonstrate that liquid metals can be used as cathodes in light emitting diodes (pLEDs). The main difference between the use of liquid cathodes and evaporated cathodes is the sharpness of the metal–polymer interface. Liquid metal cathodes result in significantly sharper metal–organic interfaces

  20. Ionic imbalance induced self-propulsion of liquid metals

    Science.gov (United States)

    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.

  1. Measurement of the differential pressure of liquid metals

    Science.gov (United States)

    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)

  2. Microstructured liquid metal electron and ion sources (MILMES/MILMIS)

    Energy Technology Data Exchange (ETDEWEB)

    Mitterauer, J [Technische Universitaet Wien (Austria). Institut fuer Allgemeine Elektrotechnik und Elektronik

    1997-12-31

    Ion or electron beams can be emitted from liquid metal wetted needles, or from capillaries or slits into which the liquid metal is allowed to flow. Large-area liquid metal field emission sources have been proposed recently, using either two-dimensional, regular arrays of cones or capillaries, or even a substrate with an intrinsically microstructured surface covered by a liquid metal film. This latter concept has been realized in a pilot experiment by in situ wicking and wetting of a porous sintered metal disc. Microstructured liquid metal ion or electron sources are capable of operating in a pulsed mode at a current level which is orders of magnitude above that for steady-state operation. (author). 3 figs., 10 refs.

  3. Measurement of the differential pressure of liquid metals

    International Nuclear Information System (INIS)

    Metz, H.J.

    1975-01-01

    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

  4. European liquid metal thermal-hydraulics R and D: present and future

    International Nuclear Information System (INIS)

    Roelofs, F.; Batta, A.; Bandini, G.; Van Tichelen, K.; Gerschenfeld, A.; Cheng, X.

    2014-01-01

    A large role is attributed in the future within the European Sustainable Nuclear Energy Technology Platform (SNE-TP) and especially the underlying European Sustainable Nuclear Industry Initiative (ESNII) to the application of fast reactors for sustainable nuclear energy production. Specifically, fast reactors are considered attractive because of their possibility to use natural resources efficiently and to reduce the volume and lifetime of nuclear waste. Currently four demonstration projects have a promising outlook in Europe, i.e. the ASTRID project in France, the MYRRHA project in Belgium, the ALFRED project developed in Europe and to be built in Romania, and the ELECTRA project in Sweden. Sodium and lead(-alloys) are envisaged as coolants for these reactors. Obviously, in the development of these reactors, thermal-hydraulics is recognized as a key challenge with emphasis on safety issues. This paper will discuss the present development status of liquid metal cooled reactor thermal-hydraulics as an outcome of the European 7. framework programme THINS (Thermal-Hydraulics for Innovative Nuclear Systems) project. The main project results with respect to liquid metal cooled reactors will be summarized, i.e. turbulence heat transfer model development, fuel assembly analysis, pool thermal-hydraulics, system behaviour, multi-phase physics, and multiscale thermal-hydraulics simulation. In conclusion, the main challenges for future developments will be indicated. Emphasis will be put on the important experimental and numerical challenges. (authors)

  5. Liquid Metals as Plasma-facing Materials for Fusion Energy Systems: From Atoms to Tokamaks

    Energy Technology Data Exchange (ETDEWEB)

    Stone, Howard A. [Princeton Univ., NJ (United States); Koel, Bruce E. [Princeton Univ., NJ (United States); Bernasek, Steven L. [Princeton Univ., NJ (United States); Carter, Emily A. [Princeton Univ., NJ (United States); Debenedetti, Pablo G. [Princeton Univ., NJ (United States); Panagiotopoulos, Athanassios Z. [Princeton Univ., NJ (United States)

    2017-06-23

    The objective of our studies was to advance our fundamental understanding of liquid metals as plasma-facing materials for fusion energy systems, with a broad scope: from atoms to tokamaks. The flow of liquid metals offers solutions to significant problems of the plasma-facing materials for fusion energy systems. Candidate metals include lithium, tin, gallium, and their eutectic combinations. However, such liquid metal solutions can only be designed efficiently if a range of scientific and engineering issues are resolved that require advances in fundamental fluid dynamics, materials science and surface science. In our research we investigated a range of significant and timely problems relevant to current and proposed engineering designs for fusion reactors, including high-heat flux configurations that are being considered by leading fusion energy groups world-wide. Using experimental and theoretical tools spanning atomistic to continuum descriptions of liquid metals, and bridging surface chemistry, wetting/dewetting and flow, our research has advanced the science and engineering of fusion energy materials and systems. Specifically, we developed a combined experimental and theoretical program to investigate flows of liquid metals in fusion-relevant geometries, including equilibrium and stability of thin-film flows, e.g. wetting and dewetting, effects of electromagnetic and thermocapillary fields on liquid metal thin-film flows, and how chemical interactions and the properties of the surface are influenced by impurities and in turn affect the surface wetting characteristics, the surface tension, and its gradients. Because high-heat flux configurations produce evaporation and sputtering, which forces rearrangement of the liquid, and any dewetting exposes the substrate to damage from the plasma, our studies addressed such evaporatively driven liquid flows and measured and simulated properties of the different bulk phases and material interfaces. The range of our studies

  6. Three-dimensional MHD [magnetohydrodynamic] flows in rectangular ducts of liquid-metal-cooled blankets

    International Nuclear Information System (INIS)

    Hua, T.Q.; Walker, J.S.; Picologlou, B.F.; Reed, C.B.

    1988-07-01

    Magnetohydrodynamic flows of liquid metals in rectangular ducts with thin conducting walls in the presence of strong nonuniform transverse magnetic fields are examined. The interaction parameter and Hartmann number are assumed to be large, whereas the magnetic Reynolds number is assumed to be small. Under these assumptions, viscous and inertial effects are confined in very thin boundary layers adjacent to the walls. A significant fraction of the fluid flow is concentrated in the boundary layers adjacent to the side walls which are parallel to the magnetic field. This paper describes the analysis and numerical methods for obtaining 3-D solutions for flow parameters outside these layers, without solving explicitly for the layers themselves. Numerical solutions are presented for cases which are relevant to the flows of liquid metals in fusion reactor blankets. Experimental results obtained from the ALEX experiments at Argonne National Laboratory are used to validate the numerical code. In general, the agreement is excellent. 5 refs., 14 figs

  7. MHD and heat transfer benchmark problems for liquid metal flow in rectangular ducts

    International Nuclear Information System (INIS)

    Sidorenkov, S.I.; Hua, T.Q.; Araseki, H.

    1994-01-01

    Liquid metal cooling systems of a self-cooled blanket in a tokamak reactor will likely include channels of rectangular cross section where liquid metal is circulated in the presence of strong magnetic fields. MHD pressure drop, velocity distribution and heat transfer characteristics are important issues in the engineering design considerations. Computer codes for the reliable solution of three-dimensional MHD flow problems are needed for fusion relevant conditions. Argonne National Laboratory and The Efremov Institute have jointly defined several benchmark problems for code validation. The problems, described in this paper, are based on two series of rectangular duct experiments conducted at ANL; one of the series is a joint ANL/Efremov experiment. The geometries consist of variation of aspect ratio and wall thickness (thus wall conductance ratio). The transverse magnetic fields are uniform and nonuniform in the axial direction

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

    Science.gov (United States)

    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.

  9. Dynamic stabilization of imploding liquid metal liner

    International Nuclear Information System (INIS)

    Itoh, Yasuyuki; Fujiie, Yoichi

    1979-01-01

    The rotational stabilization has been proposed against the Rayleigh-Taylor instability of the imploding liquid metal liner. In this paper, the discussion is made on the possibility of the dynamic stabilization by applying the oscillating azimuthal magnetic field in addition to the axial field. In contrast to the rotational stabilization, the required (field) energy for this stabilization is also used for the liner driving or the plasma confinement. In the analysis, the liner subjected to the acceleration is assumed to be infinitely long, at rest and have the situation at the start of the implosion or turnaround. At turnaround, the existence of the plasma is taken into account. The perturbed motion of the liner is discussed with a linear stability analysis. Results are as follows: (1) The dynamic stabilization at the start of the implosion is possible if the distance from the conducting wall to the liner outer surface is comparable with or less than the liner thickness. (2) At turnaround, the stability is improved with decreasing the ratio of the plasma radius to that of the liner inner surface however the kink mode (m = 1) cannot be suppressed. (author)

  10. Production of Liquid Metal Spheres by Molding

    Directory of Open Access Journals (Sweden)

    Mohammed G. Mohammed

    2014-10-01

    Full Text Available This paper demonstrates a molding technique for producing spheres composed of eutectic gallium-indium (EGaIn with diameters ranging from hundreds of microns to a couple millimeters. The technique starts by spreading EGaIn across an elastomeric sheet featuring cylindrical reservoirs defined by replica molding. The metal flows into these features during spreading. The spontaneous formation of a thin oxide layer on the liquid metal keeps the metal flush inside these reservoirs. Subsequent exposure to acid removes the oxide and causes the metal to bead up into a sphere with a size dictated by the volume of the reservoirs. This technique allows for the production and patterning of droplets with a wide range of volumes, from tens of nanoliters up to a few microliters. EGaIn spheres can be embedded or encased subsequently in polymer matrices using this technique. These spheres may be useful as solder bumps, electrodes, thermal contacts or components in microfluidic devices (valves, switches, pumps. The ease of parallel-processing and the ability to control the location of the droplets during their formation distinguishes this technique.

  11. Turbulent temperature fluctuations in liquid metals

    International Nuclear Information System (INIS)

    Lawn, C.J.

    1977-01-01

    Examination of experimental data for the spectral distribution of velocity (u and v) and temperature (theta) fluctuations in the fully turbulent region of heated pipe-flow has suggested a schematic representation which incorporates the essential features. Evidence is cited to suggest that the -vtheta correlation coefficient maintains higher values that the uv coefficient at wave-numbers in the inertial subrange. The theory of Batchelor, Howells and Townsend, and limited evidence from experiments in mercury, then suggests the form of the theta 2 spectra and -vtheta cross-spectra in liquid metals. From this information, a limiting Peclet number is deduced, above which the correlation coefficient of v and theta should be a fairly weak function of Pe alone. An attempt to check this inference from published data for the RMS level of temperature fluctuations, and for the turbulent Prandtl number, proves inconclusive, because many of the correlation coefficients so estimated have values greater than unity. It is concluded that all these results for theta tilde must therefore be in error. However, since there is no evidence of very low correlation coefficients, they almost certainly lie in the range 0.5 multiply/divide 2 over a large proportion of the radius. Thus theta tilde can be estimated for any fluid in which the fluctuations are induced by uniform heating, at least to within a factor of 2, using the analysis presented. (author)

  12. Liquid metal liner implosion systems with blade lattice for fusion

    International Nuclear Information System (INIS)

    Itoh, Yasuyuki; Fujiie, Yoichi

    1980-01-01

    In this paper, the liquid liner implosion systems with the blade lattice is proposed for the rotational stabilization of the liner inner surface which is facing a plasma in a fusion reactor. The blades are electrically conducting and inclined to the radial direction. Its major function is either acceleration or deceleration of the liner in the azimuthal direction. This system enables us to exclude the rotary mechanism for the liner rotation. In this system, the liner is formed as an annular flow of a liquid metal (the waterfall concept). Results show that there is no significant difference of the energy cost for the stabilization compared with the earlier proposed system where a liner is rotated rigidly before implosion. Furthermore, the application of the rotating blade lattice makes it possible to reduce the rotational kinetic energy required for the stabilization at turnaround, where the lattice acts as an impeller in the initial liner rotation. There is an optimum blade angle to maximize the compressed magnetic field energy inside the liner for a given driving energy. (author)

  13. Diverse transformations of liquid metals between different morphologies.

    Science.gov (United States)

    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. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  14. Dynamic interactions of Leidenfrost droplets on liquid metal surface

    Science.gov (United States)

    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.

  15. Magnetohydrodynamic instability of a cylindrical liquid-metal brush

    International Nuclear Information System (INIS)

    Hong, S.H.; Wilhelm, H.E.

    1976-01-01

    The stability of a homopolar generator brush, consisting of a liquid-metal-filled cavity between rotating (rotor) and fixed (stator) cylinder electrodes, is analyzed in the presence of radial current transport and an axial homogeneous magnetic field. Within the frame of linear magnetohydrodynamics, it is shown that the liquid-metal flow in the brush is always unstable if the brush transports current. In the absence of current flow (infinite load) the axial magnetic field stabilizes the liquid-metal flow in the brush if the magnetic energy density is larger than a certain fraction of the energy density of the rotating fluid

  16. The coalescence of heterogeneous liquid metal on nano substrate

    Science.gov (United States)

    Wang, Long; Li, Yifan; Zhou, Xuyan; Li, Tao; Li, Hui

    2017-06-01

    Molecular dynamics simulation has been performed to study the asymmetric coalescence of heterogeneous liquid metal on graphene. Simulation results show that the anomalies in the drop coalescence is mainly caused by the wettability of heterogeneous liquid metal. The silver atoms incline to distribute on the outer layer of the gold and copper droplets, revealing that the structure is determined by the interaction between different metal atoms. The coalescence and fusion of heterogeneous liquid metal drop can be predicted by comparing the wettability and the atomic mass of metallic liquid drops, which has important implications in the industrial application such as ink-jet printing and metallurgy.

  17. Analysis of postulated events for the revised ALMR/PRISM design

    International Nuclear Information System (INIS)

    Slovik, G.C.; Tuyle, G.J. Van

    1992-01-01

    The Nuclear Regulatory Commission (NRC) is continuing a pre-application review of the 471 MWt, liquid metal reactor, PRISM, by General Electric, with Brookhaven National Laboratory providing technical support. The revised design has been evaluated, using the SSC code, for an unscrammed loss of heat sink (ULOHS), an unscrammed loss of flow (ULOF) with and without the Gas Expansion Modules (GEMs), and a 40 cents unscrammed transient overpower (UTOP) event. The feedback effects for U-27Pu-10Zr metal fuel were modeled in SSC. The ULOHS accident was determined to be a benign event for the design, with the reactor power transitioning down to a decay heat level within 500s. The power during the postulated ULOF events, with the GEMs functioning, transitioned to decay heat levels without fuel damage, and included a 300K margin to sodium saturation. The case without the GEMs had only a 160K margin to sodium saturation and higher fuel temperatures. In addition, the clad was predicted to quickly pass through the eutectic phase (between fuel and clad), and some clad wastage would result. The 40 cents UTOP was predicted to raise the power to 1.8 rated, which later stabilized near 1.2 times full power. SSC predicted some localized fuel melting for the event, but the significance of this localized damage has not yet been determined. If necessary, the vendor has options to reduce the maximum reactivity insertion below 40 cents

  18. Analysis of postulated events for the revised ALMR/PRISM design

    International Nuclear Information System (INIS)

    Slovik, G.C.; Van Tuyle, G.J.

    1991-01-01

    The Nuclear Regulatory Commission (NRC) is continuing a pre-application review of the 471 MWt, Liquid Metal Reactor, PRISM by General Electric, with Brookhaven National Laboratory providing technical support. The revised design has been evaluated, using the SSC code, for an unscrammed loss of heat sink (ULOHS), an unscrammed loss of flow (ULOF) with and without the Gas Expansion Modules (GEMs), and a 40 cents unscrammed transient overpower (UTOP) event. The feedback effects for U-27Pu-10Zr metal fuel were modeled in SSC. The ULOHS accident was determined to be a benign event for the design, with the reactor power transitioning down to a decay heat level within 500s. The power during the postulated ULOF events, with the GEMs functioning, transitioned to decay heat levels without fuel damage, and included a 300K margin to sodium saturation. The case without the GEMs had only a 160K margin to sodium saturation and higher fuel temperatures. In addition, the clad was predicted to quickly pass through the eutectic phase (between fuel and clad), and some clad wastage would result. The 40 cents UTOP was predicted to raise the power to 1.8 rated, which later stabilized near 1.2 times full power. SSC predicted some localized fuel melting for the event, but the significance of this localized damage has not yet been determined. If necessary, the vendor has options to reduce the maximum reactivity insertion below 40 cents

  19. Heavy liquid metal cooled FBR. Results 2003

    International Nuclear Information System (INIS)

    Hayahune, Hiroki; Enuma, Yasuhiro; Soman, Yoshindo; Konomura, Mamoru; Mizuno, Tomoyasu

    2004-08-01

    Concepts of the reactor, SG and main coolant pump have been studied considering maintainability and aseismic capability, which is a medium size pool type lead-bismuth cooled reactor. The results are following. (1) Reconsideration of reactor design concepts concerning maintainability: In pursuit of good reactor maintainability, the structural concepts of SG, UIS and core support structures have been changed to be drawn up above the upper area of the reactor system. After a few decade of interval, lead-bismuth inventory in the reactor vessel shall be fully drained for easy ISI operation of in-vessel main components such as core support structures. From the viewpoint of the reactor aseismic capability, the axial length of reactor vessel was reduced and the reactor vessel support location was changed from the top handing to the circumference of the vessel. (2) SG concept selection in conjunction with a compact reactor vessel: The concept of SG consisting of a once through type with helical coil tube is selected. 6 units of a small scale SG are arranged on a reactor roof deck along the peripheral direction, in addition to 3 units of a centrifugal mechanical pump. (3) Aseismic structural integrity of the reactor components: Aseismic structural integrity of the reactor vessel, core support structures, UIS, FHM, SG and the main pumps has been vigorously examined respectively. These components besides FHM could keep the aseismic structural integrity for strong S2 earthquake under the design condition. FHM could also keep the integrity for S1 earthquake. (4) Safety evaluation: Thermal transients following loss of flow type accident due to plant total blackout and typical manual reactor trip incident, have been evaluated to assure the pant safety design, by analyzing thermal hydraulic behavior of transients concerning core flow rate and temperatures of the plant cooling system. Loss of flow accident due to plant total blackout: The reactor coolant pumps shall be tripped and

  20. Engineering: Liquid metal pumped at a record temperature

    Science.gov (United States)

    Lambrinou, Konstantina

    2017-10-01

    Although liquid metals are effective fluids for heat transfer, pumping them at high temperatures is limited by their corrosiveness to solid metals. A clever pump design addresses this challenge using only ceramics. See Article p.199