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Sample records for advanced sodium cooled

  1. Challenges in licensing a sodium-cooled advanced recycling reactor

    International Nuclear Information System (INIS)

    Levin, Alan E.

    2008-01-01

    As part of the Global Nuclear Energy Partnership (GNEP), the U.S. Department of Energy (DOE) has focused on the use of sodium-cooled fast reactors (SFRs) for the destruction of minor actinides derived from used reactor fuel. This approach engenders an array of challenges with respect to the licensing of the reactor: the U.S. Nuclear Regulatory Commission (NRC) has never completed the review of an application for an operating license for a sodium-cooled reactor. Moreover, the current U.S. regulatory structure has been developed to deal almost exclusively with light-water reactor (LWR) designs. Consequently, the NRC must either (1) develop a new regulatory process for SFRs, or (2) reinterpret the existing regulations to apply them, as appropriate, to SFR designs. During the 1980s and 1990s, the NRC conducted preliminary safety assessments of the Sodium Advanced Fast Reactor (SAFR) and the Power Reactor Innovative Small Module (PRISM) designs, and in that context, began to consider how to apply LWR-based regulations to SFR designs. This paper builds on that work to consider the challenges, from the reactor designer's point of view, associated with licensing an SFR today, considering (1) the evolution of SFR designs, (2) the particular requirements of reactor designs to meet GNEP objectives, and (3) the evolution of NRC regulations since the conclusion of the SAFR and PRISM reviews. (author)

  2. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  3. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Sun, K.

    2012-09-15

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  4. Analysis of advanced sodium-cooled fast reactor core designs with improved safety characteristics

    International Nuclear Information System (INIS)

    Sun, K.

    2012-09-01

    Currently, the large majority of nuclear power plants are operated with thermal-neutron spectra and need regular fuel loading of enriched uranium. According to the identified conventional uranium resources and their current consumption rate, only about 100 years’ nuclear fuel supply is foreseen. A reactor operated with a fast-neutron spectrum, on the other hand, can induce self-sustaining, or even breeding, conditions for its inventory of fissile material, which effectively allow it, after the initial loading, to be refueled using simply natural or depleted uranium. This implies a much more efficient use of uranium resources. Moreover, minor actinides become fissionable in a fast-neutron spectrum, enabling full closure of the fuel cycle and leading to a minimization of long-lived radioactive wastes. The sodium-cooled fast reactor (SFR) is one of the most promising candidates to meet the Generation IV International Forum (GIF) declared goals. In comparison to other Generation IV systems, there is considerable design experience related to the SFR, and also more than 300 reactor years of practical operation. As a fast-neutron-spectrum system, the long-term operation of an SFR core in a closed fuel cycle will lead to an equilibrium state, where both reactivity and fuel mass flow stabilize. Although the SFR has many advantageous characteristics, it has one dominating neutronics drawback: there is generally a positive reactivity effect when sodium coolant is removed from the core. This so-called sodium void effect becomes even stronger in the equilibrium closed fuel cycle. The goal of the present doctoral research is to improve the safety characteristics of advanced SFR core designs, in particular, from the viewpoint of the positive sodium void reactivity effect. In this context, particular importance has been given to the dynamic core behavior under a hypothetical unprotected loss-of-flow (ULOF) accident scenario, in which sodium boiling occurs. The proposed

  5. Sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hokkyo, N; Inoue, K; Maeda, H

    1968-11-21

    In a sodium cooled fast neutron reactor, an ultrasonic generator is installed at a fuel assembly hold-down mechanism positioned above a blanket or fission gas reservoir located above the core. During operation of the reactor an ultrsonic wave of frequency 10/sup 3/ - 10/sup 4/ Hz is constantly transmitted to the core to resonantly inject the primary bubble with ultrasonic energy to thereby facilitate its growth. Hence, small bubbles grow gradually to prevent the sudden boiling of sodium if an accident occurs in the cooling system during operation of the reactor.

  6. Conceptual design of advanced central receiver power systems sodium-cooled receiver concept. Volume 2, Book 2. Appendices. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1979-03-01

    The appendices include: (A) design data sheets and P and I drawing for 100-MWe commercial plant design, for all-sodium storage concept; (B) design data sheets and P and I drawing for 100-MWe commercial plant design, for air-rock bed storage concept; (C) electric power generating water-steam system P and I drawing and equipment list, 100-MWe commercial plant design; (D) design data sheets and P and I drawing for 281-MWe commercial plant design; (E) steam generator system conceptual design; (F) heat losses from solar receiver surface; (G) heat transfer and pressure drop for rock bed thermal storage; (H) a comparison of alternative ways of recovering the hydraulic head from the advanced solar receiver tower; (I) central receiver tower study; (J) a comparison of mechanical and electromagnetic sodium pumps; (K) pipe routing study of sodium downcomer; and (L) sodium-cooled advanced central receiver system simulation model. (WHK)

  7. An Advanced Sodium-Cooled Fast Reactor Core Concept Using Uranium-Free Metallic Fuels for Maximizing TRU Burning Rate

    Directory of Open Access Journals (Sweden)

    Wuseong You

    2017-12-01

    Full Text Available In this paper, we designed and analyzed advanced sodium-cooled fast reactor cores using uranium-free metallic fuels for maximizing burning rate of transuranics (TRU nuclides from PWR spent fuels. It is well known that the removal of fertile nuclides such as 238U from fuels in liquid metal cooled fast reactor leads to the degradation of important safety parameters such as the Doppler coefficient, coolant void worth, and delayed neutron fraction. To resolve the degradation of the Doppler coefficient, we considered adding resonant nuclides to the uranium-free metallic fuels. The analysis results showed that the cores using uranium-free fuels loaded with tungsten instead of uranium have a significantly lower burnup reactivity swing and more negative Doppler coefficients than the core using uranium-free fuels without resonant nuclides. In addition, we considered the use of axially central B4C absorber region and moderator rods to further improve safety parameters such as sodium void worth, burnup reactivity swing, and the Doppler coefficient. The results of the analysis showed that the final design core can consume ~353 kg per cycle and satisfies self-controllability under unprotected accidents. The fuel cycle analysis showed that the PWR–SFR coupling fuel cycle option drastically reduces the amount of waste going to repository and the SFR burner can consume the amount of TRUs discharged from 3.72 PWRs generating the same electricity.

  8. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  9. An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar

    2014-03-01

    Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

  10. A 100 MWe Advanced Sodium-cooled Fast Reactor (AFR-100)

    International Nuclear Information System (INIS)

    Grandy, C.; Kim, T.K.; Jin, E.

    2013-01-01

    • AFR-100 Design development is continuing in the U.S.; • Various innovations are included in the design to understand their feasibility; • Engineering and safety analyses have been performed that demonstrate the inherent safety characteristics of the AFR-100 design during severe accidents; • R&D is being performed on a number of the innovations such as advanced materials, compact fuel handing system, advanced energy conversion system, advanced core design, etc

  11. BN800: The advanced sodium cooled fast reactor plant based on close fuel cycle

    International Nuclear Information System (INIS)

    Wu Xingman

    2011-01-01

    As one of the advanced countries with actually fastest reactor technology, Russia has always taken a leading role in the forefront of the development of fast reactor technology. After successful operation of BN600 fast reactor nuclear power station with a capacity of six hundred thousand kilowatts of electric power for nearly 30 years, and after a few decades of several design optimization improved and completed on its basis, it is finally decided to build Unit 4 of Beloyarsk nuclear power station (BN800 fast reactor power station). The BN800 fast reactor nuclear power station is considered to be the project of the world's most advanced fast reactor nuclear power being put into implementation. The fast reactor technology in China has been developed for decades. With the Chinese pilot fast reactor to be put into operation soon, the Chinese model fast reactor power station has been put on the agenda. Meanwhile, the closed fuel cycle development strategy with fast reactor as key aspect has given rise to the concern of experts and decision-making level in relevant areas. Based on the experiences accumulated in many years in dealing the Sino-Russian cooperation in fast reactor technology, with reference to the latest Russian published and authoritative literatures regarding BN800 fast reactor nuclear power station, the author compiled this article into a comprehensive introduction for reference by leaders and experts dealing in the related fields of nuclear fuel cycle strategy and fast reactor technology development researches, etc. (authors)

  12. Pressure drop and heat transfer in the sodium to air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kang, H.; Eoh, J.; Cha, J.; Kim, S.

    2011-01-01

    A numerical study was performed to investigate the thermal and hydraulic characteristics and build up design model of the AHX (sodium-to-air heat exchanger) unit of a sodium-cooled fast reactor. Helical-coiled tube banks in the AHX were modeled as porous media and simulated heat and momentum transfer. Two-dimensional flow characteristic appeared at the most region of AHX annulus. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX were evaluated and compared with Zhukauskas empirical correlations. (author)

  13. Sodium-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hammers, H.W.

    1982-01-01

    The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

  14. Numerical study on pressure drop and heat transfer for designing sodium-to-air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kang, Hie-Chan; Eoh, Jae-Hyuk; Cha, Jae-Eun; Kim, Seong-O.

    2013-01-01

    Highlights: ► Numerical simulation for the heat flow characteristic of the sodium-to-air heat exchanger (AHX) and tube banks. ► Parallelogram tube banks showed almost similar thermal and hydraulic characteristics to the rectangular tube banks. ► Pressure drop and heat transfer of the staggered and rectangular tube banks compared with Zhukauskas’ correlation. ► AHX was modeled as porous media and suggested design guide to enhance the performance. - Abstract: A numerical study is performed to investigate the thermal and hydraulic characteristics and build up design model of the AHX (sodium-to-air heat exchanger) unit of a sodium-cooled fast reactor. Helical-coiled tube banks in the AHX are modeled as porous media and simulated heat and momentum transfer by a commercial program. Two-dimensional flow characteristic appears differently at the inlet region of the AHX annulus, and the required length of the inlet region is shorter for an inlet having a 45 degree chamber or a round shape than for one with a perpendicular corner. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX are evaluated and discussed. Pressure drop and heat transfer shows similar trends and underestimated values, respectively, when compared with Zhukauskas empirical correlations. The parallelogram tube bank shows similar results to the rectangular arrangement.

  15. Water experiments on thermal striping in reactor vessel of advanced sodium-cooled fast reactor. Influence of flow collector of backup CR guide tube

    International Nuclear Information System (INIS)

    Kobayashi, Jun; Ezure, Toshiki; Tanaka, Masaaki; Kamide, Hideki

    2016-01-01

    Design study of an advanced large-scale sodium-cooled fast reactor (SFR) has been conducted in JAEA. In the region between the bottom of the Upper Internal Structure (UIS) and the core outlet, the hot sodium from the fuel subassembly mixes with the cold sodium from the neighbor control rod (CR) channel. Therefore, temperature fluctuation due to mixing fluids at different temperatures may cause high cycle thermal fatigue at the bottom of the UIS. In the advanced design, installation of a flow guide structure named Flow-Collector (FC) to the backup control rod (BCR) guide tube is considered to enhance reliable operation of self-actuated shutdown system (SASS) and to ensure reactor shutdown operation. Previously, water experiments without the FC model had been examined in JAEA to investigate effective countermeasures to the significant temperature fluctuation generation at the bottom of the UIS. Since the FC may affect the thermal mixing behavior at the bottom of the UIS, influence of the FC on characteristics of the temperature fluctuation around the BCR channels was investigated using a water experimental facility with structure model of the FC. Through the experiment, small influence of the FC on the temperature fluctuation distribution at the bottom of the UIS was indicated. (author)

  16. Proposed Advanced Reactor Adaptation of the Standard Review Plan NUREG-0800 Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Poore, III, Willis P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sofu, Tanju [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-03-01

    This report proposes adaptation of the previous regulatory gap analysis in Chapter 4 (Reactor) of NUREG 0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The proposed adaptation would result in a Chapter 4 review plan applicable to certain advanced reactors. This report addresses two technologies: the sodium-cooled fast reactor (SFR) and the modular high temperature gas-cooled reactor (mHTGR). SRP Chapter 4, which addresses reactor components, was selected for adaptation because of the possible significant differences in advanced non-light water reactor (non-LWR) technologies compared with the current LWR-based description in Chapter 4. SFR and mHTGR technologies were chosen for this gap analysis because of their diverse designs and the availability of significant historical design detail.

  17. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective ...

  18. A resting bottom sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Costes, D.

    2012-01-01

    This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

  19. The BN-1800 advanced sodium cooled fast reactor meeting requirements to nuclear power engineering of the XXI century

    International Nuclear Information System (INIS)

    Poplavskij, V.M.; Tsibulya, A.M.; Kamaev, A.A.

    2004-01-01

    Basic principles and direction of the elaboration of sodium fast reactor BN-1800 are discussed. The elaboration of the BN-1800 reactor is based on the scientific justified technical feasibilities of BN-350, BN-600 and BN-800 reactors. Descriptions of power blocks and reactor core of the elaborated reactor are presented. Characteristics of the BN-1800 steam generator are given. Safety of reactor unit is estimated, fundamental technical and economic indexes of BN-1800 are discussed. Economic indexes of the BN-1800 reactor are noted to be on the level of WWER-1000 and WWER-1500 reactors [ru

  20. Development of Advanced 9Cr Ferritic-Martensitic Steels and Austenitic Stainless Steels for Sodium-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Sham, T.-L.; Tan, L.; Yamamoto, Y.

    2013-01-01

    Summary of creep, thermal aging and weldability aspects: • The creep resistance of advanced 9Cr FM steels was greatly enhanced by optimizing their compositions as well as by using TMT. – Up to about 700 times increase in creep life, compared to Gr91, was achieved under the accelerated test conditions at 600°C. • The increased density of ultrafine precipitates facilitated the increase in strength and thermal aging resistance, leading to the improved creep resistance. • Properties of four candidate austenitic alloys, HT-UPS, NF709, and two modified HT-UPS alloy (designated Alloys A and B), have been evaluated and compared with 316H. – Alloys A and B showed successful improvement in weldability. – Only a little difference in thermal stability of the alloys in solution annealed conditions. 10% cold work increased the yield strength of the alloys for more than 200% compared to the HT-UPS without cold work. – HT-UPS exhibited the best creep properties among the alloys with and without cold work, and NF709 followed

  1. Hydrogen detector for sodium cooled reactors

    International Nuclear Information System (INIS)

    Roy, P.; Rodgers, D.N.

    1975-01-01

    An improved hydrogen detector for use in sodium cooled reactors is described. The improved detector basically comprises a diffusion tube of either pure nickel or stainless steel having a coating on the vacuum side (inside) of a thin layer of refractory metal, e.g., tungsten or molybdenum. The refractory metal functions as a diffusion barrier in the path of hydrogen diffusing from the sodium on the outside of the detector into the vacuum on the inside, thus by adjusting the thickness of the coating, it is possible to control the rate of permeation of hydrogen through the tube, thereby providing a more stable detector. (U.S.)

  2. Design considerations for economically competitive sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Zhang, Hongbin; Zhao, Haihua; Mousseau, Vincent; Szilard, Ronaldo

    2009-01-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phenix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design. (author)

  3. Design study on sodium-cooled large-scale reactor

    International Nuclear Information System (INIS)

    Shimakawa, Yoshio; Nibe, Nobuaki; Hori, Toru

    2002-05-01

    In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled large-scale reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled large-scale reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2001, which is the first year of Phase 2. In the JFY2001 design study, a plant concept has been constructed based on the design of the advanced loop type reactor, and fundamental specifications of main systems and components have been set. Furthermore, critical subjects related to safety, structural integrity, thermal hydraulics, operability, maintainability and economy have been examined and evaluated. As a result of this study, the plant concept of the sodium-cooled large-scale reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

  4. Startup of the FFTF sodium cooled reactor

    International Nuclear Information System (INIS)

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

    1981-03-01

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

  5. Single- and two-phase flow modeling for coupled neutronics / thermal-hydraulics transient analysis of advanced sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Chenu, A.

    2011-10-01

    Nuclear power is nowadays in the front rank as regards helping to meet the growing worldwide energy demand while avoiding an excessive increase in greenhouse gas emissions. However, the operating nuclear power plants are mainly thermal-neutron reactors and, as such, can not be maintained on the basis of the currently identified uranium resources beyond one century at the present consumption rate. Sustainability of nuclear power thus involves closure of the fuel cycle through breeding. With a uranium-based fuel, breeding can only be achieved using a fast-neutron reactor. Sodium-cooled fast reactor (SFR) technology benefits from 400 reactor-years of accumulated experience and is thus a prime candidate for the implementation of so-called Generation-IV nuclear energy systems. In this context, the safety demonstration of SFRs remains a major Research and Development related issue. The current research aims at the development of a computational tool for the in-depth understanding of SFR core behaviour during accidental transients, particularly those including boiling of the coolant. An accurate modelling of the core physics during such transients requires the coupling between 3D neutron kinetics and thermal-hydraulics in the core, to account for the strong interactions between the two-phase coolant flow and power variations caused by the sodium void effect. The present study is specifically focused upon models for the representation of sodium two-phase flow. The extension of the thermal-hydraulics TRACE code, previously limited to the simulation of single-phase sodium flow, has been carried out through the implementation of equations-of-state and closure relations specific to sodium. The different correlations have then been implemented as options. From the validation study carried out, it has been possible to recommend a set of models which provide satisfactory results, while considering annular flow as the dominant regime up to dryout and a smooth breakdown of the

  6. Passive safety optimization in liquid-sodium cooled reactors

    International Nuclear Information System (INIS)

    Cahalan, J. E.; Hahn, D.; Chang, W.-P.; Kwon, Y.-M.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2004-01-01

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4)

  7. Methods for the sodium cooled fast reactor fire safety provisions

    International Nuclear Information System (INIS)

    Gryaznov, B.V.; Dergachev, N.P.

    1983-01-01

    Problems of fire safety provision on NPPs with sodium cooled fast reactor are under discussion. Methods of sodium leak localization, measures eliminating sodium flaring up during leaks and main means of sodium fire extinguishing are considered. An extinguishing of sodium flaring up is performed by means of sodium temperatUre decrease and by limitation of hydrogen access to the flaring up surface. A conclusion is made that the most effective methods of extinguishing are the following: self-extinguishing (due to hydrogen burning out in a limiting volume); extinguishing by a gas mixture of nitrogen and carbonic acid (initial filling and blowing of rooms during sodium flaring up); extinguishing by special powders

  8. Design study on sodium cooled large-scale reactor

    International Nuclear Information System (INIS)

    Murakami, Tsutomu; Hishida, Masahiko; Kisohara, Naoyuki

    2004-07-01

    In Phase 1 of the 'Feasibility Studies on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled large-scale reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2, design improvement for further cost reduction of establishment of the plant concept has been performed. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2003, which is the third year of Phase 2. In the JFY2003 design study, critical subjects related to safety, structural integrity and thermal hydraulics which found in the last fiscal year has been examined and the plant concept has been modified. Furthermore, fundamental specifications of main systems and components have been set and economy has been evaluated. In addition, as the interim evaluation of the candidate concept of the FBR fuel cycle is to be conducted, cost effectiveness and achievability for the development goal were evaluated and the data of the three large-scale reactor candidate concepts were prepared. As a results of this study, the plant concept of the sodium-cooled large-scale reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

  9. Development Status on Innovative Sodium-Cooled Fast Reactor (JSFR)

    International Nuclear Information System (INIS)

    Yanagisawa, Tsutomu; Sato, Kazujiro

    2006-01-01

    The first step in Japan's nuclear fuel cycle policy is to introduce MOX recycle in light water reactors (LWRs) and the final step is to establish multiple TRU recycle in fast reactors (FRs), with the goal of realizing a stable supply, effective use of nuclear fuel resources, and the environmentally friendly production of energy. Therefore, a feasibility study on commercialized FR cycle systems has been launched since July 1999 by a Japanese joint project team of Japan Atomic Energy Agency (JAEA) and the Japan Atomic Power Company (JAPC: the representative of the electric utilities) in cooperation with Central Research Institute of Electric Power Industry (CRIEPI) and vendors. In the period from July 1999 to March 2001, the feasibility study phase-I was conducted to screen out representative FR cycle concepts. In the feasibility study phase-II (April 2001 - March 2006), investigations in to the representative FR concepts were carried out to clarify the most promising concept for commercial deployment. This paper describes an innovative sodium-cooled FR, which is named as the JAEA Sodium-cooled FR (JSFR), as the most promising FR concept that meets the Generation-IV performance target. The JSFR employs several advanced technologies, such as an oxide dispersion strengthened (ODS) cladding for higher burn-up, a short-piping configuration with less elbows by adopting high chromium steel, a large scale integrated intermediate heat exchanger with a primary circulation pump, etc. Based on the design, construction and operation experiences of JOYO and MONJU, there are extensive technology bases for sodium-cooled FRs. Nevertheless, several innovative technologies implemented into the JSFR have to be developed in order to realize higher economic competitiveness by reducing construction costs and improving plant availability

  10. Cooling Performance of ALIP according to the Air or Sodium Cooling Type

    Energy Technology Data Exchange (ETDEWEB)

    Ye, Huee-Youl; Yoon, Jung; Lee, Tae-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    ALIP pumps the liquid sodium by Lorentz force produced by the interaction of induced current in the liquid metal and their associated magnetic field. Even though the efficiency of the ALIP is very low compared to conventional mechanical pumps, it is very useful due to the absence of moving parts, low noise and vibration level, simplicity of flow rate regulation and maintenance, and high temperature operation capability. Problems in utilization of ALIP concern a countermeasure for elevation of internal temperature of the coil due to joule heating and how to increase magnetic flux density of Na channel gap. The conventional ALIP usually used cooling methods by circulating the air or water. On the other hand, GE-Toshiba developed a double stator pump adopting the sodium-immersed self-cooled type, and it recovered the heat loss in sodium. Therefore, the station load factor of the plant could be reduced. In this study, the cooling performance with cooling types of ALIP is analyzed. We developed thermal analysis models to evaluate the cooling performance of air or sodium cooling type of ALIP. The cooling performance is analyzed for operating parameters and evaluated with cooling type. 1-D and 3-D thermal analysis model for IHTS ALIP was developed, and the cooling performance was analyzed for air or sodium cooling type. The cooling performance for air cooling type was better than sodium cooling type at higher air velocity than 0.2 m/s. Also, the air temperature of below 270 .deg. demonstrated the better cooling performance as compared to sodium.

  11. Conceptual design of advanced central receiver power systems sodium-cooled receiver concept. Volume 2, Book 1. Commercial plant conceptual design. Final report

    Energy Technology Data Exchange (ETDEWEB)

    1979-03-01

    The conceptual design of the 100-MW solar tower focus commercial power plant is described in detail. Sodium is pumped up to the top of a tall tower where the receiver is located. The sodium is heated in the receiver and then flows down the tower, through a pressure reducing device, and thence into a large, hot storage tank which is located at ground level and whose size is made to meet a specific thermal energy storage capacity requirement. From this tank, the sodium is pumped by a separate pump, through a system of sodium-to-water steam generators. The steam generator system consists of a separate superheater and reheater operating in parallel and an evaporator unit operating in series with the other two units. The sodium flowing from the evaporator unit is piped to a cold storage tank. From the cold storage tank, sodium is then pumped up to the tip of the tower to complete the cycle. The steam generated in the steam generators is fed to a conventional off-the-shelf, high-efficiency turbine. The steam loop operates in a conventional rankine cycle with the steam generators serving the same purpose as a conventional boiler and water being fed to the evaporator with conventional feedwater pumps. The pressure reducing device (a standard drag valve, for example) serves to mitigate the pressure caused by the static head of sodium and thus allows the large tanks to operate at ambient pressure conditions. (WHK)

  12. Evolution of design of steam generator for sodium cooled reactors

    International Nuclear Information System (INIS)

    Chetal, S.C.; Vaidyanathan

    1997-01-01

    The first sodium cooled reactor was the experimental breeder reactor (EBR-I) in usa which was commissioned in 1951 and was incidentally the first nuclear reactor to generate electrical energy. This was followed by fast breeder reactors in USSR, UK, france, USA, japan, germany and India. The use of sodium as a coolant is due to its low moderation which helps in breeding fissile fuel from fertile materials and also its high heat transfer coefficient at comparatively low velocities. The good heat transfer properties introduce thermal stresses when there are rapid changes in the sodium temperatures. Also sodium has a chemical affinity with air and water. The steam generators for sodium cooled reactors have to allow for these novel conditions and in addition, unlike other components. Choices have to be made whether it is a recirculation type as in most fossil plants or an once through unit, the power rating, shape of the tube (straight, helical, U-tube), materials (Ferritic or austenitic), with free level of sodium or not, sodium on tube side or shell side and so on. With higher pressures and steam temperatures reheating steam after partial expansion in the turbine becomes essential as in conventional turbines. For this purpose the choice of reheating fluid viz sodium or live main steam has to be made. This paper traces the evolution of steam generator designs in the different sodium cooled reactors (chronologically) and the operation experience. 16 figs., 1 tab

  13. Simplified numerical simulation of hot channel in sodium cooled reactor

    International Nuclear Information System (INIS)

    Fonseca, F. de A.S. da; Silva Filho, E.

    1988-12-01

    The thermal-hydraulic parameter values that restrict the operation of a liquid sodium cooled reactor are not established by the average conditions of the coolant in the reactor core but by the extreme conditions of the hot channel. The present work was developed to analysis of hot channel of a sodium cooled reactor, adapting to this reactor an existent simplified model for hot channel of pressurized water reactor. The model was applied for a standard sodium reactor and the results are considered satisfatory. (author) [pt

  14. Proceedings: Cooling tower and advanced cooling systems conference

    International Nuclear Information System (INIS)

    1995-02-01

    This Cooling Tower and Advanced Cooling Systems Conference was held August 30 through September 1, 1994, in St. Petersburg, Florida. The conference was sponsored by the Electric Power Research Institute (EPRI) and hosted by Florida Power Corporation to bring together utility representatives, manufacturers, researchers, and consultants. Nineteen technical papers were presented in four sessions. These sessions were devoted to the following topics: cooling tower upgrades and retrofits, cooling tower performance, cooling tower fouling, and dry and hybrid systems. On the final day, panel discussions addressed current issues in cooling tower operation and maintenance as well as research and technology needs for power plant cooling. More than 100 people attended the conference. This report contains the technical papers presented at the conference. Of the 19 papers, five concern cooling tower upgrades and retrofits, five to cooling tower performance, four discuss cooling tower fouling, and five describe dry and hybrid cooling systems. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  15. Ultrasonic sweep arm for sodium cooled reactors

    International Nuclear Information System (INIS)

    Rohrbacher, H.A.; Bartholomay, R.

    1975-05-01

    This report describes experience in the use of a new type of monitoring and testing device to be applied in conjunction with components under sodium. In the method outlined, ultrasonic pulses are used which are emitted into the sodium plenum of fast breeder reactors by newly developed high temperature transducers. The basic work was conducted under out-of-pile conditions in a sodium tank of the sodium tank facility of the Karlsruhe Institute for Reactor Development. The sensor development, which preceded this phase, resulted in the use of soldered lithium niobate crystals whose operating characteristics were improved by the preliminary treatment outlined in the report. Special materials and techniques suitable for sensor fabrication are proposed. An alternative to soldering is suggested for contacting the crystals with their diaphragms, i.e. a contact pressure concept for the range of application up to 2 MHz. (orig.) [de

  16. Sodium pool fire analysis of sodium-cooled fast reactor by calculation

    International Nuclear Information System (INIS)

    Yu Hong; Xu Mi; Jin Degui

    2002-01-01

    Theoretical models were established according to the characteristic of sodium pool fire, and the SPOOL code was created independently. Some transient processes in sodium pool fire were modeled, including chemical reaction of sodium and oxygen; sodium combustion heat transfer modes in several kids of media; production, deposition and discharge of sodium aerosol; mass and energy exchange between different media in different ventilating conditions. The important characteristic parameters were calculated, such as pressure and temperature of gas, temperature of building materials, mass concentration of sodium aerosol, and so on. The SPOOL code, which provided available safety analysis tool for sodium pool fire accidents in sodium-cooled fast reactor, was well demonstrated with experimental data

  17. Analysis of the formation of local cooling disturbances in sodium-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1976-09-01

    The aim of this analysis of the formation of local cooling disturbances in sodium-cooled fast breeder reactors is to get results on the possible extent of blockages and the time necessary for growth which may be used for a safety evaluation. After an introduction where the thermohydraulic and physical/chemical aspects of the problems are considered, the causes for the local cooling disturbances and the phenomena arising with it are freated in more detail. (orig./TK) [de

  18. A reflux capsule steam generator for sodium cooled reactors

    International Nuclear Information System (INIS)

    Lantz, E.

    Pressurized water reactor plants at numerous sites have sustained significant leakage through their steam generators. The consequent shutdowns for repairs and replacements have damaged their economics. This experience suggests that if steam generators for liquid metal fast breeder reactors (LMFBR's) continue to be built as presently designed some of them will have similar problems. Because of their larger capital investment, the consequent damage to the economics of LMFBR's could be more serious. Reflux capsules provide a way to separate sodium from water and to reduce thermal stresses in steam generators for sodium cooled reactors. Their use would also eliminate the need for a primary heat exchanger and a secondary sodium loop pump. (author)

  19. Shape optimization of a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Schmitt, D.; Allaire, G.; Pantz, O.; Pozin, N.

    2013-01-01

    Traditional designs of sodium cooled fast reactors have a positive sodium expansion feedback. During a loss of flow transient without scram, sodium heating and boiling thus insert a positive reactivity and prevents the power from decreasing. Recent studies led at CEA, AREVA and EDF show that cores with complex geometries can feature a very low or even a negative sodium void worth. Usual optimization methods for core conception are based on a parametric description of a given core design. New core concepts and shapes can then only be found by hand. Shape optimization methods have proven very efficient in the conception of optimal structures under thermal or mechanical constraints. First studies show that these methods could be applied to sodium cooled core conception. In this paper, a shape optimization method is applied to the conception of a sodium cooled fast reactor core with low sodium void worth. An objective function to be minimized is defined. It includes the reactivity change induced by a 1% sodium density decrease. The optimization variable is a displacement field changing the core geometry from one shape to another. Additionally, a parametric optimization of the plutonium content distribution of the core is made, so as to ensure that the core is kept critical, and that the power shape is flat enough. The final shape obtained must then be adjusted to a given realistic core layout. Its characteristics can be checked with reference neutronic codes such as ERANOS. Thanks to this method, new shapes of reactor cores could be inferred, and lead to new design ideas. (authors)

  20. Parametric study of sodium aerosols in the cover-gas space of sodium-cooled reactors

    International Nuclear Information System (INIS)

    Sheth, A.

    1975-03-01

    A mathematical model has been developed to describe the behavior of sodium aerosols in the cover-gas space of a sodium-cooled reactor. A review of the literature was first made to examine methods of aerosol generation, mathematical expressions representing aerosol behavior, and pertinent experimental investigations of sodium aerosols. In the development of the model, some terms were derived from basic principles and other terms were estimated from available correlations. The model was simulated on a computer, and important parameters were studied to determine their effects on the overall behavior of sodium aerosols. The parameters studied were sodium pool temperature, source and initial size of particles, film thickness at the sodium pool/cover gas interface, wall plating parameters, cover-gas flow rate, and type of cover gas (argon and helium). The model satisfactorily describes the behavior of sodium aerosol in argon, but not in helium. Possible reasons are given for the failure of the model with helium, and further experimental work is recommended. The mathematical model, with appropriate modifications to describe the behavior of sodium aerosols in helium, would be very useful in designing traps to remove aerosols from the cover gas of sodium-cooled reactors. (U.S.)

  1. Design and selection of materials for sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Chetal, S.C.

    2011-01-01

    Sodium cooled fast reactors are currently in operation, under construction or under design by a number of countries. The design of sodium cooled fast reactor is covered by French RCC - MR code and ASME code NH. The codes cover rules as regards to materials, design and construction. These codes do not cover the effect of irradiation and environment. Elevated temperature design criteria in nuclear codes are much stringent in comparison to non nuclear codes. Sodium corrosion is not an issue in selection of materials provided oxygen impurity in sodium is controlled for which excellent reactor operating experience is available. Austenitic stainless steels have remained the choice for the permanent structures of primary sodium system. Stabilized austenitic stainless steel are rejected because of poor operating experience and non inclusion in the design codes. Route for improved creep behaviour lies in compositional modifications in 316 class steel. However, the weldability needs to be ensured. For cold leg component is non creep regime, SS 304 class steel is favoured from overall economics. Enhanced fuel burn up can be realized by the use of 9-12%Cr 1%Mo class steel for the wrapper of MOX fuel design, and cladding and wrapper for metal fuel reactors. Minor compositional modifications of 20% cold worked 15Cr-15Ni class austenitic stainless steel will be a strong candidate for the cladding of MOX fuel design in the short term. Long term objective for the cladding will be to develop oxide dispersion strengthened steel. 9%Cr 1%Mo class steel (Gr 91) is an ideal choice for integrated once through sodium heated steam generators. One needs to incorporate operating experience from reactors and thermal power stations, industrial capability and R and D feedback in preparing the technical specifications for procurement of wrought products and welding consumables to ensure reliable operation of the components and systems over the design life. The paper highlights the design approach

  2. Advanced sodium fast reactor accident source terms :

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-09-01

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

  3. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

    1992-09-01

    The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

  4. Conceptual design study on simplified and safer cooling systems for sodium cooled FBRs

    International Nuclear Information System (INIS)

    Hayafune, Hiroki; Shimakawa, Yoshio; Ishikawa, Hiroyasu; Kubota, Kenichi; Kobayashi, Jun; Kasai, Shigeo

    2000-06-01

    The objective of this study is to create the FBR plant concepts increasing economy and safety for the Phase-I 'Feasibility Studies on Commercialized Fast Reactor System'. In this study, various concepts of simplified 2ry cooling system for sodium cooled FBRs are considered and evaluated from the view points of technological feasibility, economy, and safety. The concepts in the study are considered on the basis of the following points of view. 1. To simplify 2ry cooling system by moderating and localizing the sodium-water reaction in the steam generator of the FBRs. 2. To simplify 2ry cooling system by eliminating the sodium-water reaction using integrated IHX-SG unit. 3. To simplify 2ry cooling system by eliminating the sodium-water reaction using a power generating system other than the steam generator. As the result of the study, 12 concepts and 3 innovative concepts are proposed. The evaluation study for those concepts shows the following technical prospects. 1. 2 concepts of integrated IHX-SG unit can eliminate the sodium-water reaction. Separated IHX and SG tubes unit using Lead-Bismuth as the heat transfer medium. Integrated IHX-SG unit using copper as the heat transfer medium. 2. Cost reduction effect by simplified 2ry cooling system using integrated IHX-SG unit is estimated 0 to 5%. 3. All of the integrated IHX-SG unit concepts have more weight and larger size than conventional steam generator unit. The weight of the unit during transporting and lifting would limit capacity of heat transfer system. These evaluation results will be compared with the results in JFY 2000 and used for the Phase-II study. (author)

  5. Core Seismic Tests for a Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Gyeong Hoi; Lee, J. H

    2007-01-15

    This report describes the results of the comparison of the core seismic responses between the test and the analysis for the reduced core mock-up of a sodium-cooled fast reactor to verify the FAMD (Fluid Added Mass and Damping) code and SAC-CORE (Seismic Analysis Code for CORE) code, which implement the application algorithm of a consistent fluid added mass matrix including the coupling terms. It was verified that the narrow fluid gaps between the duct assemblies significantly affect the dynamic characteristics of the core duct assemblies and it becomes stronger as a number of duct increases within a certain level. As conclusion, from the comparison of the results between the tests and the analyses, it is verified that the FAMD code and the SAC-CORE code can give an accurate prediction of a complex core seismic behavior of the sodium-cooled fast reactor.

  6. Safety Design Criteria of Indian Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Pillai, P.; Chellapandi, P.; Chetal, S.C.; Vasudeva Rao, P.R.

    2013-01-01

    • Important feedback has been gained through the design and safety review of PFBR. • The safety criteria document prepared by AERB and IGCAR would provide important input to prepare the dedicated document for the Sodium cooled Fast Reactors at the national and international level. • A common approach with regard to safety, among countries pursuing fast reactor program, is desirable. • Sharing knowledge and experimental facilities on collaborative basis. • Evolution of strong safety criteria – fundamental to assure safety

  7. Design of sodium cooled reactor systems and components for maintainability

    International Nuclear Information System (INIS)

    Carr, R.W.; Charnock, H.O.; McBride, J.P.

    1978-09-01

    Special maintenability problems associated with the design and operation of sodium cooled reactor plants are discussed. Some examples of both good and bad design practice are introduced from the design of the FFTF plant and other plants. Subjects include design for drainage, cleaning, decontamination, access, component removal, component disassembly and reassembly, remote tooling, jigs, fixtures, and design for minimizing radiation exposure of maintenance personnel. Check lists are included

  8. Apparatus for removing impurities in the sodium of sodium cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yamauchi, A

    1970-11-11

    An apparatus is provided for removing oxygen from liquid sodium flowing in a sodium cooled reactor. The removal of oxygen is complete with high efficiency. The liquid sodium to be purified is disposed outside a cylindrical wall and negatively charged, whereas sodium as a reducing material is disposed inside the same wall. The cylindrical wall is made of zirconia-calcia (ZrO/sub 2/)sub(0.87)(CaO)sub(0.13) solid electrolyte, the cylinder having a thickness of 2.5mm, a diameter of 3cm and a depth of 20cm under the sodium level. Electric resistance of the solid electrolyte is 2.3 ohm at 500/sup 0/C. A current of 1A by the application of 25 volts treats 0.3g of oxygen. Consequently, 1 liter or 1kg of liquid sodium containing 1,000ppm of oxygen can be purified for about 3 hours at an electrical consumption of 7.5 watt-hour. In one embodiment, a cylindrical electrolytic solid made of zirconia-calcia or zirconia-yttria was disposed in a container. Liquid sodium containing oxygen flowed outside of the cylinder. Liquid sodium as a reducing material was present inside the cylinder and the container and the cylinder were electrically insulated. An electrode was inserted at the center of the cylinder and a baffle plate at the upper portion of the electrode to shield heat and rising sodium vapor was provided. The space above the container was filled with an inert gas. The oxygen in the liquid sodium to be purified transferred through the wall of the cylinder into the interior of the cylinder so as to oxydize the reducing sodium material. The supersaturated sodium oxide inside the cylinder was deposited.

  9. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  10. Thermohydraulic relationships for advanced water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-04-01

    This report was prepared in the context of the IAEA's Co-ordinated Research Project (CRP) on Thermohydraulic Relationships for Advanced Water Cooled Reactors, which was started in 1995 with the overall goal of promoting information exchange and co-operation in establishing a consistent set of thermohydraulic relationships which are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. For advanced water cooled reactors, some key thermohydraulic phenomena are critical heat flux (CHF) and post CHF heat transfer, pressure drop under low flow and low pressure conditions, flow and heat transport by natural circulation, condensation of steam in the presence of non-condensables, thermal stratification and mixing in large pools, gravity driven reflooding, and potential flow instabilities. The objectives of the CRP are (1) to systematically list the requirements for thermohydraulic relationships in support of advanced water cooled reactors during normal and accident conditions, and provide details of their database where possible and (2) to recommend and document a consistent set of thermohydraulic relationships for selected thermohydraulic phenomena such as CHF and post-CHF heat transfer, pressure drop, and passive cooling for advanced water cooled reactors. Chapter 1 provides a brief discussion of the background for this CRP, the CRP objectives and lists the participating institutes. Chapter 2 provides a summary of important and relevant thermohydraulic phenomena for advanced water cooled reactors on the basis of previous work by the international community. Chapter 3 provides details of the database for critical heat flux, and recommends a prediction method which has been established through international co-operation and assessed within this CRP. Chapter 4 provides details of the database for film boiling heat transfer, and presents three methods for predicting film boiling heat transfer coefficients developed by institutes

  11. Design study on sodium-cooled middle-scale modular reactor

    International Nuclear Information System (INIS)

    Shimakawa, Yoshio; Nibe, Nobuaki; Hori, Toru

    2002-05-01

    In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled middle-scale modular reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled middle-scale modular reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled middle-scale modular reactor performed in JFY2001, which is the first year of Phase 2. As the construction cost of the sodium-cooled middle-scale modular reactor, which has been constructed in Phase 1, was about 10% higher than that of the sodium-cooled large-scale reactor, a new concept of the middle-scale modular reactor, which is expected to be equal to the large-scale reactor from a viewpoint of economic competitiveness, has been re-constructed based on the design of the advanced loop type reactor. After that, fundamental specifications of main systems and components for the new concept have been set, and critical subjects related to safety, structural integrity, thermal hydraulics, operability, maintainability and economy have been examined and evaluated. As a result of this study, the plant concept of the sodium-cooled middle-scale modular reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

  12. Mapping of sodium void worth and doppler effect for sodium-cooled fast reactor - 15458

    International Nuclear Information System (INIS)

    Krepel, J.; Pelloni, S.; Bortot, S.; Panadero, A.L.; Mikityuk, K.

    2015-01-01

    The sodium-cooled fast reactor (SFR) represents the reference and the most technologically mastered system among the Generation-IV reactors. Nevertheless, the sodium void worth in the fuel regions of SFR is usually positive. To overcome this safety drawback, low-void sodium-cooled fast spectrum core (CFV) was proposed by CEA. Such a CFV core is used in the frame of WP6 'Core safety' of the FP7 Euratom ESNII+ project as a reference SFR design. The overall sodium void effect is negative for the CFV core. Nevertheless, locally it is positive in the fuel region and negative in the sodium plenum. Similarly, also the Doppler effect is spatially dependent and it varies between the inner and outer fuel regions and between the middle and lower blankets. Accordingly, knowledge of the local distributions or actually mappings of the two safety-related parameters will be necessary, before safety assessment and transient analysis can be done. In this study these maps have been produced using the deterministic code ERANOS. The obtained mapping shows strong local dependency of both safety-related effects. A sensitivity of the void effect to the sodium plenum modeling was also demonstrated. The results may serve as an input for the transient analysis of the CFV core or as a cross-check for the Monte Carlo method based maps. (authors)

  13. Advanced gas-cooled reactors (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Yeomans, R. M. [South of Scotland Electricity Board, Hunterston Power Station, West Kilbride, Ayshire, UK

    1981-01-15

    The paper describes the advanced gas-cooled reactor system, Hunterston ''B'' power station, which is a development of the earlier natural uranium Magnox type reactor. Data of construction, capital cost, operating performance, reactor safety and also the list of future developments are given.

  14. Sodium flow measurement in large pipelines of sodium cooled fast breeder reactors with bypass type flow meters

    International Nuclear Information System (INIS)

    Rajan, K.K.; Jayakumar, T.; Aggarwal, P.K.; Vinod, V.

    2016-01-01

    Highlights: • Bypass type permanent magnet flow meters are more suitable for sodium flow measurement. • A higher sodium velocity through the PMFM sensor will increase its sensitivity and resolution. • By modifying the geometry of bypass line, higher sodium velocity through sensor is achieved. • With optimized geometry the sensitivity of bypass flow meter system was increased by 70%. - Abstract: Liquid sodium flow through the pipelines of sodium cooled fast breeder reactor circuits are measured using electromagnetic flow meters. Bypass type flow meter with a permanent magnet flow meter as sensor in the bypass line is selected for the flow measurement in the 800 NB main secondary pipe line of 500 MWe Prototype Fast Breeder Reactor (PFBR), which is at the advanced stage of construction at Kalpakkam. For increasing the sensitivity of bypass flow meters in future SFRs, alternative bypass geometry was considered. The performance enhancement of the proposed geometry was evaluated by experimental and numerical methods using scaled down models. From the studies it is observed that the new configuration increases the sensitivity of bypass flow meter system by around 70%. Using experimentally validated numerical tools the volumetric flow ratio for the bypass configurations is established for the operating range of Reynolds numbers.

  15. In service inspection and repair of sodium cooled ASTRID prototype

    Energy Technology Data Exchange (ETDEWEB)

    Baque, F.; Jadot, F. [French Atomic Commission, Cadarache Centre, 13108 Saint Paul lez Durance Cedex, (France); Marlier, R. [AREVA, 10 rue Recamier, 69456 Lyon cedex 06, (France); Saillant, J-F. [AREVA/NDE Solutions, 4 rue Thomas Dumorey, BP 70385, 71109 Chalon sur Saone Cedex, (France); Delalande, V. [EDF R and D, 6, quai Watier, 78400 Chatou, (France)

    2015-07-01

    In the frame of the large R and D work which is performed for the future ASTRID sodium cooled prototype, In Service Inspection and Repair (ISI and R) has been identified as a major issue to be taken into account in order to enlarge the plant safety, to consolidate its availability and to protect the associated investment. After the first part of pre-conceptual design phase (2008-2012), the running second part of pre-conceptual phase (2013-2015) allows to increase the ISI and R tool ability for immersed sodium structures of ASTRID, at about 200 deg. C, on the basis of consolidated specifications and thanks to their qualification through more and more realistic laboratory tests and simulation with CIVA code. ISI and R items are being developed and qualified during a pluri-annual program which mainly deals with the reactor block structures, the primary components and circuit, and the Power Conversion System. It ensures a strong connection between the reactor designers and inspection specialists, as the optimization of inspectability and repairability is looked at: this already induced specific rules for design, in order to shorten and ease the ISI and R operations, which have been merged into RCC-MRx rules. In the frame of increasing technology readiness level with corresponding performance demonstration, this paper presents R and D dealing with the ISI and R items: it highlights the sensor development (both ultrasonic and electromagnetic concepts, compatible with sodium at 200 deg. C), then their applications for ASTRID structure control (under sodium telemetry, imaging and NDE). Activity for repair is also presented (a single laser tool for sodium sweeping, machining and welding), and finally the effort for associated robotic (generic program for ASTRID applications, specific technological tools for sodium medium, tight immersed bell). The main results of testing and simulation are given for telemetry, vision, NDE applications, laser process repair and under sodium

  16. ASTRID, Generation IV advanced sodium technological reactor for industrial demonstration

    International Nuclear Information System (INIS)

    Gauche, F.

    2013-01-01

    ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) is an integrated technology demonstrator designed to demonstrate the operability of the innovative choices enabling fast neutron reactor technology to meet the Generation IV criteria. ASTRID is a sodium-cooled fast reactor with an electricity generating power of 600 MWe. In order to meet the generation IV goals, ASTRID will incorporate the following decisive innovations: -) an improved core with a very low, even negative void coefficient; -) the possible installation of additional safety devices in the core. For example, passive anti-reactivity insertion devices are explored; -) more core instrumentation; -) an energy conversion system with modular steam generators, to limit the effects of a possible sodium-water reaction, or sodium-nitrogen exchangers; -) considerable thermal inertia combined with natural convection to deal with decay heat; -)elimination of major sodium fires by bunkerization and/or inert atmosphere in the premises; -) to take into account off-site hazards (earthquake, airplane crash,...) right from the design stage; -) a complete rethink of the reactor architecture in order to limit the risk of proliferation. ASTRID will also include systems for reducing the length of refueling outages and increasing the burn-up and the duration of the cycle. In-service inspection, maintenance and repair are also taken into account right from the start of the project. The ASTRID prototype should be operational by about 2023. (A.C.)

  17. Delayed gamma power measurement for sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

    2011-01-15

    Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  18. Under-Sodium-Viewing as one technique for periodic inspections in sodium-cooled fast reactors-- possibilities and limits

    International Nuclear Information System (INIS)

    Weiss, H.

    1979-07-01

    Periodic inspections are gaining increasingly technical importance for fast sodium cooled reactors. Among others the reactor tank and its internals have to be inspected, whereby licensing experts partly are requesting the standards of Light Water Reactors. This leads to difficulties in sodium cooled reactors because of the non-transparent coolant sodium and their compact structure. In order to avoid the complete dumping of the sodium, the under sodium viewing shall be applied besides other inspection methods. Since this is a new method, which is still in its development phase, this report presents and discusses the technical and physical basis and outlines possibilities and limits [de

  19. An Assessment of Fission Product Scrubbing in Sodium Pools Following a Core Damage Event in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bucknor, M.; Farmer, M.; Grabaskas, D.

    2017-06-26

    The U.S. Nuclear Regulatory Commission has stated that mechanistic source term (MST) calculations are expected to be required as part of the advanced reactor licensing process. A recent study by Argonne National Laboratory has concluded that fission product scrubbing in sodium pools is an important aspect of an MST calculation for a sodium-cooled fast reactor (SFR). To model the phenomena associated with sodium pool scrubbing, a computational tool, developed as part of the Integral Fast Reactor (IFR) program, was utilized in an MST trial calculation. This tool was developed by applying classical theories of aerosol scrubbing to the decontamination of gases produced as a result of postulated fuel pin failures during an SFR accident scenario. The model currently considers aerosol capture by Brownian diffusion, inertial deposition, and gravitational sedimentation. The effects of sodium vapour condensation on aerosol scrubbing are also treated. This paper provides details of the individual scrubbing mechanisms utilized in the IFR code as well as results from a trial mechanistic source term assessment led by Argonne National Laboratory in 2016.

  20. Operating Temperatures of a Sodium-Cooled Exhaust Valve as Measured by a Thermocouple

    Science.gov (United States)

    Sanders, J. C.; Wilsted, H. D.; Mulcahy, B. A.

    1943-01-01

    A thermocouple was installed in the crown of a sodium-cooled exhaust valve. The valve was then tested in an air-cooled engine cylinder and valve temperatures under various engine operating conditions were determined. A temperature of 1337 F was observed at a fuel-air ratio of 0.064, a brake mean effective pressure of 179 pounds per square inch, and an engine speed of 2000 rpm. Fuel-air ratio was found to have a large influence on valve temperature, but cooling-air pressure and variation in spark advance had little effect. An increase in engine power by change of speed or mean effective pressure increased the valve temperature. It was found that the temperature of the rear spark-plug bushing was not a satisfactory indication of the temperature of the exhaust valve.

  1. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  2. Selection of steam generator materials for sodium cooled fast breeders

    International Nuclear Information System (INIS)

    Berge, P.

    1977-01-01

    The sodium water heat exchangers are now considered as the stumbling block in the development of liquid metal cooled fast breeders, due to the risk of sodium-water reactions. The selection of the materials for these tube-bundles has been very broad, for the different existing, or in-project, reactors in the world: low alloy 2 1/4 Cr - 1 Mo steels (unstabilized or stabilized); 9 Cr - 1 Mo ferritic steel; 18 Cr - 10 Ni austenitic stainless steels; alloy 800. On can also add other ferritic steels, as 9 Cr - 2 Mo stabilized, which are studied for this application. In the framework of the E.D.F.-C.E.A. working group a major effort was undertaken to study the characteristics of these various materials with respect to the main criteria governing construction of the tube bundles and their performance in service: mechanical characteristics at high temperature; fabrication and welding; behavior with respect to mass transfer in sodium; carburization and decarburization; corrosion resistance. The main lines and results of this program are described [fr

  3. Design study on sodium-cooled middle-scale modular reactor

    International Nuclear Information System (INIS)

    Kisohara, Naoyuki; Hishida, Masahiko; Nibe, Nobuaki

    2003-09-01

    In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled middle-scale modular reactor, which has a possibility to fulfill the design requirements of the F/S. This report summarizes the results of the design study on the sodium-cooled middle-scale modular reactor performed in JFY2002, which is the second year of Phase 2. The construction cost of the sodium-cooled middle-scale modular reactor, which has been constructed in JFY2002, was almost achieved the economical goal. But its achievability was not sufficient to accept the concept. In order to reduce the construction cost, the plant concept has been re-constructed based on the 50 MWe plant studied in JFY2002. After that, fundamental specifications of main systems and components for the new concept have been set, and critical subjects have been examined and evaluated. In addition, in order to achieve the further cost reduction, the plant with simplified secondary system, the plant with electric magnetic pump in secondary system, and the fuel handling system are examined and evaluated. As a result of this study, the plant concept of the sodium-cooled middle-scale modular reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000 yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

  4. 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

  5. Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sieger, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Moe, Wayne [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); HolbrookINL, Mark [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

  6. Control of radioactive material transport in sodium-cooled reactors

    International Nuclear Information System (INIS)

    Brehm, W.F.

    1980-03-01

    The Radioactivity Control Technology (RCT) program was established by the Department of Energy to develop and demonstrate methods to control radionuclide transport to ex-core regions of sodium-cooled reactors. This radioactive material is contained within the reactor heat transport system with any release to the environment well below limits established by regulations. However, maintenance, repair, decontamination, and disposal operations potentially expose plant workers to radiation fields arising from radionuclides transported to primary system components. This paper deals with radioactive material generated and transported during steady-state operation, which remains after 24 Na decay. Potential release of radioactivity during postulated accident conditions is not discussed. The control methods for radionuclide transport, with emphasis on new information obtained since the last Environmental Control Symposium, are described. Development of control methods is an achievable goal

  7. Control rod homogenization in heterogeneous sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Andersson, Mikael

    2016-01-01

    The sodium-cooled fast reactor is one of the candidates for a sustainable nuclear reactor system. In particular, the French ASTRID project employs an axially heterogeneous design, proposed in the so-called CFV (low sodium effect) core, to enhance the inherent safety features of the reactor. This thesis focuses on the accurate modeling of the control rods, through the homogenization method. The control rods in a sodium-cooled fast reactor are used for reactivity compensation during the cycle, power shaping, and to shutdown the reactor. In previous control rod homogenization procedures, only a radial description of the geometry was implemented, hence the axially heterogeneous features of the CFV core could not be taken into account. This thesis investigates the different axial variations the control rod experiences in a CFV core, to determine the impact that these axial environments have on the control rod modeling. The methodology used in this work is based on previous homogenization procedures, the so-called equivalence procedure. The procedure was newly implemented in the PARIS code system in order to be able to use 3D geometries, and thereby be take axial effects into account. The thesis is divided into three parts. The first part investigates the impact of different neutron spectra on the homogeneous control-rod cross sections. The second part investigates the cases where the traditional radial control-rod homogenization procedure is no longer applicable in the CFV core, which was found to be 5-10 cm away from any material interface. In the third part, based on the results from the second part, a 3D model of the control rod is used to calculate homogenized control-rod cross sections. In a full core model, a study is made to investigate the impact these axial effects have on control rod-related core parameters, such as the control rod worth, the capture rates in the control rod, and the power in the adjacent fuel assemblies. All results were compared to a Monte

  8. 4. generation sodium-cooled fast reactors. The ASTRID technological demonstrator

    International Nuclear Information System (INIS)

    2012-12-01

    The sodium-cooled fast reactor (SFR) concept is one of the four fast neutron concepts selected by the Generation IV International Forum (GIF). SFRs have favourable technical characteristics and they are the sole type of reactor for which significant industrial experience feedback is available. After a discussion of the past experience gained on fast breeder reactors in the world (benefits, difficulties and problematics), the authors discuss the main improvement domains and the associated R and D advances (reactor safety, prevention and mitigation of severe accidents, the sodium-water risk, detection of sodium leaks, increased availability, instrumentation and inspection, control and repairability, assembly handling and washing). Then, they describe the technical requirements and safety objectives of the ASTRID experimental project, notably with its reactivity management, cooling management, and radiological containment management functions. They describe and discuss requirements to be met and choices made for Astrid, and the design options for its various components (core and fuels, nuclear heater, energy conversion system, fuel assembly handling, instrumentation and in-service inspection, control and command). They present the installations which are associated with the ASTRID cycle, evoke the development and use of simulations and codes, describe the industrial organization and the international collaboration about the ASTRID project, present the planning and cost definition

  9. Towards the Characterization of the Bubble Presence in Liquid Sodium of Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Cavaro, M.; Jeannot, J.P.; Payan, C.

    2013-06-01

    In a Sodium cooled Fast Reactors (SFR), different phenomena such as gas entrainment or nucleation can lead to gaseous micro-bubbles presence in the liquid sodium of the primary vessel. Although this free gas presence has no direct impact on the core neutronics, the French Atomic Energy and Alternative Energies Commission (CEA) currently works on its characterization to, among others, check the absence of risk of large gas pocket formation and to assess the induced modifications of the sodium acoustic properties. The main objective is to evaluate the void fraction values (volume fraction of free gas) and the radii histogram of the bubbles present in liquid sodium. Acoustics and electromagnetic techniques are currently developed at CEA: - The low-frequency speed of sound measurement, which allows us to link - thanks to Wood's model - the measured speed of sound to the actual void fraction. - The nonlinear mixing of two frequencies, based on the nonlinear resonance behavior of a bubble. This technique allows knowing the radius histogram associated to a bubble cloud. Two different mixing techniques are presented in this paper: the mixing of two high frequencies and the mixing of a high and a low frequency. - The Eddy-current flowmeter (ECFM), the output signal of which is perturbed by free gas presence and in consequence allows detecting bubbles. For each technique, initial results are presented. Some of them are really promising. So far, acoustic experiments have been led with an air-water experimental set-up. Micro-bubbles clouds are generated with a dissolved air flotation device and monitored by an optical device which provides reference measurements. Generated bubbles have radii range from few micrometers to several tens of micrometers. Present and future air/water experiments are presented. Furthermore, a development plan of in-sodium tests is presented in terms of a device set-up, instrumentation, modeling tools and experiments. (authors)

  10. Neutron noise analysis for malfunction diagnosis at sodium cooled reactors

    International Nuclear Information System (INIS)

    Hoppe, P.

    1978-09-01

    For the investigation of the potential use of neutron noise analysis at sodium cooled power reactors, measurements have been performed at the KNK I reactor over a period of 18 month under different operational conditions. The signal fluctuations of the following tranducers have been recorded: In-core and Ex-core neutron detectors, temperature-, flow-, pressure-, vibration- and acoustic sensors. These extensive measurements have been analyzed in the frequency range from 0,001 Hz to 1000 Hz with all currently known methods for the identification of noise sources. The following results have been found: - Neutron noise for f 20 Hz the white detection noise prevails. In the region from 1 Hz to 20 Hz the vibrations of core components contribute to neutron noise. - Neutron noise is influenced by the state of the plant. - The contributions to neutron noise due to the fluctuations of coolant flow and inlet temperature are small compared to those produced by the movements of the control rod initiated by the reactor control system. The quantitatively unidentifiable amount of reactivity fluctuations (0,6 time-dependent thermal bowing of the core. With respect to these results and by calculation of the neutron noise patterns to be expected for the SNR 300, the following possible applications for neutron noise analysis have been found: By means of neutron noise analysis only reactivity fluctuations can be identified and supervised which are produced by time dependent changes of the core geometry. Furthermore neutron noise analysis is well suited for a sensitive detection of control rod vibrations and of local sodium boiling. Finally it can be used for the surveillance of the proper functioning of the reactor control system and of the control rod drive mechanism. (orig./HP) 891 HP [de

  11. Thermal analysis experiment for elucidating sodium-water chemical reaction mechanism in steam generator of sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kikuchi, Shin; Kurihara, Akikazu; Ohshima, Hiroyuki

    2012-01-01

    For the purpose of elucidating the mechanism of the sodium-water surface reaction in a steam generator of sodium-cooled fast reactors, kinetic study of the sodium (Na)-sodium hydroxide (NaOH) reaction has been carried out by using Differential Thermal Analysis (DTA) technique. The parameters, including melting points of Na and NaOH, phase transition temperature of NaOH, Na-NaOH reaction temperature, and decomposition temperature of sodium hydride (NaH) have been identified from DTA curves. Based on the measured reaction temperature, rate constant of sodium monoxide (Na 2 O) generation was obtained. Thermal analysis results indicated that Na 2 O generation at the secondary overall reaction should be considered during the sodium-water reaction. (author)

  12. Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors

    International Nuclear Information System (INIS)

    Vega R, A. K.; Espinosa P, G.; Gomez T, A. M.

    2016-09-01

    The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

  13. Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ferroni, Paolo [Westinghouse Electric Company LLC, Cranberry Township, PA (United States). Global Technology Development; Tatli, Emre [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Czerniak, Luke [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Chien, Hual-Te [Argonne National Lab. (ANL), Argonne, IL (United States); Yoichi, Momozaki [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-06-29

    The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na2O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO2 (sCO2) Brayton cycle power conversion system. In SFRs, Na2O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na2O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na2O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na2O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO2 systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test

  14. Construction within cooling system of a sodium cooler reactor

    International Nuclear Information System (INIS)

    1977-01-01

    A procedure is described for the manufacture and the construction of a bundle of a large number of pipes, at least near their outer ends lying practically evenly spaced which pipes lie with one of their outermost ends in a pipe plate and with their other outer ends in a second pipe plate, where the procedure involves placing at or near the derived place a means for holding the bundle of pipes, as well as eventually holding a pipe plate with stub pipes near the outer ends of the bundle of pipes, the successive attachment by means of welding of the pipes in the plate of the above mentioned assembly with the stub pipes, characterized in that to each of the pipes in the bundle is welded to an outer end directly a corresponding short pipe which is also welded to a pipe end of a stub pipe, so that a connection is made by the short pipe which lies between the outer end of the pipe in the bundle and the stub pipe. Such a construction is used in the heat exchanger of sodium cooled reactors. (G.C.)

  15. FY 2017-Influence of Sodium Environment on the Tensile Properties of Advanced Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Li, Meimei [Argonne National Lab. (ANL), Argonne, IL (United States); Chen, Wei-Ying [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-08-01

    This report provides an update on the understanding of the effects of sodium exposures on tensile properties of advanced alloy 709 in support of the design and operation of structural components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY17 (M3AT-17AN1602093), under the Work Package AT-17AN160209, “Sodium Compatibility” performed by Argonne National Laboratory (ANL), as part of Advanced Reactor Technologies Program. Three laboratory-size heats of Alloy 709 austenitic steel were investigated in liquid sodium environments at 550-650°C to understand its corrosion behaviour, microstructural evolution, and tensile properties. In addition, a commercial scale heat has been produced and hot-rolled into plates.

  16. Development of electro-magnetic pump for the ASTRID Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Suzuki, Tetsu; Aizawa, Rie; Wakasaki, Shingo; Dechelette, Frank; Benoit, Fabrice

    2017-01-01

    In the framework of the SFR (Sodium-cooled Fast Reactor) prototype called ASTRID (Advance Sodium Technological Reactor for Industrial Demonstration), the large capacity Electro-Magnetic Pumps (EMP) as main circulating pumps on the intermediate sodium circuits has been considered instead of mechanical pumps by CEA. The use of EMP has several decisive technological merits compared with mechanical pump in the reactor design, operation and maintenance. Nevertheless, some theoretical and technological developments have to be carried out in order to validate the design tools which take Magneto Hydro Dynamic (MHD) phenomena into account and the applicability of the EMP to the steady state and transient operating conditions of ASTRID. To move forward to developments, a collaboration agreement between the CEA and TOSHIBA Corporation was made and entered into to carry out a joint work program on the EMP for ASTRID design and development. CEA performed the theoretical analysis, and the EMP experimental model is constructed by CEA to support these theoretical developments. This model consists of a middle-size annular EMP for the liquid metal sodium. The various testing program using this model has been started in 2016. TOSHIBA performed the examination of design specification for ASTRID, an electromagnetic design, a structural design and various analyses. The structure design has been examined the placement of the sodium boundary and the withstand pressure, etc. And, if the thicknesses of the structure increase for withstanding pressure, the pump efficiency falls because the loss of the electromagnetic force increases. Therefore the balance between withstanding pressure and the efficiency has been considered by an electromagnetism design. This paper presents the design studies and experimental activities for the EMP development in the framework of the CEA-TOSHIBA collaborations. (author)

  17. Advanced gas cooled reactors - Designing for safety

    International Nuclear Information System (INIS)

    Keen, Barry A.

    1990-01-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme

  18. Advanced gas cooled reactors - Designing for safety

    Energy Technology Data Exchange (ETDEWEB)

    Keen, Barry A [Engineering Development Unit, NNC Limited, Booths Hall, Knutsford, Cheshire (United Kingdom)

    1990-07-01

    The Advanced Gas-Cooled Reactor Power Stations recently completed at Heysham in Lancashire, England, and Torness in East Lothian, Scotland represent the current stage of development of the commercial AGR. Each power station has two reactor turbo-generator units designed for a total station output of 2x660 MW(e) gross although powers in excess of this have been achieved and it is currently intended to uprate this as far as possible. The design of both stations has been based on the successful operating AGRs at Hinkley Point and Hunterston which have now been in-service for almost 15 years, although minor changes were made to meet new safety requirements and to make improvements suggested by operating experience. The construction of these new AGRs has been to programme and within budget. Full commercial load for the first reactor at Torness was achieved in August 1988 with the other three reactors following over the subsequent 15 months. This paper summarises the safety principles and guidelines for the design of the reactors and discusses how some of the main features of the safety case meet these safety requirements. The paper also summarises the design problems which arose during the construction period and explains how these problems were solved with the minimum delay to programme.

  19. Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

    Science.gov (United States)

    Natesan, K.; Li, Meimei; Chopra, O. K.; Majumdar, S.

    2009-07-01

    Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.

  20. Study of guided wave transmission through complex junction in sodium cooled reactor

    International Nuclear Information System (INIS)

    Elie, Q.; Le Bourdais, F.; Jezzine, K.; Baronian, V.

    2015-01-01

    Ultrasonic guided wave techniques are seen as suitable candidates for the inspection of welded structures within sodium cooled fast reactors (SFR), as the long range propagation of guided waves without amplitude attenuation can overcome the accessibility problem due to the liquid sodium. In the context of the development of the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID), the French Atomic Commission (CEA) investigates non-destructive testing techniques based on guided wave propagation. In this work, guided wave NDT methods are applied to control the integrity of welds located in a junction-type structure welded to the main vessel. The method presented in this paper is based on the analysis of scattering matrices peculiar to each expected defect, and takes advantage of the multi-modal and dispersive characteristics of guided wave generation. In a simulation study, an algorithm developed using the CIVA software is presented. It permits selecting appropriate incident modes to optimize detection and identification of expected flawed configurations. In the second part of this paper, experimental results corresponding to a first validation step of the simulation results are presented. The goal of the experiments is to estimate the effectiveness of the incident mode selection in plates. The results show good agreement between experience and simulation. (authors)

  1. Materials Options of Steam Generator for Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Fu Xiaogang; Long Bin; Han Liqing; Qin Bo; Zhang Jinquan; Wang Shuxing

    2013-01-01

    Overview of the material options of steam generator for sodium-cooled fast reactors, the method to calculate the service life, the thinning of wall thickness and the sodium corrosion rate, the degradation of mechanical properties (thermal aging and decarburization) and the calculation results of theoretical models

  2. Final report-passive safety optimization in liquid sodium-cooled reactors

    International Nuclear Information System (INIS)

    Cahalana, J. E.; Hahn, D.

    2007-01-01

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  3. Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

    International Nuclear Information System (INIS)

    Song, Wei; Wu, Yuanyu; Hu, Wenjun; Zuo, Jiaxu

    2015-01-01

    Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

  4. Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

    Energy Technology Data Exchange (ETDEWEB)

    Song, Wei [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China); Wu, Yuanyu [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint-Paul-lès-Durance (France); Hu, Wenjun [China Institute of Atomic Energy, P. O. Box 275(34), Beijing (China); Zuo, Jiaxu, E-mail: zuojiaxu@chinansc.cn [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China)

    2015-11-15

    Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

  5. Experience in handling core subassemblies in sodium cooled reactor KNK and test rigs

    International Nuclear Information System (INIS)

    Althaus; Jansing; Kesseler; Kirchner; Menck

    1974-01-01

    Compared with a water cooled reactor plant a sodium cooled reactor plant presents a number of problems which result from the specific nature of sodium. These problems that must be faced during all handling operations are mainly: 1. The rapid reaction of sodium in air requires handling to be done only under cover gas. 2. The temperature of all sodium-wetted components is to be kept above the melting point of sodium. 3. Poor draining of removed reactor components due to the high surface tension of sodium and the associated danger of dripping radioactive sodium may produce radiation or contamination problems. 4. Sodium is not transparent. The sum of these and further influences dictate that the general handling usually is carried out without visual means, though a method is under development in the USA to use ultrasonic for under sodium 'viewing'. These limitations to sodium component handling are applicable to all sodium reactor plants, several of which are discussed in this report. After the description of the handling systems of the KNK plant now operating at Karlsruhe, the experience with the SNR test rig and finally the handling systems for SNR 300 and SNR 2 are discussed

  6. Sustained Recycle in Light Water and Sodium-Cooled Reactors

    International Nuclear Information System (INIS)

    Piet, Steven J.; Bays, Samuel E.; Pope, Michael A.; Youinou, Gilles J.

    2010-01-01

    From a physics standpoint, it is feasible to sustain recycle of used fuel in either thermal or fast reactors. This paper examines multi-recycle potential performance by considering three recycling approaches and calculating several fuel cycle parameters, including heat, gamma, and neutron emission of fresh fuel; radiotoxicity of waste; and uranium utilization. The first recycle approach is homogeneous mixed oxide (MOX) fuel assemblies in a light water reactor (LWR). The transuranic portion of the MOX was varied among Pu, NpPu, NpPuAm, or all-TRU. (All-TRU means all isotopes through Cf-252.) The Pu case was allowed to go to 10% Pu in fresh fuel, but when the minor actinides were included, the transuranic enrichment was kept below 8% to satisfy the expected void reactivity constraint. The uranium portion of the MOX was enriched uranium. That enrichment was increased (to as much as 6.5%) to keep the fuel critical for a typical LWR irradiation. The second approach uses heterogeneous inert matrix fuel (IMF) assemblies in an LWR - a mix of IMF and traditional UOX pins. The uranium-free IMF fuel pins were Pu, NpPu, NpPuAm, or all-TRU. The UOX pins were limited to 4.95% U-235 enrichment. The number of IMF pins was set so that the amount of TRU in discharged fuel from recycle N (from both IMF and UOX pins) was made into the new IMF pins for recycle N+1. Up to 60 of the 264 pins in a fuel assembly were IMF. The assembly-average TRU content was 1-6%. The third approach uses fast reactor oxide fuel in a sodium-cooled fast reactor with transuranic conversion ratio of 0.50 and 1.00. The transuranic conversion ratio is the production of transuranics divided by destruction of transuranics. The FR at CR=0.50 is similar to the CR for the MOX case. The fast reactor cases had a transuranic content of 33-38%, higher than IMF or MOX.

  7. Recent advances in cooled-semen technology.

    Science.gov (United States)

    Aurich, Christine

    2008-09-01

    The majority of horse registries approve the use of artificial insemination, and horse breeding has widely taken benefit from the use of cooled-stored semen. New insights into cooled-semen technology open possibilities to reduce problems such as impaired semen quality after cooled-storage in individual stallions. The stallion itself has major impacts on quality and fertility of cooled-stored semen. Dietary supplementation of antioxidants and polyunsaturated fatty acids improves semen quality in a variety of species, but only few studies on this topic exist in the horse. Proper semen collection and handling is the main key to the maintenance of semen quality during cooled-storage. Semen collection should be achieved by minimal sexual stimulation with a single mount; this results in high sperm concentration, low content of seminal plasma and minimal contamination with bacteria. Milk-based semen extenders are most popular for semen processing and storage. The development of more defined extenders containing only the beneficial milk ingredients has made extender quality more constant and reliable. Semen is often centrifuged to decrease the seminal plasma content. Centrifugation results in a recovery rate of only 75% of spermatozoa in the semen pellet. Recovery rates after centrifugation may be improved with use of a "cushion technique" allowing higher centrifugation force and duration. However, this is not routinely used in cooled-semen technology. After slow-cooling, semen-storage and shipping is best performed at 5 degrees C, maintaining semen motility, membrane integrity and DNA integrity for up to 40 h after collection. Shipping containers created from Styrofoam boxes provide maintenance of semen quality at low cost.

  8. Corrosion performance of advanced structural materials in sodium.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Momozaki, Y.; Li, M.; Rink, D.L. (Nuclear Engineering Division)

    2012-05-16

    This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory, the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.' Two reports were issued in 2009 (Natesan and Meimei Li 2009, Natesan et al. 2009) which examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design specifications for the ANL pumped loop for testing advanced structural materials. Available information was presented on solubility of several metallic and nonmetallic elements along with a discussion of the possible mechanisms for the accumulation of impurities in sodium. That report concluded that the solubility of many metals in sodium is low (<1 part per million) in the temperature range of interest in sodium reactors and such trace amounts would not impact the mechanical integrity of structural materials and components. The earlier report also analyzed the solubility and transport mechanisms of nonmetallic elements such as oxygen, nitrogen, carbon, and hydrogen in laboratory sodium loops and in reactor systems such as Experimental Breeder Reactor-II, Fast Flux

  9. Corrosion performance of advanced structural materials in sodium

    International Nuclear Information System (INIS)

    Natesan, K.; Momozaki, Y.; Li, M.; Rink, D.L.

    2012-01-01

    This report gives a description of the activities in design, fabrication, construction, and assembling of a pumped sodium loop for the sodium compatibility studies on advanced structural materials. The work is the Argonne National Laboratory (ANL) portion of the effort on the work project entitled, 'Sodium Compatibility of Advanced Fast Reactor Materials,' and is a part of Advanced Materials Development within the Reactor Campaign. The objective of this project is to develop information on sodium corrosion compatibility of advanced materials being considered for sodium reactor applications. This report gives the status of the sodium pumped loop at Argonne National Laboratory, the specimen details, and the technical approach to evaluate the sodium compatibility of advanced structural alloys. This report is a deliverable from ANL in FY2010 (M2GAN10SF050302) under the work package G-AN10SF0503 'Sodium Compatibility of Advanced Fast Reactor Materials.' Two reports were issued in 2009 (Natesan and Meimei Li 2009, Natesan et al. 2009) which examined the thermodynamic and kinetic factors involved in the purity of liquid sodium coolant for sodium reactor applications as well as the design specifications for the ANL pumped loop for testing advanced structural materials. Available information was presented on solubility of several metallic and nonmetallic elements along with a discussion of the possible mechanisms for the accumulation of impurities in sodium. That report concluded that the solubility of many metals in sodium is low (<1 part per million) in the temperature range of interest in sodium reactors and such trace amounts would not impact the mechanical integrity of structural materials and components. The earlier report also analyzed the solubility and transport mechanisms of nonmetallic elements such as oxygen, nitrogen, carbon, and hydrogen in laboratory sodium loops and in reactor systems such as Experimental Breeder Reactor-II, Fast Flux Test Facility, and

  10. Monte Carlo transport correction of sodium reactivity worth spatial distribution in perspective Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Raskach, K.F.; Blyskavka, V; Kislitsyna, T.S.

    2011-01-01

    In this paper we apply Monte Carlo for calculating spatial distribution of sodium reactivity worth in the perspective Russian sodium-cooled fast reactor BN-1200. A special Monte Carlo technique applicable for calculating perturbations and derivatives of the effective multiplication factor is used. The numerical results obtained show that Monte Carlo has a good perspective to deal with such problems and to be used as a reference solution for engineering codes based on the diffusion approximation. They also allow to conclude that in the sodium blanket and in the neighboring region of the core the diffusion code used likely overestimates sodium reactivity worth. This conclusion has to be verified in future work. (author)

  11. Optimized evaporative cooling for sodium Bose-Einstein condensation against three-body loss

    International Nuclear Information System (INIS)

    Shobu, Takahiko; Yamaoka, Hironobu; Imai, Hiromitsu; Morinaga, Atsuo; Yamashita, Makoto

    2011-01-01

    We report on a highly efficient evaporative cooling optimized experimentally. We successfully created sodium Bose-Einstein condensates with 6.4x10 7 atoms starting from 6.6x10 9 thermal atoms trapped in a magnetic trap by employing a fast linear sweep of radio frequency at the final stage of evaporative cooling so as to overcome the serious three-body losses. The experimental results such as the cooling trajectory and the condensate growth quantitatively agree with the numerical simulations of evaporative cooling on the basis of the kinetic theory of a Bose gas carefully taking into account our specific experimental conditions. We further discuss theoretically a possibility of producing large condensates, more than 10 8 sodium atoms, by simply increasing the number of initial thermal trapped atoms and the corresponding optimization of evaporative cooling.

  12. Materials Performance in Sodium-Cooled Fast Reactors: Past, Present, and Future

    International Nuclear Information System (INIS)

    Natesan, K.; Li Meimei

    2013-01-01

    • This paper gives an overview of the requirements, selection, and performance of materials for in-core and out-of-core components in SFRs. • Globally, sodium-cooled fast reactors have been designed, built, and operated in several countries. A substantial database exists for the existing materials on their functional and mechanical performance. • The 60-yr design life of the SFR presents a significant challenge to the development of database, extrapolation/prediction of long-term performance, and high-temperature design methodology for the structural components. • Licensing of SFR requires a valid assessment of the environmental effects (irradiation, thermal aging, and sodium) on materials performance. • Advanced materials such as, ODS alloys for cladding, Gr91 and 92 F/M steels, and austenitic alloys such as NF709 for structures can improve the economy, safety, and flexibility of SFRs. A substantial database is needed for all these materials and global effort is underway to develop the needed information through experimentation and modeling

  13. Advanced intermediate temperature sodium copper chloride battery

    Science.gov (United States)

    Yang, Li-Ping; Liu, Xiao-Min; Zhang, Yi-Wei; Yang, Hui; Shen, Xiao-Dong

    2014-12-01

    Sodium metal chloride batteries, also called as ZEBRA batteries, possess many merits such as low cost, high energy density and high safety, but their high operation temperature (270-350 °C) may cause several issues and limit their applications. Therefore, decreasing the operation temperature is of great importance in order to broaden their usage. Using a room temperature ionic liquid (RTIL) catholyte composed of sodium chloride buffered 1-ethyl-3-methylimidazolium chloride-aluminum chloride and a dense β″-aluminates solid electrolyte film with 500 micron thickness, we report an intermediate temperature sodium copper chloride battery which can be operated at only 150 °C, therefore alleviating the corrosion issues, improving the material compatibilities and reducing the operating complexities associated with the conventional ZEBRA batteries. The RTIL presents a high ionic conductivity (0.247 S cm-1) at 150 °C and a wide electrochemical window (-2.6 to 2.18 vs. Al3+/Al). With the discharge plateau at 2.64 V toward sodium and the specific capacity of 285 mAh g-1, this intermediate temperature battery exhibits an energy density (750 mWh g-1) comparable to the conventional ZEBRA batteries (728-785 mWh g-1) and superior to commercialized Li-ion batteries (550-680 mWh g-1), making it very attractive for renewable energy integration and other grid related applications.

  14. Computational methodology of sodium-water reaction phenomenon in steam generator of sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Takata, Takashi; Yamaguchi, Akira; Uchibori, Akihiro; Ohshima, Hiroyuki

    2009-01-01

    A new computational methodology of sodium-water reaction (SWR), which occurs in a steam generator of a liquid-sodium-cooled fast reactor when a heat transfer tube in the steam generator fails, has been developed considering multidimensional and multiphysics thermal hydraulics. Two kinds of reaction models are proposed in accordance with a phase of sodium as a reactant. One is the surface reaction model in which water vapor reacts directly with liquid sodium at the interface between the liquid sodium and the water vapor. The reaction heat will lead to a vigorous evaporation of liquid sodium, resulting in a reaction of gas-phase sodium. This is designated as the gas-phase reaction model. These two models are coupled with a multidimensional, multicomponent gas, and multiphase thermal hydraulics simulation method with compressibility (named the 'SERAPHIM' code). Using the present methodology, a numerical investigation of the SWR under a pin-bundle configuration (a benchmark analysis of the SWAT-1R experiment) has been carried out. As a result, the maximum gas temperature of approximately 1,300degC is predicted stably, which lies within the range of previous experimental observations. It is also demonstrated that the maximum temperature of the mass weighted average in the analysis agrees reasonably well with the experimental result measured by thermocouples. The present methodology will be promising to establish a theoretical and mechanical modeling of secondary failure propagation of heat transfer tubes due to such as an overheating rupture and a wastage. (author)

  15. Conceptual design for accelerator-driven sodium-cooled sub-critical transmutation reactors using scale laws

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwang Gu; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    The feasibility study on conceptual design methodology for accelerator-driven sodium-cooled sub-critical transmutation reactors has been conducted to optimize the design parameters from the scale laws and validates the reactor performance with the integrated code system. A 1000 MWth sodium-cooled sub-critical transmutation reactor has been scaled and verified through the methodology in this paper, which is referred to Advanced Liquid Metal Reactor (ALMR). A Pb-Bi target material and a partitioned fuel are the liquid phases, and they are cooled by the circulation of secondary Pb-Bi coolant and by primary sodium coolant, respectively. Overall key design parameters are generated from the scale laws and they are improved and validated by the integrated code system. Integrated Code System (ICS) consists of LAHET, HMCNP, ORIGEN2, and COMMIX codes and some files. Through ICS the target region, the core region, and thermal-hydraulic related regions are analyzed once-through Results of conceptual design are attached in this paper. 5 refs., 4 figs., 1 tab. (Author)

  16. Conceptual core design study for Japan sodium-cooled fast reactor: Review of sodium void reactivity worth evaluation

    International Nuclear Information System (INIS)

    Ohki, Shigeo

    2012-01-01

    The conceptual core design study for a large-scale Japan sodium-cooled fast reactor (JSFR) have been carried out in the framework of the FaCT project. The reference “High-internal conversion” core can satisfy the requirements for enhanced safety, as well as achieving economic competitiveness. In order to increase the design reliability, more rigorous uncertainty evaluation is important. Development of the verification and validation methodology of the core neutronic design method is currently underway. (author)

  17. Inverted Steam Generators for Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Matal, Oldřich; Šimo, Tomáš; Matal, Oldřich Jr.

    2013-01-01

    Conclusions: Two inverted steam generators of the Czech industry provenience have still been in successful operation with no water into sodium leaks at BOR 60 (RIAR Dimitrovgrad, Russian Federation). Micromodular inverted steam generator (MMISG) since 1981 and modular inverted steam generator (MISG) since 1991. In the framework of the CP ESFR project predesign studies of 100 MW (thermal) ISG modules were performed with the consideration of MMISG and MISG design, operational and safety benefits and experience. Development of material and technology for sodium heated steam generators components reflecting contemporary domestic industrial conditions in the Czech Republic was restarted in the years 2003 to 2004 and supported in the years 2008 to 2011 by the European CP ESFR project and by the Ministry of Industry and Trade of the Czech Republic

  18. Radioactive material transport in sodium-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Brehm, W.F.; McGuire, J.C.; Colburn, R.P.; Maffei, H.P.; Olson, W.H.

    1980-03-01

    Trapping devices which remove nuclides from the sodium stream in pre-selected locations away from maintenance areas have been developed and proven successful in in-reactor testing. The release of corrosion product radionuclides as a function of system temperature and oxygen content has been quantitatively evaluated. Ongoing work concentrates on further in-reactor testing of radionuclide removal devices, and characterization of fission product release and deposition from fuel pins with breached-cladding

  19. Sodium tests on an integrated purification prototype for a sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Abramson, R.

    1984-04-01

    This paper describes sodium tests performed on the integrated primary sodium purification prototype of the Creys Malville Super Phenix 1 fast breeder reactor. These tests comprised: - hydrostatic test, - thermal tests, - handling tests. They enabled a number of new technological arrangements to be qualified and provided the necessary information for the design and construction of the Super Phenix 1 purification units

  20. Testing aspects of advanced coherent electron cooling technique

    Energy Technology Data Exchange (ETDEWEB)

    Litvinenko, V.; Jing, Y.; Pinayev, I.; Wang, G.; Samulyak, R.; Ratner, D.

    2015-05-03

    An advanced version of the Coherent-electron Cooling (CeC) based on the micro-bunching instability was proposed. This approach promises significant increase in the bandwidth of the CeC system and, therefore, significant shortening of cooling time in high-energy hadron colliders. In this paper we present our plans of simulating and testing the key aspects of this proposed technique using the set-up of the coherent-electron-cooling proof-of-principle experiment at BNL.

  1. Evaluation of advanced cooling therapy's esophageal cooling device for core temperature control.

    Science.gov (United States)

    Naiman, Melissa; Shanley, Patrick; Garrett, Frank; Kulstad, Erik

    2016-05-01

    Managing core temperature is critical to patient outcomes in a wide range of clinical scenarios. Previous devices designed to perform temperature management required a trade-off between invasiveness and temperature modulation efficiency. The Esophageal Cooling Device, made by Advanced Cooling Therapy (Chicago, IL), was developed to optimize warming and cooling efficiency through an easy and low risk procedure that leverages heat transfer through convection and conduction. Clinical data from cardiac arrest, fever, and critical burn patients indicate that the Esophageal Cooling Device performs very well both in terms of temperature modulation (cooling rates of approximately 1.3°C/hour, warming of up to 0.5°C/hour) and maintaining temperature stability (variation around goal temperature ± 0.3°C). Physicians have reported that device performance is comparable to the performance of intravascular temperature management techniques and superior to the performance of surface devices, while avoiding the downsides associated with both.

  2. Tribological behavior of inconel 718 in sodium cooled reactor environments

    International Nuclear Information System (INIS)

    Wilson, W.L.; Galioto, T.A.; Schrock, S.L.

    1976-01-01

    Results of the present study on the tribological behavior of Inconel 718 in a sodium environment are summarized as follows: (a) Stroke lengths less than or equal to one-half the test pin diameter result in higher friction coefficients. (b) At elevated temperatures, the formation of a lubricative surface film can significantly influence the frictional behavior. (c) Tangential forces present during static dwell periods result in greater bonding tendencies. (d) Increasing contact pressure during static dwell periods results in lower breakaway friction coefficients

  3. Design study on simplification of secondary sodium cooling system for sodium cooled FBRs. Study result from JFY2000 to JFY2001

    International Nuclear Information System (INIS)

    Hori, Toru; Kawasaki, Nobuchika; Konomura, Mamoru

    2002-09-01

    For the 'Feasibility Studies on Commercialized Fast Reactor System' , various concepts with the simplified secondary sodium cooling system were designed, and the feasibility of technical issues was evaluated by focusing on improvement of economy and safety, especially elimination or mitigation of sodium-water direct interaction on heat transfer tube failure accident. In JFY 2000, 8 concepts with inert intermediate media were evaluated from standpoints of economy, safety, and structure integrity. And as promising candidates, the Pb-Bi pool type SG and the Pb-Bi tube type SG (concentric triple-walled tube) were selected, which had low cost compared with conventional IHX and SG system, and had potential of eliminating sodium-water direct interaction by separation of sodium and water tube zone. In JFY 2001, for the Pb-Bi tube type SG, important technical issues on 'Pb-Bi triple-walled tube specification suitable for safety demand', 'safety frame work corresponded to tube failure accident', and 'measures for Pb-Bi leakage into primary sodium loop' were studied, and the SG concept was constructed. In order to eliminate the design supposition of guillotine failure, available design measures for tube specification were tried to extract. But based on vibration characteristics of Pb-Bi triple-walled tube, the time required difference between outer and inner tube failure could not increase largely compared with known double-walled tube. The Pb-Bi tube type SG had potential of cost reduction (81% of cooling system, and 97% of plant), compared with conventional IHX and SG. But finally it was judged that design study on this type SG would not be executed after JFY 2002, due to impossibility of eliminating the design supposition of guillotine failure. (author)

  4. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R. [Inst. for Neutron Physic and Reactor Technology, Karlsruhe Inst. of Technology, Campus Nord (Germany)

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  5. Materials and manufacturing for sodium cooled breeder and fusion power reactor

    International Nuclear Information System (INIS)

    Baldev Raj

    2013-01-01

    The paper narrates definitions of challenges relating to materials and manufacturing for sodium cooled fast reactors thermonuclear fusion reactors. Science and technology developed indigenously but in the context of bench marks in the world is described through examples. Solutions to challenges requires synergy among theoretical physicists, computational chemists, material scientists, metallurgists and engineers with their domains of expertise along with foresight effective management

  6. Study of an electromagnetic pump in a sodium cooled reactor. Design study of secondary sodium main pumps (Joint research)

    International Nuclear Information System (INIS)

    Chikazawa, Yoshitaka; Kisohara, Naoyuki; Hishida, Masahiko; Fujii, Tadashi; Konomura, Mamoru; Ara, Kuniaki; Hori, Toru; Uchida, Akihito; Nishiguchi, Youhei; Nibe, Nobuaki

    2006-07-01

    In the feasibility study on commercialized fast breeder cycle system, a medium scale sodium cooled reactor with 750 MW electricity has been designed. In this study, EMPs are applied to the secondary sodium main pump. The EMPs type is selected to be an annular linear induction pump (ALIP) type with double stators which is used in the 160 m 3 /min EMP demonstration test. The inner structure and electromagnetic features are decided reviewing the 160 m 3 /min EMP. Two dimensional electromagnetic fluid analyses by EAGLE code show that Rms (magnetic Reynolds number times slip) is evaluated to be 1.08 which is less than the stability limit 1.4 confirmed by the 160 m 3 /min EMP test, and the instability of the pump head is evaluated to be 3% of the normal operating pump head. Since the EMP stators are cooled by contacting coolant sodium duct, reliability of the inner structures are confirmed by temperature distribution and stator-duct contact pressure analyses. Besides, a power supply system, maintenance and repair feature and R and D plan of EMP are reported. (author)

  7. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yoon, K. H.; Lee, C. B.

    2014-01-01

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness

  8. Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  9. Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Chan Bock Lee

    2016-10-01

    Full Text Available Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic–martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  10. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  11. Design Requirements of an Advanced HANARO Reactor Core Cooling System

    International Nuclear Information System (INIS)

    Park, Yong Chul; Ryu, Jeong Soo

    2007-12-01

    An advanced HANARO Reactor (AHR) is an open-tank-type and generates thermal power of 20 MW and is under conceptual design phase for developing it. The thermal power is including a core fission heat, a temporary stored fuel heat in the pool, a pump heat and a neutron reflecting heat in the reflector vessel of the reactor. In order to remove the heat load, the reactor core cooling system is composed of a primary cooling system, a primary cooling water purification system and a reflector cooling system. The primary cooling system must remove the heat load including the core fission heat, the temporary stored fuel heat in the pool and the pump heat. The purification system must maintain the quality of the primary cooling water. And the reflector cooling system must remove the neutron reflecting heat in the reflector vessel of the reactor and maintain the quality of the reflector. In this study, the design requirement of each system has been carried out using a design methodology of the HANARO within a permissible range of safety. And those requirements are written by english intend to use design data for exporting the research reactor

  12. Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Sang June; Ha, Kwi Seok; Chang, Won Pyo; Kang, Seok Hun; Lee, Kwi Lim; Choi, Chi Woong; Lee, Seung Won; Yoo, Jin; Jeong, Jae Ho; Jeong, Tae Kyeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-08-15

    The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

  13. Application of the SHARP Toolkit to Sodium-Cooled Fast Reactor Challenge Problems

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Yu, Y. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

    2017-09-30

    The Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit is under development by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of the U.S. Department of Energy, Office of Nuclear Energy. To better understand and exploit the benefits of advanced modeling simulations, the NEAMS Campaign initiated the “Sodium-Cooled Fast Reactor (SFR) Challenge Problems” task, which include the assessment of hot channel factors (HCFs) and the demonstration of zooming capability using the SHARP toolkit. If both challenge problems are resolved through advanced modeling and simulation using the SHARP toolkit, the economic competitiveness of a SFR can be significantly improved. The efforts in the first year of this project focused on the development of computational models, meshes, and coupling procedures for multi-physics calculations using the neutronics (PROTEUS) and thermal-hydraulic (Nek5000) components of the SHARP toolkit, as well as demonstration of the HCF calculation capability for the 100 MWe Advanced Fast Reactor (AFR-100) design. Testing the feasibility of the SHARP zooming capability is planned in FY 2018. The HCFs developed for the earlier SFRs (FFTF, CRBR, and EBR-II) were reviewed, and a subset of these were identified as potential candidates for reduction or elimination through high-fidelity simulations. A one-way offline coupling method was used to evaluate the HCFs where the neutronics solver PROTEUS computes the power profile based on an assumed temperature, and the computational fluid dynamics solver Nek5000 evaluates the peak temperatures using the neutronics power profile. If the initial temperature profile used in the neutronics calculation is reasonably accurate, the one-way offline method is valid because the neutronics power profile has weak dependence on small temperature variation. In order to get more precise results, the proper temperature profile for initial neutronics calculations was obtained from the

  14. Aspects of unconventional cores for large sodium cooled power reactors; evaluation of a literature survey

    International Nuclear Information System (INIS)

    Kiefhaber, E.

    1978-10-01

    The report gives an overview of a literature study on the application of unconventional cores for sodium cooled fast reactors. Different types of unconventional cores (heterogeneous cores, pancake cores, moderated cores and others) are compared with conventional cores, which are characterized by a cylindrical geometry with two or three fissile zones surrounded by an axial and a radial blanket. The main parameters of interest in this comparison are the neutronic parameters sodium void and Doppler effect, the breeding properties and the steel damage. Consequences for the core safety and the overall plant design are also mentioned

  15. Mechanical Design Features of the KALIMER-600 Sodium-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Park, Chang Gyu; Kim, Jong Bum [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    KALIMER-600 is a sodium cooled reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types.

  16. Mechanical Design Features of the KALIMER-600 Sodium-Cooled Reactor

    International Nuclear Information System (INIS)

    Lee, Jae Han; Park, Chang Gyu; Kim, Jong Bum

    2005-01-01

    KALIMER-600 is a sodium cooled reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types

  17. Health and Safety Considerations Associated with Sodium-Cooled Experimental Nuclear Fuel Dismantlement

    Energy Technology Data Exchange (ETDEWEB)

    Carvo, Alan E. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    Between the mid-1970s and the mid-1980s Sandia National Laboratory constructed eleven experimental assemblies to simulate debris beds formed in a sodium-cooled fast breeder reactor. All but one of the assemblies were irradiated. The experimental assemblies were transferred to the Idaho National Laboratory (INL) in 2007 and 2008 for storage, dismantlement, recovery of the uranium for reuse in the nuclear fuel cycle, and disposal of unneeded materials. This paper addresses the effort to dismantle the assemblies down to the primary containment vessel and repackage them for temporary storage until such time as equipment necessary for sodium separation is in place.

  18. Gas-cooled reactors for advanced terrestrial applications

    International Nuclear Information System (INIS)

    Kesavan, K.; Lance, J.R.; Jones, A.R.; Spurrier, F.R.; Peoples, J.A.; Porter, C.A.; Bresnahan, J.D.

    1986-01-01

    Conceptual design of a power plant on an inert gas cooled nuclear coupled to an open, air Brayton power conversion cycle is presented. The power system, called the Westinghouse GCR/ATA (Gas-Cooled Reactors for Advanced Terrestrial Applications), is designed to meet modern military needs, and offers the advantages of secure, reliable and safe electrical power. The GCR/ATA concept is adaptable over a range of 1 to 10 MWe power output. Design descriptions of a compact, air-transportable forward base unit for 1 to 3 MWe output and a fixed-base, permanent installation for 3 to 10 MWe output are presented

  19. Natural Circulation Phenomena and Modelling for Advanced Water Cooled Reactors

    International Nuclear Information System (INIS)

    2012-03-01

    The role of natural circulation in advanced water cooled reactor design has been extended with the adoption of passive safety systems. Some designs utilize natural circulation to remove core heat during normal operation. Most passive safety systems used in evolutionary and innovative water cooled reactor designs are driven by natural circulation. The use of passive systems based on natural circulation can eliminate the costs associated with the installation, maintenance and operation of active systems that require multiple pumps with independent and redundant electric power supplies. However, considering the weak driving forces of passive systems based on natural circulation, careful design and analysis methods must be employed to ensure that the systems perform their intended functions. Several IAEA Member States with advanced reactor development programmes are actively conducting investigations of natural circulation to support the development of advanced water cooled reactor designs with passive safety systems. To foster international collaboration on the enabling technology of passive systems that utilize natural circulation, in 2004 the IAEA initiated a coordinated research project (CRP) on Natural Circulation Phenomena, Modelling and Reliability of Passive Systems that Utilize Natural Circulation. Three reports were published within the framework of this CRP. The first report (IAEA-TECDOC-1474) contains the material developed for the first IAEA training course on natural circulation in water cooled nuclear power plants. The second report (IAEA-TECDOC-1624) describes passive safety systems in a wide range of advanced water cooled nuclear power plant designs, with the goal of gaining insights into system design, operation and reliability. This third, and last, report summarizes the research studies completed by participating institutes during the CRP period.

  20. Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Weaver, K.D.; Herring, J.S.; Macdonald, P.E.

    2001-01-01

    Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

  1. Application of objective provision tree to development of standard review plan for sodium-cooled fast reactor nuclear design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo-Hoon; Suh, Namduk; Choi, Yongwon; Shin, Andong [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2016-06-15

    A systematic methodology was developed for the standard review plan for sodium-cooled fast reactor nuclear design. The process is first to develop an objective provision tree of sodium-cooled fast reactor for the reactivity control safety function. The provision tree is generally developed by designer to confirm whether the design satisfies the defense-in-depth concept. Then applicability of the current standard review plan of nuclear design for light water reactor to sodium-cooled fast reactor was evaluated and complemented by the developed objective provision tree.

  2. Sodium-cooled reactors, objectives, achieved technical state and development trends

    International Nuclear Information System (INIS)

    Wolff, U.

    1988-01-01

    The use of fossil fuels to cover the future world-wide energy demand alone would rapidly deplete these ressources, especially oil and gas. Today's knowledge suggests the enhanced exploitation of solar energy, nuclear fusion and the application of uranium in sodium-cooled breeder reactors as the alternative energies offering a great potential. The sodium-cooled reactor outdistances the other options in terms of development. Its technical feasibility and safe operation have been verified and its profitability appears to be possible when using today's technology. The verification of its profitability while maintaining a high safety level is the overriding task for the future. The paper discusses corresponding activities in the USA, the USSR, Japan and Western Europe. (orig.) [de

  3. Effects of Nuclear Energy on Sustainable Development and Energy Security: Sodium-Cooled Fast Reactor Case

    Directory of Open Access Journals (Sweden)

    Sungjoo Lee

    2016-09-01

    Full Text Available We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indicators to meet data availability, nuclear energy relevancy, comparability among energy options, and fit with Korean energy policy objectives. The results show that sodium-cooled fast reactors is a better alternative than existing nuclear power as well as coal electricity generation across social, economic and environmental dimensions. Our method makes comparison between energy alternatives easier, thereby clarifying consequences of different energy policy decisions.

  4. Thermal hydraulics of sodium-cooled fast reactors - key issues and highlights

    International Nuclear Information System (INIS)

    Ninokata, H.; Kamide, H.

    2011-01-01

    In this paper key issues and highlighted topics in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. In particular, design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are on-going. Here, progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed. (author)

  5. A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

  6. New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

    Energy Technology Data Exchange (ETDEWEB)

    Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2017-07-15

    The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

  7. Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors

    OpenAIRE

    Verma, Vasudha

    2017-01-01

    Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility stud...

  8. Application of hafnium hydride control rod to large sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Kazumi, E-mail: kazumi_ikeda@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Moriwaki, Hiroyuki, E-mail: hiroyuki_moriwaki@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Ohkubo, Yoshiyuki, E-mail: yoshiyuki_okubo@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Iwasaki, Tomohiko, E-mail: tomohiko.iwasaki@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, Aoba, Aramaki, Aoba-ku, Sendai-shi, Miyagi-ken 980-8579 (Japan); Konashi, Kenji, E-mail: konashi@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki-ken 311-1313 (Japan)

    2014-10-15

    Highlights: • Application of hafnium hydride control rod to large sodium cooled fast breeder reactor. • This paper treats application of an innovative hafnium hydride control rod to a large sodium cooled fast breeder reactor. • Hydrogen absorption triples the reactivity worth by neutron spectrum shift at H/Hf ratio of 1.3. • Lifetime of the control rod quadruples because produced daughters of hafnium isotopes are absorbers. • Nuclear and thermal hydraulic characteristics of the reactor are as good as or better than B-10 enriched boron carbide. - Abstract: This study treats the feasibility of long-lived hafnium hydride control rod in a large sodium-cooled fast breeder reactor by nuclear and thermal analyses. According to the nuclear calculations, it is found that hydrogen absorption of hafnium triples the reactivity by the neutron spectrum shift at the H/Hf ratio of 1.3, and a hafnium transmutation mechanism that produced daughters are absorbers quadruples the lifetime due to a low incineration rate of absorbing nuclides under irradiation. That is to say, the control rod can function well for a long time because an irradiation of 2400 EFPD reduces the reactivity by only 4%. The calculation also reveals that the hafnium hydride control rod can apply to the reactor in that nuclear and thermal characteristics become as good as or better than 80% B-10 enriched boron carbide. For example, the maximum linear heat rate becomes 3% lower. Owing to the better power distribution, the required flow rate decreases approximately by 1%. Consequently, it is concluded on desk analyses that the long lived hafnium hydride control rod is feasible in the large sodium-cooled fast breeder reactor.

  9. IAEA Workshop (Training Course) on Codes and Standards for Sodium Cooled Fast Reactors. Working Material

    International Nuclear Information System (INIS)

    2010-01-01

    The training course consisted of lectures and Q&A sessions. The lectures dealt with the history of the development of Design Codes and Standards for Sodium Cooled Fast Reactors (SFRs) in the respective country, the detailed description of the current design Codes and Standards for SFRs and their application to ongoing Fast Reactor design projects, as well as the ongoing development work and plans for the future in this area. Annex 1 contains the detailed Workshop program

  10. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Plevacova, K. [CEA, DEN, STRI, LMA, Cadarache, 3108 St. Paul lez Durance (France); Journeau, C., E-mail: christophe.journeau@cea.fr [CEA, DEN, STRI, LMA, Cadarache, 3108 St. Paul lez Durance (France); Piluso, P. [CEA, DEN, STRI, LMA, Cadarache, 3108 St. Paul lez Durance (France); Zhdanov, V.; Baklanov, V. [IAE, National Nuclear Centre, Material Structure Investigation Dept., Krasnoarmeiskaya, 10, Kurchatov City (Kazakhstan); Poirier, J. [CEMHTI, 1D, av. de la Recherche Scientifique, 45071 Orleans Cedex 2 (France)

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U{sub x}, Zr{sub y})O{sub 2-z} water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO{sub 2}, the zirconium carbide coating keeps its role of protective barrier with UO{sub 2}-Al{sub 2}O{sub 3} below 2000 deg. C but does not resist to a UO{sub 2}-Eu{sub 2}O{sub 3} mixture.

  11. Extended stability of intravenous 0.9% sodium chloride solution after prolonged heating or cooling.

    Science.gov (United States)

    Puertos, Enrique

    2014-03-01

    The primary objective of this study was to evaluate the stability and sterility of an intravenous 0.9% sodium chloride solution that had been cooled or heated for an extended period of time. Fifteen sterile 1 L bags of 0.9% sodium chloride solution were randomly selected for this experiment. Five bags were refrigerated at an average temperature of 5.2°C, 5 bags were heated at an average temperature of 39.2°C, and 5 bags were stored at an average room temperature of 21.8°C to serve as controls. All samples were protected from light and stored for a period of 199 days prior to being assayed and analyzed for microbial and fungal growth. There was no clinically significant difference in the mean sodium values between the refrigerated samples, the heated samples, and the control group. There were no signs of microbial or fungal growth for the duration of the study. A sterile intravenous solution of 0.9% sodium chloride that was heated or cooled remained stable and showed no signs of microbial or fungal growth for a period of 199 days. This finding will allow hospitals and emergency medical technicians to significantly extend the expiration date assigned to these fluids and therefore obviate the need to change out these fluids every 28 days as recommended by the manufacturer.

  12. Advanced technologies for water cooled reactors 1990. Pt. 1

    International Nuclear Information System (INIS)

    1991-05-01

    The meeting was attended by 20 participants from 12 countries who reviewed and discussed the status and progress of national programmes on advanced water-cooled reactors and recommended to the Scientific Secretary a comprehensive programme for 1991/1992 which would support technology development programmes in IWGATWR Member States. This summary report outlines the activities of IWGATWR since its Second Meeting in June 1988 and main results of the Third Meeting

  13. Advanced technologies for water cooled reactors 1990. Pt. 2

    International Nuclear Information System (INIS)

    1991-05-01

    The main purpose of the meeting was to review and discuss the status of national programmes, the progress achieved since the last meeting held in June 1988 in the field of advanced technologies and design trends for existing and future water cooled reactors. 24 specialists from 14 countries and the IAEA took part in the meeting and 12 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  14. Numerical approach for quantification of self wastage phenomena in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Jang, Sung Hyun; Takata, Takashi; Yamaguchi, Akira; Uchbori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

    2015-01-01

    Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called 'self-wastage phenomena'. A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodium-water reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm)

  15. Numerical approach for quantification of self wastage phenomena in sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Sung Hyun; Takata, Takashi [Graduate School of Engineering, Osaka University, Osaka (Japan); Yamaguchi, Akira [Graduate School of Engineering, The University of Tokyo, Ibaraki (Japan); Uchbori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki [Japan Atomic Energy Agency, Ibaraki (Japan)

    2015-10-15

    Sodium-cooled fast breeder reactors use liquid sodium as a moderator and coolant to transfer heat from the reactor core. The main hazard associated with sodium is its rapid reaction with water. Sodium-water reaction (SWR) takes place when water or vapor leak into the sodium side through a crack on a heat-transfer tube in a steam generator. If the SWR continues for some time, the SWR will damage the surface of the defective area, causing it to enlarge. This self-enlargement of the crack is called 'self-wastage phenomena'. A stepwise numerical evaluation model of the self-wastage phenomena was devised using a computational code of multicomponent multiphase flow involving a sodium-water chemical reaction: sodium-water reaction analysis physics of interdisciplinary multiphase flow (SERAPHIM). The temperature of gas mixture and the concentration of NaOH at the surface of the tube wall are obtained by a numerical calculation using SERAPHIM. Averaged thermophysical properties are used to assess the local wastage depth at the tube surface. By reflecting the wastage depth to the computational grid, the self-wastage phenomena are evaluated. A two-dimensional benchmark analysis of an SWAT (Sodium-Water reAction Test rig) experiment is carried out to evaluate the feasibility of the numerical model. Numerical results show that the geometry and scale of enlarged cracks show good agreement with the experimental result. Enlarged cracks appear to taper inward to a significantly smaller opening on the inside of the tube wall. The enlarged outer diameter of the crack is 4.72 mm, which shows good agreement with the experimental data (4.96 mm)

  16. The role of the IAEA in advanced technologies for water-cooled reactors

    International Nuclear Information System (INIS)

    Cleveland, J.

    1996-01-01

    The role of the IAEA in advanced technologies for water-cooled reactors is described, including the following issues: international collaboration ways through international working group activities; IAEA coordinated research programmes; cooperative research in advanced water-cooled reactor technology

  17. Transport of sodium through the cover gas of a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Clement, C.F.; Hawtin, P.

    1977-01-01

    Idealised models are presented for sodium vapour transport through argon or helium and the subsequent roof condensation. For both gases the dominant heat transfer mechanism from the pool is radiation but the mass transport process is convection for argon and diffusion for helium. For argon a theory based on work of Hills and Szekely is presented which predicts a heat transfer rate independent of the actual amount of condensation occurring in the cavity, and which suggests a mass transfer rate close to that calculated in the absence of condensation. Experimental determination of the temperature and velocity flow characteristics are desirable to examine and improve on the suspect basic assumption of the theory that the velocity flow pattern is unaffected by condensation. For helium diffusion theory predicts a mass transfer rate an order of magnitude smaller than for argon, but only a slightly smaller overall heat transfer rate because of the dominance of radiation. (author)

  18. The Modification of Sodium Polyacrylate Water Solution Cooling Properties by AL2O3

    Directory of Open Access Journals (Sweden)

    Wojciech Gęstwa

    2010-01-01

    Based on cooling curves, it can be concluded that for the water solution of sodium polyacrylate with AL2O3 nanoparticles in comparison to water and 10% polymer water solution lower cooling speed is obtained. The cooling medium containing nanoparticles provides lower cooling speed in the smallest surface austenite occurance (500–600 C in the charts of the CTP for most nonalloy structural steels and low-alloy steels. However lower cooling temperature at the beginning of martensitic transformation causes the formation of smaller internal stresses, leading to smaller dimensional changes and hardening deformation. For the quenching media the wetting angle was appointed by the drop-shape method. These studies showed the best wettability of polymer water solution (sodium polyacrylate with the addition of AL2O3 nanoparticles, whose wetting angle was about 65 degrees. Obtaining the smallest wetting angle for the medium containing nanoparticles suggests that the heat transfer to the cooling medium is larger. This allows slower cooling at the same time ensuring its homogeneity. The obtained values of wetting angle confirm the conclusions drawn on the basis of cooling curves and allowus to conclude that in the case of the heat transfer rate it will have a lower value than for water and 10% polymer water solution. In the research on hardened carburized steel samples C10 and 16MnCr5 surface hardness, impact strength and changes in the size of cracks in Navy C-ring sample are examined. On this basis of the obtained results it can be concluded that polymer water solution with nanoparticles allows to obtain a better impact strength at comparable hardness on the surface. Research on the dimensional changes on the basis of the sample of Navy C-ring also shows small dimensional changes for samples carburized and hardened in 10% polymer water solution with the addition of nanoparticles AL2O3. Smaller dimensional changes were obtained for samples of steel 16MnCr5 thanfar C10. The

  19. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Yun, Sunghwan; Kim, Sang Ji

    2015-01-01

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

  20. Subchannel analysis of sodium-cooled reactor fuel assemblies with annular fuel pins

    International Nuclear Information System (INIS)

    Memmott, Matthew; Buongiorno, Jacopo; Hejzlar, Pavel

    2009-01-01

    Using a RELAP5-3D subchannel analysis model, the thermal-hydraulic behavior of sodium-cooled fuel assemblies with internally and externally cooled annular fuel rods was investigated, in an effort to enhance the economic performance of sodium-fast reactors by increasing the core power density, decreasing the core pressure drop, and extending the fuel discharge burnup. Both metal and oxide fuels at high and low conversion ratios (CR=0.25 and CR=1.00) were investigated. The externally and internally cooled annular fuel design is most beneficial when applied to the low CR core, as clad temperatures are reduced by up to 62.3degC for the oxide fuel, and up to 18.5degC for the metal fuel. This could result in a power uprates of up to ∼44% for the oxide fuel, and up to ∼43% for the metal fuel. The use of duct ribs was explored to flatten the temperature distribution at the core outlet. Subchannel analyses revealed that no fuel melting would occur in the case of complete blockage of the hot interior-annular channel for both metal and oxide fuels. Also, clad damage would not occur for the metal fuel if the power uprate is 38% or less, but would indeed occur for the oxide fuel. (author)

  1. Preparation of a monoenergetic sodium beam by laser cooling and deflection

    International Nuclear Information System (INIS)

    Nellessen, J.; Sengstock, K.; Muller, J.H.; Ertmer, W.; Wallis, H.

    1989-01-01

    This paper reports on a sodium atomic beam with a density of approx. 10 5 at cm 3 within a velocity interval of less than 3 m/s with a mean velocity of typically 50-160 m/s which has been produced by laser deflection of a laser cooled atomic beam. Laser cooling with the frequency chirp method decelerates and cools a considerable part of an atomic beam into a narrow velocity group with a temperature of approx 30 mK as a part of the resulting atomic beam. This velocity group has been selectively deflected up to 30 degrees - 40 degrees using a light field with k vectors always perpendicular to the atomic trajectory. If the light field is prepared by use of a cylindrical lens, the angle of deflection is nearly independent from the actual orbit radius. For a laser frequency detuning of about one natural linewidth to the red, the strong frequency dependence of the light pressure force leads to a beam collimation via detuning-locking of the atomic trajectory. To avoid optical pumping we used a frequency modulated laser beam with a sideband spacing matched to the hyperfine splitting of the ground state. As the cooling was performed by the frequency chirp method, one can use a part of the cooling laser beam as deflecting laser beam. Typical velocity distributions in the deflected and undeflected atomic beam, measured 22 cm downstream the deflection zone. It shows the perfect transfer of the cooled velocity group from the laser cooled beam into the deflected beam; curve c) shows as comparison the result for the deflection of the initial thermal atomic beam

  2. Utility industry evaluation of the Sodium Advanced Fast Reactor

    International Nuclear Information System (INIS)

    Burstein, S.; DelGeorge, L.O.; Tramm, T.R.; Gibbons, J.P.; High, M.D.; Neils, G.H.; Pilmer, D.F.; Tomonto, J.R.; Wells, J.T.

    1990-02-01

    A team of utility industry representatives evaluated the Sodium Advanced Fast Reactor plant design, a current liquid metal reactor design created by an industrial team led by Rockwell International under Department of Energy sponsorship. The utility industry team concluded that the plant design offers several attractive characteristics, especially in the safety arena, as well as preserving the traditional attraction of liquid metal reactors, very high fuel utilization. Specific comments and recommendations are provided as a contribution towards improving an already attractive plant design. 18 refs

  3. Analysis of self-wastage phenomena of micro leak caused by sodium-water reaction in sodium-cooled fast breeder reactor through simulant experiment

    International Nuclear Information System (INIS)

    Jang, Sunghyon; Takata, Takashi; Yamaguchi, Akira

    2014-01-01

    Self-wastage phenomena are an enlargement of a leak on the heat transfer tube caused by a corrosive sodium-water reaction (SWR) in a steam generator (SG) of sodium-cooled fast breeder reactor (SFR). If the steam generator operates for sometimes under this condition, the self-wastage phenomena start from the sodium side and advance through the tube thickness. The leak rate stays almost constant level until the wastage reaches the sodium side, however, when the thin diaphragm of the tube wall is removed, the leak rate sharply increase, and it may bring a secondary failure of the surrounding heat transfer tubes. The design and safety concern is a possibility of the secondary failure of nearby SG tubes that could cause undesirable development of the accidents. One needs to evaluate the increased resultant leak rate due to the self-wastage phenomenon. Therefore, a quantification of the diameter of enlarged leak is needed to estimate the resultant leak rate. For this purpose, a simulant self-wastage experiment was proposed to investigate the self-enlargement of the leak so that evaluate the mechanism of the Self-wastage. In the experiment, high concentrated hydrochloric acid (HCl) is injected to the reaction tank that is filled sodium hydroxide (NaOH) solution through a nozzle made by paraffin wax. The self-enlargement of the leak was evaluated by considering the melted nozzle due to the reaction heat released from the Neutralization reaction. Also, a numerical investigation has been carried out to evaluate the enlarged nozzle and validate the results of experimental methodology. Based on the experimental and computational results, it is found that despite initial leak rate, there is an upper limit in the enlarged nozzle. These results show a similar tendency with the experimental result of SWAT-4 experiment carried out by Power Reactor and Nuclear Fuel Development Corporation (PNC), Japan. Furthermore, the increased resultant leak rate is evaluated using the enlarged

  4. Progress Report on Sodium Cooled Fast Breeder Reactor Development in Japan, April 1975

    International Nuclear Information System (INIS)

    Tomabechi, K.

    1975-01-01

    The progress of the sodium cooled fast Breeder Reactor development in Japan in the past 12 months can be summarized as follows. Installation of all the components of the Experimental Fast Reactor, ''JOYO'', was completed in the end of the last year and various commissioning tests of the reactor began in January 1975. It is planned to charge sodium into the reactor in coming fall and the first criticality experiment is currently planned in the summer 1976. Most of the research and development works for ''JOYO'' are nearing completion. These include an endurance test of 3 prototype primary sodium pump for 12,000 hours. 86 core fuel subassemblies and 220 blanket subassemblies, a sufficient number for composing the initial core, have already been fabricated. Concerning the Prototype Fast Breeder Reactor, ''MONJU'', design activity as well as relevant research and development works are continued. A siting problem exists and it is hoped to be resolved soon. Of the research and development works, a significant achievement in the past 12 months can be a successful operation at full power of the 50 MW Steam Generator Test Facility. This facility was put into operation at full power in June 1974. No leak of water into sodium has been experienced with operation of the steam generator tested. The steam generator is being dismantled for a detailed inspection originally planned

  5. KNK II, Compact Sodium-Cooled Reactor in the Nuclear Research Center Karlsruhe

    International Nuclear Information System (INIS)

    1978-01-01

    The report gives an overview of the project of the sodium-cooled fast reactor KNK II in the nuclear research center KfK in Karlsruhe. This test reactor was the preparatory stage of the prototype plant SNR 300 and had several goals: to train operating personal, to practice the licensing procedures in Germany, to get experience with the sodium technology and to serve as a test bed for fast breeder core components. The report contains contributions of KfK as the owner and project managing organization, of INTERATOM as the design and construction company and of the KBG as the plant operating organization. Experience with and results of relevant aspects of the project are tackled: project management, reactor core and component design, safety questions and licensing, plant design and test programs [de

  6. Heat-transfer in a partially-blocked sodium-cooled rod bundle

    International Nuclear Information System (INIS)

    Han, J.T.

    1979-01-01

    Heat transfer coefficients were experimentally determined for 31-rod sodium-cooled bundle with a 6-subchannel central blockage. The Nusselt number is presented as a function of the Peclet number for both the free flow region undisturbed by the blockage and the wake region immediately downstream of the blockage. Results are compared with the existing correlations for liquid metals. The heat transfer coefficient was generally higher in the unblocked free flow region than in the wake region. A leak at the blockage improved the heat transfer coefficient in the wake region

  7. The cryogenic cooling program at the Advanced Photon Source

    International Nuclear Information System (INIS)

    Rogers, C.S.; Mills, D.M.; Assoufid, L.

    1994-06-01

    This paper describes the experimental and analytical program in cryogenic cooling of high-heat-load optics at the Advanced-Photon Source. A prototype liquid nitrogen pumping system has been procured. This pump provides a variable flow rate of 1 to 10 gpm of pressurized liquid nitrogen and is sized to handle up to 5 kW of optic heat load. Also, a high-vacuum, double-crystal monochromator testing tank has been fabricated. This system will be used to test cryogenic crystals at existing synchrotron sources. A finite element analysis has been performed for a cryogenically cooled Si crystal in the inclined geometry for Undulator A at 100 mA. The inclination angle was 80 degrees. It was set to diffract from the (111) planes at the first harmonic energy of 4.2 keV. The maximum slope error in the diffraction plane was calculated to be about 1 μrad with a peak temperature of 94 K. An analysis has also been performed for a cryogenically-cooled ''thin'' crystal oriented in the Bragg geometry which accepts 87% of the lst harmonic photons at 3.866 keV. The total absorbed power was 131 W at 100 mA current and the peak temperature was 124 K

  8. Recycling option search for a 600 MWE sodium-cooled transmutation fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Kyo; Kim, Myung Hyun [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

    2015-02-15

    Four recycling scenarios involving pyroprocessing of spent fuel (SF) have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR), KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU) SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro- SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC) decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs). If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE) isotopes. The RE isotope recovery factor should be lowered to 20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

  9. Recycling option search for a 600-MWe sodium-cooled transmutation fast reactor

    Directory of Open Access Journals (Sweden)

    Yong Kyo Lee

    2015-02-01

    Full Text Available Four recycling scenarios involving pyroprocessing of spent fuel (SF have been investigated for a 600-MWe transmutation sodium-cooled fast reactor (SFR, KALIMER. Performance evaluation was done with code system REBUS connected with TRANSX and TWODANT. Scenario Number 1 is the pyroprocessing of Canada deuterium uranium (CANDU SF. Because the recycling of CANDU SF does not have any safety problems, the CANDU-Pyro-SFR system will be possible if the pyroprocessing capacity is large enough. Scenario Number 2 is a feasibility test of feed SF from a pressurized water reactor PWR. The sensitivity of cooling time before prior to pyro-processing was studied. As the cooling time increases, excess reactivity at the beginning of the equilibrium cycle (BOEC decreases, thereby creating advantageous reactivity control and improving the transmutation performance of minor actinides. Scenario Number 3 is a case study for various levels of recovery factors of transuranic isotopes (TRUs. If long-lived fission products can be separated during pyroprocessing, the waste that is not recovered is classified as low- and intermediate-level waste, and it is sufficient to be disposed of in an underground site due to very low-heat-generation rate when the waste cooling time becomes >300 years at a TRU recovery factor of 99.9%. Scenario Number 4 is a case study for the recovery factor of rare earth (RE isotopes. The RE isotope recovery factor should be lowered to ≤20% in order to make sodium void reactivity less than <7$, which is the design limit of a metal fuel.

  10. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    2012-12-01

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  11. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    2012-06-01

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  12. Thermal-hydraulic numerical simulation of fuel sub-assembly for Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Saxena, Aakanksha

    2014-01-01

    The thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR). First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry. The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (∼20 C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (∼1 C-2 C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10 Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures. (author) [fr

  13. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  14. Contribution to perfecting eddy current testing of steam generator tubes of sodium cooled breeders: description of the Monacault loop for the study of sodium deposit influence

    International Nuclear Information System (INIS)

    Lapicore, A.; Lemarquis, J.C.; Oberlin, C.; Pigeon, M.

    1981-12-01

    In the event of sodium-water reaction in the steam generator of a sodium cooled breeder reactor, it is essential to be able to monitor the local loss of thickness of the tubes located in the reaction area. A method for monitoring the tubes by an eddy current probe is being developed for Super Phenix. The sodium deposits on the outer wall of the tubes, as well as their prolonged contact with high temperature sodium are likely to bring about a change in the signals picked up. A test loop, Monacault, has been built in order to clarify the importance of these parameters (effect of sodium deposits, reproducibility of the wetting at different temperatures). It includes three test cells containing the sample tubes having a total of 61 standard defects to be tested. The first results on the wetting of tubes are given and discussed [fr

  15. Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

    Energy Technology Data Exchange (ETDEWEB)

    Scaller, K; Vrillon, B

    1980-02-01

    Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component.

  16. Preliminary Design of Compressor Impeller for innovative Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jekyoung; Cho, Seongkuk; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Cha, Jae Eun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    For nuclear power plant application, applying S-CO{sub 2} Brayton cycle to Sodium cooled Fast Reactors and Small Modular Reactors are currently considered and active research is being performed by various research institutions and universities. As a part of research activities on the SCO{sub 2} Brayton cycle development for a nuclear power system, KAIST joint research team is currently working on an innovative Sodium cooled Fast Reactor (iSFR) development which utilizes S-CO{sub 2} Brayton cycle as its power conversion system. Various research subjects including reactor physics, thermo-hydraulics, material, cycle analysis and system integration are being considered as research issues currently. However, technical issues rising from dramatic change of thermodynamic property of CO{sub 2} near the critical point still remain as problems to be solved. As a result, 3D impeller model generation based on 1D mean stream line analysis results was successfully performed for non-airfoil blades. Since 3D model generation module works successfully, KAIST{sub T}MD can support 3D CFD analysis for internal flow structure in the designed impeller. Compressor loss mechanisms are complex phenomena and these are difficulties to be modeled while considering each loss mechanism separately.

  17. Tentative design-philosophy for bellows in sodium cooled fast breeder reactors pipings

    International Nuclear Information System (INIS)

    Scaller, K.; Vrillon, B.

    1980-01-01

    Expansion joints have proved to be reliable components, when properly designed and realized, in normal industrial equipment. But nevertheless bellows have not been employed widely in nuclear reactors and almost not in sodium cooled fast breeder reactors, where use of expansion-joints could considerably shorten the length of pipelines and, in consequence, lower the cost of the power plant. In the framework of its research and development program on fast reactors the French Atomic Energy.Commission, in cooperation with the industry, develops guidelines, backed up by experiments, to allow a safe design of pipe-lines and compensating-devices. The main points of these guidelines are discussed in this paper with the understanding, that they are tentative rules subject to changes. The guidelines are a complement to existing rules, like ASME - Code III, Code Case 1481, standards of the EJMA Preliminary Draft for Code Case Class I, Expansion Joints in Piping systems and suppliers' rules for the special case of application to sodium cooled fast breeder reactors. Relatively small diameters and easily accessible expansion joints, on control rods and valves for example, are not concerned. These guidelines do not apply to the bellows which are used as an integral part of a component

  18. Preliminary Validation of the MATRA-LMR Code Using Existing Sodium-Cooled Experimental Data

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Kim, Sangji

    2014-01-01

    The main objective of the SFR prototype plant is to verify TRU metal fuel performance, reactor operation, and transmutation ability of high-level wastes. The core thermal-hydraulic design is used to ensure the safe fuel performance during the whole plant operation. The fuel design limit is highly dependent on both the maximum cladding temperature and the uncertainties of the design parameters. Therefore, an accurate temperature calculation in each subassembly is highly important to assure a safe and reliable operation of the reactor systems. The current core thermalhydraulic design is mainly performed using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code, which has been already validated using the existing sodium-cooled experimental data. In addition to the SLTHEN code, a detailed analysis is performed using the MATRA-LMR (Multichannel Analyzer for Transient and steady-state in Rod Array-Liquid Metal Reactor) code. In this work, the MATRA-LMR code is validated for a single subassembly evaluation using the previous experimental data. The MATRA-LMR code has been validated using existing sodium-cooled experimental data. The results demonstrate that the design code appropriately predicts the temperature distributions compared with the experimental values. Major differences are observed in the experiments with the large pin number due to the radial-wise mixing difference

  19. Conceptual core designs for a 1200 MWe sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Joo, H. K.; Lee, K. B.; Yoo, J. W.; Kim, Y. I.

    2008-01-01

    The conceptual core design for a 1200 MWe sodium cooled fast reactor is being developed under the framework of the Gen-IV SFR development program. To this end, three core concepts have been tested during the development of a core concept: a core with an enrichment split fuel, a core with a single-enrichment fuel with a region-wise varying clad thickness, and a core with a single-enrichment fuel with non-fuel rods. In order to optimize a conceptual core configuration which satisfies the design targets, a sensitivity study of the core design parameters has been performed. Two core concepts, the core with an enrichment-split fuel and the core with a single-enrichment fuel with a region-wise varying clad thickness, have been proposed as the candidates of the conceptual core for a 1200 MWe sodium cooled fast reactor. The detailed core neutronic, fuel behavior, thermal, and safety analyses will be performed for the proposed candidate core concepts to finalize the core design concept. (authors)

  20. Neutronic/Thermal-hydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    International Nuclear Information System (INIS)

    Ragusa, Jean; Siegel, Andrew; Ruggieri, Jean-Michel

    2010-01-01

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  1. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

    2010-09-28

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  2. Sodium-cooled fast reactor core designs for transmutation of MHR spent fuel

    International Nuclear Information System (INIS)

    Hong, S. G.; Kim, Y. H.; Venneri, F.

    2010-01-01

    In this paper, the core design analyses of sodium cooled fast reactors (SFR) are performed for the effective transmutation of the DB (Deep Burn)-MHR (Modular Helium Reactor). In this concept, the spent fuels of DB-MHR are transmuted in SFRs with a closed fuel cycle after TRUs from LWR are first incinerated in a DB-MHR. We introduced two different type SFR core designs for this purpose, and evaluated their core performance parameters including the safety-related parameters. In particular, the cores are designed to have lower transmutation rate relatively to our previous work so as to make the fuel characteristics more feasible. The first type cores which consist of two enrichment regions are typical homogeneous annular cores and they rate 900 MWt power. On the other hand, the second type cores which consist of a central non-fuel region and a single enrichment fuel region rate relatively higher power of 1500 MWt. For these cores, the moderator rods (YH 1.8 ) are used to achieve less positive sodium void worth and the more negative Doppler coefficient because the loading of DB-MHR spent fuel leads to the degradation of these safety parameters. The analysis results show that these cores have low sodium void worth and negative reactivity coefficients except for the one related with the coolant expansion but the coolant expansion reactivity coefficient is within the typical range of the typical SFR cores. (authors)

  3. Demonstration of leak-before-break in Japan Sodium cooled Fast Reactor (JSFR) pipes

    International Nuclear Information System (INIS)

    Wakai, Takashi; Machida, Hideo; Yoshida, Shinji; Xu, Yang; Tsukimori, Kazuyuki

    2014-01-01

    This paper describes the leak-before-break (LBB) assessment procedure applicable to Japan Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr–1Mo steel. For the sodium pipes of JSFR, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. Firstly, a LBB assessment flowchart eliminating uncertainty resulted from small scale leakage, such as self plugging phenomenon and influence of crack surface roughness on leak rate, was proposed. Secondly, a rational unstable fracture assessment technique, taking the compliance changing with crack extension into account, was also proposed. Thirdly, a crack opening displacement (COD) assessment technique was developed, because COD assessment method applicable to JSFR pipes – thin wall and small work hardening material – had not been proposed yet. In addition, fracture toughness tests were performed using compact tension (CT) specimens to obtain the fracture toughness, J IC , and the crack growth resistance (J–R) curve at elevated temperature. Finally, by using the flowchart, proposed techniques and collected data, LBB assessment for the primary sodium pipes of JSFR was conducted. As a result, LBB aspect was successfully demonstrated with sufficient margins

  4. Challenges and innovative technologies on fuel handling systems for future sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Chassignet, Mathieu; Dumas, Sebastien; Penigot, Christophe; Prele, Gerard; Capitaine, Alain; Rodriguez, Gilles; Sanseigne, Emmanuel; Beauchamp, Francois

    2011-01-01

    The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into account various items and in particular operating strategies such as core design and management and core configuration. Moreover, the FHS will have to cope with safety assessments: a permanent cooling strategy to prevent fuel clad rupture, plus provisions to handle short-cooled fuel and criteria to ensure safety during handling. In addition, the handling and elimination of residual sodium must be investigated; it implies specific cleaning treatment to prevent chemical risks such as corrosion or excess hydrogen production. The objective of this study is to identify the challenges of a SFR fuel handling system. It will then present the range of technical options incorporating innovative technologies under development to answer the GENERATION IV SFR requirements. (author)

  5. Water cooled metal optics for the Advanced Light Source

    International Nuclear Information System (INIS)

    McKinney, W.R.; Irick, S.C.; Lunt, D.L.J.

    1991-01-01

    The program for providing water cooled metal optics for the Advanced Light Source at Berkeley is reviewed with respect to fabrication and metrology of the surfaces. Materials choices, surface figure and smoothness specifications, and metrology systems for measuring the plated metal surfaces are discussed. Results from prototype mirrors and grating blanks will be presented, which show exceptionally low microroughness and mid-period error. We will briefly describe out improved version of the Long Trace Profiler, and its importance to out metrology program. We have completely redesigned the mechanical, optical and computational parts of the profiler system with the cooperation of Peter Takacs of Brookhaven, Continental Optical, and Baker Manufacturing. Most important is that one of our profilers is in use at the vendor to allow testing during fabrication. Metrology from the first water cooled mirror for an ALS beamline is presented as an example. The preplating processing and grinding and polishing were done by Tucson Optical. We will show significantly better surface microroughness on electroless nickel, over large areas, than has been reported previously

  6. Advanced Spectral Library (ASTRAL): Atomic Fluorescence in Cool, Evolved Stars

    Science.gov (United States)

    Carpenter, Ken G.; Nielsen, Krister E.; Kober, Gladys V.; Rau, Gioia

    2018-01-01

    The "Advanced Spectral Library (ASTRAL) Project: Cool Stars" (PI = T. Ayres) collected a definitive set of representative, high-resolution (R~46,000 in the FUV up to ~1700 Å, R~30,000 for 1700-2150 Å, and R~114,000 >2150 Å) and high signal/noise (S/N>100) UV spectra of eight F-M evolved cool stars. These extremely high-quality STIS UV echelle spectra are available from the HST archive and from the Univ. of Colorado (http://casa.colorado.edu/~ayres/ASTRAL/) and will enable investigations of a broad range of problems -- stellar, interstellar, and beyond -- for many years. In this paper, we extend our study of the very rich emission-line spectra of the four evolved K-M stars in the sample, Beta Gem (K0 IIIb), Gamma Dra (K5 III), Gamma Cru (M3.4 III), and Alpha Ori (M2 Iab), to study the atomic fluorescence processes operating in their outer atmospheres. We summarize the pumping transitions and fluorescent line products known on the basis of previous work (e.g. Carpenter 1988, etc.) and newly identified in our current, on-going analysis of these extraordinary ASTRAL STIS spectra.

  7. Cryogenically cooled monochromators for the Advanced Photon Source

    International Nuclear Information System (INIS)

    Mills, D.M.

    1996-01-01

    The use of cryogenically cooled monochromators looks to be a very promising possibility for the Advanced Photon Source. This position has recently been bolstered by several experiments performed on beamlines at the ESRF and CHESS. At the ESRF, several crystal geometries have been tested that were designed for high power densities (approx-gt 150 W/mm 2 ) and moderate total absorbed powers (<200 W). These geometries have proven to be very successful at handling these power parameters with measured strains on the arc-second level. The experiments performed at CHESS were focused on high total power (approx-gt 1000 W) but moderate power densities. As with the previously mentioned experiments, the crystals designed for this application performed superbly with no measurable broadening of the rocking curves on the arc-second level. These experiments will be summarized and, based on these results, the performance of cryogenic monochromators for the APS will be assessed. copyright 1996 American Institute of Physics

  8. Advanced gas cooled nuclear reactor materials evaluation and development program

    International Nuclear Information System (INIS)

    1977-01-01

    Results of work performed from January 1, 1977 through March 31, 1977 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed

  9. Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

    International Nuclear Information System (INIS)

    Pattinson, A.

    2003-01-01

    The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

  10. An evaluation of the fluid-elastic instability for Intermediate Heat Exchanger of Prototype Sodium-cooled fast Reactor

    International Nuclear Information System (INIS)

    Cho, Jaehun; Kim, Sungkyun; Koo, Gyeonghoi

    2014-01-01

    The sodium-cooled fast reactor (SFR) module consists of the vessel, containment vessel, head, rotating plug (RP), upper internal structure (UIS), intermediate heat exchanger (IHX), decay heat exchanger (DHX), primary pump, internal structure, internal components and reactor core. The IHXs transfer heat from the radioactive sodium coolant (primary sodium) in the primary heat transport system to the nonradioactive sodium coolant (secondary sodium) in the intermediate heat transport system. Each sodium flows like Fig. 1. Primary sodium flows inside of tube and secondary sodium flows outside. During transferring heat two sodium to sodium, the fluid-elastic instability is occurred among tube bundle by cross flow. Large amplitude vibration occurred by the fluid-elastic instability is caused such as crack and wear of tube. Thus it is important to decrease the fluid-elastic instability in terms of a safety. The purpose of this paper is to evaluate the fluid-elastic instability for tube bundle in the IHX following ASME code. This paper evaluated the fluid-elastic instability of tube bundle in the SFR IHX. According evaluation results, the fluid-elastic instability of IHX tube bundle is occurred. A installing an additional TSP under the upper tubesheet can decrease a probability of fluid-elastic instability. If a location of an additional TSP does not exceed tube length to become a 750 mm, tube bundle of IHX is safety from the fluid-elastic instability

  11. Studies of decay heat removal by natural convection using the SONACO sodium-cooled 37-pin bundle

    International Nuclear Information System (INIS)

    Wydler, P.; Dury, T.V.; Hudina, M.; Weissenfluh, T. von; Sigg, B.; Dutton, P.

    1986-01-01

    Natural convection measurements in an electrically heated sodium-cooled rod bundle are being performed with the aim of contributing to a better understanding of natural convection effects in subassemblies with stagnant sodium and providing data for code validation. Measurements include temperature distributions in the bundle for different cooling configurations which simulate heat transfer to the intersubassembly gap and neighbouring subassemblies and possible thermosyphonic interaction between a subassembly and the reactor plenum above. Conditions for which stable natural convection patterns exist are identified, and results are compared with predictions of different computer codes of the porous-medium type. (author)

  12. Basic concept of fuel safety design and assessment for sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Nakae, Nobuo; Baba, Toshikazu; Kamimura, Katsuichiro

    2013-03-01

    'Philosophy in Safety Evaluation of Fast Breeder Reactors' was published as a guideline for safety design and safety evaluation of Sodium-Cooled Fast Reactor in Japan. This guideline points out that cladding creep and swelling due to internal pressure should be taken into account since the fuel is used under high temperature and high burnup, and that fuel assembly deformation and the prevention from coolant channel blockage should be taken into account in viewpoints of nuclear and thermal hydraulic design. However, the requirements including their criteria and evaluation items are not described. Two other domestic guidelines related to core design are applied for fuel design of fast reactor, but the description is considered to not be enough to practically use. In addition, technical standard for nuclear fuel used in power reactors is also applied for fuel inspection. Therefore, the technical standard and guideline for fuel design and safety evaluation are considered to be very important issue for nuclear safety regulation. This document has been developed according to the following steps: The guidelines and the technical standards, which are prepared in foreign countries and international organization, were reviewed. The technical background concerning fuel design and safety evaluation for fast reactor was collected and summarized in the world wide scale. The basic concept of fuel safety design and assessment for sodium-cooled fast reactor was developed by considering a wide range of views of the specialists in Japan. In order to discuss the content with foreign specialists IAEA Consultancy Meetings have been held on January, 2011 and January, 2012. The participants of the meeting came from USA, UK, EC, India, China and South Korea. The specialists of IAEA and JNES were also joined. Although this document is prepared for application to 'Monju'(prototype LMFR), it may be applied to experimental, demonstration and commercial types of LMFR after revising it by taking

  13. Material System Engineering for Advanced Electrocaloric Cooling Technology

    Science.gov (United States)

    Qian, Xiaoshi

    Electrocaloric effect refers to the entropy change and/or temperature change in dielectrics caused by the electric field induced polarization change. Recent discovery of giant ECE provides an opportunity to realize highly efficient cooling devices for a broad range of applications ranging from household appliances to industrial applications, from large-scale building thermal management to micro-scale cooling devices. The advances of electrocaloric (EC) based cooling device prototypes suggest that highly efficient cooling devices with compact size are achievable, which could lead to revolution in next generation refrigeration technology. This dissertation focuses on both EC based materials and cooling devices with their recent advances that address practical issues. Based on better understandings in designing an EC device, several EC material systems are studied and improved to promote the performances of EC based cooling devices. In principle, applying an electric field to a dielectric would cause change of dipolar ordering states and thus a change of dipolar entropy. Giant ECE observed in ferroelectrics near ferroelectric-paraelectric (FE-PE) transition temperature is owing to the large dipolar orientation change, between random-oriented dipolar states in paraelectric phase and spontaneous-ordered dipolar states in ferroelectric phases, which is induced by external electric fields. Besides pursuing large ECE, studies on EC cooling devices indicated that EC materials are required to possess wide operational temperature window, in which large ECE can be maintained for efficient operations. Although giant ECE was first predicted in ferroelectric polymers, where the large effect exhibits near FEPE phase transition, the narrow operation temperature window poses obstacles for these normal ferroelectrics to be conveniently perform in wide range of applications. In this dissertation, we demonstrated that the normal ferroelectric polymers can be converted to relaxor

  14. Advanced Pumps and Cold Plates for Two-Phase Cooling Loops, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Advanced instruments used for earth science missions require improved cooling systems to remove heat from high power electronic components and maintain tight...

  15. Studies on plant dynamics of sodium-cooled fast breeder reactors - verification of a plant model

    International Nuclear Information System (INIS)

    Schubert, B.

    1988-01-01

    For the analysis of sodium-cooled FBR safety and dynamics theoretical models are used, which have to be verified. In this report the verification of the plant model SSC-L is conducted by the comparison of calculated data with measurements of the experimental reactors KNK II and RAPSODIE. For this the plant model is extended and adapted. In general only small differences between calculated and measured data are recognized. The results are used to improve and complete the plant model. The extensions of the plant model applicability are used for the calculation of a loss of heat sink transient with reactor scram, considering pipes as passive heat sinks. (orig./HP) With 69 figs., 10 tabs [de

  16. Assessment of the dry process fuel sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on 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 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 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.

  17. Assessment of the dry process fuel sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2004-04-01

    The feasibility of using dry-processed oxide fuel in a Sodium-cooled Fast Reactor (SFR) was analyzed for the equilibrium fuel cycle of two reference cores: Hybrid BN-600 benchmark core with a enlarged lattice pitch and modified BN-600 core. The dry process technology assumed in this study based on 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 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 self-sustainable breakeven core was achievable without blanket fuels when the fuel volume fraction was ∼50% and most of the fission products were removed

  18. Impact of nuclear data on sodium-cooled fast reactor calculations

    International Nuclear Information System (INIS)

    Aures, A.; Bostelmann, F.; Zwermann, W.; Velkov, K.

    2016-01-01

    Neutron transport and depletion calculations are performed in combination with various nuclear data libraries in order to assess the impact of nuclear data on safety-relevant parameters of sodium-cooled fast reactors. These calculations are supplemented by systematic uncertainty analyses with respect to nuclear data. Analysed quantities are the multiplication factor and nuclide densities as a function of burn-up and the Doppler and Na-void reactivity coefficients at begin of cycle. While ENDF/B-VII.0 / -VII.1 yield rather consistent results, larger discrepancies are observed between the JEFF libraries. While the newest evaluation, JEFF-3.2, agrees with the ENDF/B-VII libraries, the JEFF-3.1.2 library yields significant larger multiplication factors. (authors)

  19. Effect of Reflector Material on the Neutronic Characteristics of the Small Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sung Hwan; Baek, Min Ho; Yoo, Jae Woon; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    The sodium-cooled fast reactor (SFR) has been chosen as a candidate for the Gen-IV Nuclear Energy Systems Initiative due to the advantages in utilization of uranium resources and reduction of radioactive wastes. Recently, the uranium blanket concept is omitted for a purpose of the non-proliferation, hence the reflector material plays a more important role in reactor core design. Moreover, especially in the Korean prototype SFR, the initial core should startup with low-enriched uranium ({<=} 20 w/o) for 100 {approx} 150 MWe power. This restriction causes significant difficulties to achieve sufficient excess reactivity. Thus, in this paper, core characteristic studies of various reflector materials (HT9, BeO, MgO, and ZrH{sub 1.6}) are performed to enhance the initial core excess reactivity

  20. Linear programming optimization of nuclear energy strategy with sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Lee, Je Whan; Jeong, Yong Hoon; Chang, Yoon Il; Chang, Soon Heung

    2011-01-01

    Nuclear power has become an essential part of electricity generation to meet the continuous growth of electricity demand. A Sodium-cooled Fast Reactor (SFR) was developed to extend uranium resource utilization under a growing nuclear energy scenario while concomitantly providing a nuclear waste management solution. Key questions in this scenario are when to introduce SFRs and how many reactors should be introduced. In this study, a methodology using Linear Programming is employed in order to quantify an optimized growth pattern of a nuclear energy system comprising light water reactors and SFRs. The optimization involves tradeoffs between SFR capital cost premiums and the total system U3O8 price premiums. Optimum nuclear growth patterns for several scenarios are presented, as well as sensitivity analyses of important input parameters

  1. Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    International Nuclear Information System (INIS)

    Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook

    2016-01-01

    The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper

  2. Performance of the diffusion barrier in the metallic fuel in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Ryu, Ho Jin; Yang, Seong Woo; Lee, Byoung Oon; Oh, Seok Jin; Lee, Chan Bock; Hahn, Dohee

    2009-01-01

    The objectives in this study are to propose several kinds of barrier materials and to evaluate their performance to prevent a fuel-clad interaction situation between the metallic fuel and the clad material in the Sodium-cooled Fast Reactor (SFR). Metallic foil made from refractory element, electrodeposition of the Cr on the clad surface, and the vapor deposition of the Zr were used as the barrier layers. The diffusion couple test was performed at the temperature of 800degC for 25 hour. The results showed that considerable amount of reaction occurred at the specimen without barrier, whereas excellent performance was observed in that neither reaction nor inter-diffusion occurred in the case of metallic foil made of Cr or V. Electrodeposition was revealed to be excellent provided that optimum deposition condition can be found. Similar to the electro-deposition result, excellent performance observed in the case of vapor deposition condition. (author)

  3. Overall System Description and Safety Characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Directory of Open Access Journals (Sweden)

    Jaewoon Yoo

    2016-10-01

    Full Text Available The Prototype Gen IV sodium cooled fast reactor (PGSFR has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  4. Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

    International Nuclear Information System (INIS)

    Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

    2008-01-01

    This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities

  5. BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

    International Nuclear Information System (INIS)

    Hetrick, D.L.; Sowers, G.W.

    1978-06-01

    This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. A list of variable names and a listing for BRENDA are included as appendices

  6. Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

    2008-01-15

    This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities.

  7. Advanced applications of water cooled nuclear power plants

    International Nuclear Information System (INIS)

    2008-07-01

    By August 2007, there were 438 nuclear power plants (NPPs) in operation worldwide, with a total capacity of 371.7 GW(e). Further, 31 units, totaling 24.1 GW(e), were under construction. During 2006 nuclear power produced 2659.7 billion kWh of electricity, which was 15.2% of the world's total. The vast majority of these plants use water-cooled reactors. Based on information provided by its Member States, the IAEA projects that nuclear power will grow significantly, producing between 2760 and 2810 billion kWh annually by 2010, between 3120 and 3840 billion kWh annually by 2020, and between 3325 and 5040 billion kWh annually by 2030. There are several reasons for these rising expectations for nuclear power: - Nuclear power's lengthening experience and good performance: The industry now has more than 12 000 reactor years of experience, and the global average nuclear plant availability during 2006 reached 83%; - Growing energy needs: All forecasts project increases in world energy demand, especially as population and economic productivity grow. The strategies are country dependent, but usually involve a mix of energy sources; - Interest in advanced applications of nuclear energy, such as seawater desalination, steam for heavy oil recovery and heat and electricity for hydrogen production; - Environmental concerns and constraints: The Kyoto Protocol has been in force since February 2005, and for many countries (most OECD countries, the Russian Federation, the Baltics and some countries of the Former Soviet Union and Eastern Europe) greenhouse gas emission limits are imposed; - Security of energy supply is a national priority in essentially every country; and - Nuclear power is economically competitive and provides stability of electricity price. In the near term most new nuclear plants will be evolutionary water cooled reactors (Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs), often pursuing economies of scale. In the longer term, innovative designs that

  8. Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut [Joint Research Centre, EC-JRC, Westerduinweg 3, P.O. Box 2, NL-0 1755 ZG Petten (Netherlands)

    2006-07-01

    This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

  9. Comparison of lead and sodium-cooled reactors - Safety, fuel cycle performance and some economical aspects

    International Nuclear Information System (INIS)

    Carlsson, Johan; Tucek, Kamil; Wider, Hartmut

    2006-01-01

    This paper compares the Lead-cooled Fast Reactor (LFR) and the Sodium-cooled Fast Reactor (SFR) regarding different aspects of the coolant, safety and economics. A brief review of design and safety experience of an SFR (BN-600) and some safety philosophy of the most developed LFR (BREST) are presented as well. The pros and cons of the lead and the sodium coolants are discussed. This paper presents results concerning the coolant temperature evolution during three accident scenarios, i.e. Loss-Of- Flow (LOF), Loss-Of-Heat-Sink (LOHS), and Total-Loss-Of-Power (TLOP). It also studies possible moderators, like BeO and hydrides, for the core designs to have negative reactivity feedbacks and favorable reactivity swings. LFR seems to be able to accommodate more minor actinides than SFR at comparable coolant and Doppler feedbacks. We show that LFR can be designed both to breed and burn transuranics from LWRs. The hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. It is shown that the LFR can handle the LOF transient better than the SFR. This is due to the much lower pressure drop in the LFR core. The coolant outlet temperatures stabilize at 2050 K and 940 K for SFR and LFR, respectively when no feedbacks are considered. Investigations also concern the SFR's performance when the pitch-to-diameter was increased from 1.2 to 1.4. For the LOHS and TLOP accidents their temperature evolutions are milder for the LFR since lead has a 50% larger volumetric heat capacity. For the TLOP the core outlet temperature of the LFR peaks at 1080 K after 2 days. Regarding economics it appears easier to avoid an intermediate cycle in an LFR than an SFR. (authors)

  10. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jin Ha; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2016-05-15

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O{sub 2} and (U,TRU)O{sub 2} which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O{sub 2}, (Th,Pu)O{sub 2} and (Th,TRU)O{sub 2}, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  11. Comparison of Core Performance with Various Oxide fuels on Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Choi, Jin Ha; Kim, Myung Hyun

    2016-01-01

    The system is called Prototype GenIV Sodium-cooled Fast Reactor (PGSFR). Ultimate goal of PGSFR is test for capability of TRU transmutation. Purpose of this study is test for evaluation of in-core performance and TRU transmutation performance by applying various oxide fuel loaded TRU. Fuel type of reference core is changed to uranium-based oxide fuel. Oxide fuel has a lot of experience through fuel fabrication and reactor operation. This study performed by compared and analyzed a core performance of various oxide fuels. (U,Pu)O_2 and (U,TRU)O_2 which various oxide fuel types are selected as extreme case for comparison with core performance and transmutation capability of TRU isotopes. Thorium-based fuel is known that it has good performance for burner reactor due to low proliferation characteristic. To check the performance of TRU incineration for comparison with uranium-based fuel on prototype SFR, Thorium-based fuel, (Th,U)O_2, (Th,Pu)O_2 and (Th,TRU)O_2, is selected. Calculations of core performance for various oxide fuel are performed using the fast calculation tool, TRANSX / DANTSTS / REBUS-3. In this study, comparison of core performance and transmutation performance is conducted with various fuel types in a sodium-cooled fast reactor. Mixed oxide fuel with TRU can produce the energy with small amount of fissile material. However, the TRU fuel is confirmed to bring a potential decline of the safety parameters. In case of (Th,U)O2 fuel, the flux level in thermal neutron region becomes lower because of higher capture cross-section of Th-232 than U-238. However, Th-232 has difficulty in converting to TRU isotopes. Therefore, the TRU consumption mass is relatively high in mixed oxide fuel with thorium and TRU.

  12. A new concept of hydrogen production system for sodium cooled FBR

    International Nuclear Information System (INIS)

    Nakagiri, Toshio; Aoto, Kazumi; Hoshiya, Taiji

    2004-01-01

    A new thermo-chemical and electrolytic hybrid hydrogen production process (thermo-chemical and electrolytic Hybrid Hydrogen process in Lower Temperature range: HHLT) is newly proposed by the Japan Nuclear Cycle Development Institute (JNC) to realize the hydrogen production from water by using the heat generation of sodium cooled Fast Breeding Reactor (FBR). The HHLT process is based on the sulfuric acid (H 2 SO 4 ) synthesis and decomposition processes developed earlier (Westinghouse process), and sulfur trioxide (SO 3 ) decomposition process of HHLT is facilitated by electrolysis with ionic oxygen conductive solid electrolyte to reduce operating temperature 200degC-300degC lower than Westinghouse process. Decomposition processes of SO 3 were confirmed with the cell voltage lower than 0.5 V at 500degC-600degC using 8mol yttria stabilized zirconia (8molYSZ) solid electrolyte and platinum electrode. Therefore, total voltage required for HHLT is expected to be lower than 1.0 V, because the voltage required for sulfuric acid synthesis is about 0.5V. Thermal efficiency of HHLT based on chemical reactions was roughly estimated to be within the range of 35% to 55% under the influence of H 2 SO 4 concentration and heat recovery. These results show the possibility of development of a new hydrogen production process which needs low splitting voltage and has high efficiency at around 500degC, utilizing the heat generation of sodium cooled FBR. SO 3 splitting with the voltage lower than 0.5V was confirmed at about 500degC experimentally, and ideal thermal efficiency of the cycle based on chemical reactions was evaluated. Furthermore, test apparatus to substantiate whole process of HHLT was manufactured. (author)

  13. Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

    Energy Technology Data Exchange (ETDEWEB)

    Sun Kaichao, E-mail: kaichao.sun@psi.ch [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), 1015 Lausanne (Switzerland); Krepel, Jiri; Mikityuk, Konstantin; Pelloni, Sandro [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Chawla, Rakesh [Paul Scherrer Institut (PSI), 5232 Villigen PSI (Switzerland); Ecole Polytechnique Federale de Lausanne (EPFL), 1015 Lausanne (Switzerland)

    2011-07-15

    Highlights: > We analyze the void reactivity effect for three ESFR core fuel cycle states. > The void reactivity effect is decomposed by neutron balance method. > Novelly, the normalization to the integral flux in the active core is applied. > The decomposition is compared with the perturbation theory based results. > The mechanism and the differences of the void reactivity effect are explained. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case - the voiding of both inner and outer fuel zones - has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly by the

  14. Void reactivity decomposition for the Sodium-cooled Fast Reactor in equilibrium fuel cycle

    International Nuclear Information System (INIS)

    Sun Kaichao; Krepel, Jiri; Mikityuk, Konstantin; Pelloni, Sandro; Chawla, Rakesh

    2011-01-01

    Highlights: → We analyze the void reactivity effect for three ESFR core fuel cycle states. → The void reactivity effect is decomposed by neutron balance method. → Novelly, the normalization to the integral flux in the active core is applied. → The decomposition is compared with the perturbation theory based results. → The mechanism and the differences of the void reactivity effect are explained. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many advantages, but has one dominating neutronic drawback - a positive sodium void reactivity. The aim of this study is to develop and apply a methodology, which should help better understand the causes and consequences of the sodium void effect. It focuses not only on the beginning-of-life (BOL) state of the core, but also on the beginning of open and closed equilibrium (BOC and BEC, respectively) fuel cycle conditions. The deeper understanding of the principal phenomena involved may subsequently lead to appropriate optimization studies. Various voiding scenarios, corresponding to different spatial zones, e.g. node or assembly, have been analyzed, and the most conservative case - the voiding of both inner and outer fuel zones - has been selected as the reference scenario. On the basis of the neutron balance method, the corresponding SFR void reactivity has been decomposed reaction-, isotope-, and energy-group-wise. Complementary results, based on generalized perturbation theory and sensitivity analysis, are also presented. The numerical analysis for both neutron balance and perturbation theory methods has been carried out using appropriate modules of the ERANOS code system. A strong correlation between the flux worth, i.e. the product of flux and adjoint flux, and the void reactivity importance distributions has been found for the node- and assembly-wise voiding scenarios. The neutron balance based decomposition has shown that the void effect is caused mainly

  15. Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, Liancheng; Zhang, Bin

    2014-01-01

    The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

  16. Level-1 PSA to support the design of the KALIMER-600 Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Han, Sang Hoon; Kim, Tae-Woon; Jeong, Hae-Yong; Han, Seok Joong; Ahn, Kwang-Il; Yang, Joon-Eon

    2012-01-01

    A sodium-cooled fast reactor, KALIMER-600, is under development. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as a coolant. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing KALIMER-600 from the aspects of risk informed design. A preliminary level-1 internal full power PSA has been performed to evaluate the safety level and its applicability for the KALIMER-600 conceptual design. Various design alternatives are evaluated from the viewpoint of PSA in order to support the design of the KALIMER-600. Sensitivity studies are also performed to evaluate the assumptions made for the PSA. The applicability and weakness of the KALIMER-600 PSA are discussed. The technical issues to be solved in performing the PSA will be discussed. (authors)

  17. Minor actinides transmutation potential: state of art for GEN IV sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Buiron, Laurent

    2015-01-01

    In the frame of the R and D program relative to the 1991 French act on nuclear waste management, fast neutron systems have shown relevant characteristics that meet both requirements on sustainable resources management and waste minimization. They also offer flexibility by mean of burner or breeder configurations allowing mastering plutonium inventory without significant impact on core safety. From the technological point of view, sodium cooled fast reactor are considered in order to achieve mean term industrial deployment. The present document summaries the main results of R and D program on minor actinides transmutation in sodium fast reactor since 2006 following recommendation of the first part of the 1991 French act. Both homogeneous and heterogeneous management achievable performances are presented for 'evolutionary' SFR V2B core as well as low void worth CFV core for industrial scale configurations (1500 MWe). Minor actinides transmutation could be demonstrated in the ASTRID reactor with the following configurations: - a 2%vol Americium content for the homogeneous mode, - a 10%vol Americium content for the heterogeneous mode, without any substantial modification of the main core safety parameters and only limited impacts on the associated fuel cycle (manufacturing issues are not considered here). In order to achieve such goal, a wide range of experimental irradiations driven by transmutation scenarios have to be performed for both homogeneous and heterogeneous minor actinides management. (author) [fr

  18. Study of various Brayton cycle designs for small modular sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Ahn, Yoonhan; Lee, Jeong Ik

    2014-01-01

    Highlights: • Application of closed Brayton cycle for small and medium sized SFRs is reviewed. • S-CO 2 , helium and nitrogen cycle designs for small modular SFR applications are analyzed and compared in terms of cycle efficiency, component performance and physical size. • Several new layouts for each Brayton cycle are suggested to simplify the turbomachinery designs. • S-CO 2 cycle design shows the best efficiency and compact size compared to other Brayton cycles. - Abstract: Many previous sodium cooled fast reactors (SFRs) adopted steam Rankine cycle as the power conversion system. However, the concern of sodium water reaction has been one of the major design issues of a SFR system. As an alternative to the steam Rankine cycle, several closed Brayton cycles including supercritical CO 2 cycle, helium cycle and nitrogen cycle have been suggested recently. In this paper, these alternative gas Brayton cycles will be compared to each other in terms of cycle performance and physical size for small modular SFR application. Several new layouts are suggested for each fluid while considering the turbomachinery design and the total system volume

  19. Development of Preliminary HT9 Cladding Tube for Sodium-cooled Fast Reactor (SFR)

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Baek, Jong Hyuk; Heo, Hyeong Min; Park, Sang Gyu; Kim, Sung Ho; Lee, Chan Bock

    2013-01-01

    To achieve manufacturing technology of the fuel cladding tube in order to keep pace with the predetermined schedule in developing SFR fuel, KAERI has launched in developing fuel cladding tube in cooperation with a domestic steelmaking company. After fabricating medium-sized 1.1 ton HT9 ingot, followed by the multiple processes of hot and cold working, preliminary samples of HT9 seamless cladding tube having 7.4mm in outer diameter, 0.56mm in thickness, and 3m in length were fabricated. The objective of this study is to summarize the brief development status of the HT9 cladding tubes. Mechanical properties like axial tension, biaxial burst, pressurized creep and sodium compatibility of the cladding tubes were carried out to set up the performance evaluation technology to test the prototype FMS cladding tube which is going to be manufactured in next stage. As a part of developing fuel cladding for the Sodium-cooled Fast Reactor (SFR), preliminary HT9 cladding tube was fabricated in cooperation with a domestic steelmaking company. Microstructure as well as mechanical tests like axial tensile test, biaxial burst test, and pressurized creep test of the fuel cladding were carried out. Performance of the domestic HT9 tube was revealed to be similar in the previously fabricated foreign HT9 tube. Further prototype FMS cladding tube is going to be manufactured in next year based on this experience. Various test items like mechanical test, sodium compatibility test, microstructural analysis, basic property, cladding performance under transient situation, and performance under ion and neutron irradiation are going be performed in the future to set up the relevant technology for the licensing of the SFR cladding tube

  20. 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

  1. Unprotected Accident Analyses of the 1200MWe GEN-IV Sodium-Cooled Fast Reactor Using the SSC-K Code

    International Nuclear Information System (INIS)

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

    2010-02-01

    A conceptual design of an advanced breakeven sodium-cooled fast reactor (G4SFR) has recently been developed by KAERI under the national nuclear R and D plan. The G4SFR is a 1,200MWe metal-fueled pool-type sodium-cooled fast reactor adopting advanced safety design features. The G4SFR development plan focuses on particular technology development efforts to effectively meet the goals of the Generation-IV (GEN-IV) nuclear system such as efficient utilization of resources, economic competitiveness, a high standard of safety, and enhanced proliferation resistance. To enhance the safety of G4SFR, advanced design features of metal-fueled core, simple and large sodium-inventory primary heat transport system, and passive safety decay heat removal system are included in the reactor design. To evaluate potential safety characteristics of such advanced design features, the plant responses and safety margins were investigated using the system transient code SSC-K for three unprotected accidents of UTOP, ULOF, and ULOHS. It was shown that the G4SFR design has inherent and passive safety characteristics and is accommodating the selected ATWS events. The inherent safety mechanism of the reactor design makes the core shutdown with sufficient margin and passive removal of decay heat with matching the core power to heat sink by passive self-regulation. The self-regulation of power without scram is mainly due to the inherent negative reactivity feedback in conjunction with the large thermal inertia of the primary heat transport system and the passive decay heat removal. Such favorable inherent and passive safety behaviors of G4SFR are expected to virtually exclude the probability of severe accidents with potential for core damage

  2. The effect of steam cycle conditions upon the economics and design of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Philpott, E.F.; Pounder, F.; Willby, C.R.

    1978-01-01

    The paper studies the effect of variation of steam and feedwater conditions upon the economics, design and layout of a sodium-cooled fast reactor. The parameters investigated are steam temperature and pressure, feedwater temperature, and boiler recirculation ratio. The paper also includes an assessment of the effects of associating the fast reactor with saturated steam cycle conditions. (author)

  3. Review of aerosol problems and the theory of aerosol physics with particular reference to sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Williams, R.J.

    1978-01-01

    Processes that would govern the development, transport, and removal of aerosols, which are of interest in the study of hypothetical core disruptive situations in pool type sodium cooled fast reactors, are discussed. Theoretical descriptions of these processes are presented and known inadequacies indicated. The interpretation of experimental data and numeric solution of the governing equations is briefly considered. (author)

  4. Study on In-Service Inspection Program and Inspection Technologies for Commercialized Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Masato Ando; Shigenobu Kubo; Yoshio Kamishima; Toru Iitsuka

    2006-01-01

    The objective of in-service inspection of a nuclear power plant is to confirm integrity of function of components necessary to safety, and satisfy the needs to protect plant investment and to achieve high plant ability. The sodium-cooled fast reactor, which is designed in the feasibility study on commercialized fast reactor cycle systems in Japan, has two characteristics related to in-service inspection. The first is that all sodium coolant boundary structures have double-wall system. Continuous monitoring of the sodium coolant boundary structures are adopted for inspection. The second characteristic is the steam generator with double-wall-tubes. Volumetric testing is adopted to make sure that one of the tubes can maintain the boundary function in case of the other tube failure. A rational in-service inspection concept was developed taking these features into account. The inspection technologies were developed to implement in-service inspection plan. The under-sodium viewing system consisted of multi ultrasonic scanning transducers, which was used for imaging under-sodium structures. The under-sodium viewing system was mounted on the under-sodium vehicle and delivered to core internals. The prototype of under-sodium viewing system and vehicle were fabricated and performance tests were carried out under water. The laboratory experiments of volumetric testing for double-wall-tubes of steam generator, such as ultrasonic testing and remote-field eddy current testing, were performed and technical feasibility was assessed. (authors)

  5. The development of advanced gas cooled reactor iodine adsorber systems

    International Nuclear Information System (INIS)

    Meddings, P.

    1986-01-01

    Advanced Gas Cooled Reactors (AGRs) are provided with plants to process the carbon dioxide coolant prior to its discharge to atmosphere. Included in these are beds of granular activated charcoal, contained within a suitable pressure vessel, through which the high pressure carbon dioxide is passed for the purpose of retaining iodine and iodine-containing compounds. Carry-over carbon dust from the adsorption beds was identified during active in-situ commissioning testing, radio-iodine being transported with the particulate material due to gross disturbance of the adsorber carbon bed and displacement of the vessel internals. The methods used to identify the causes of the problems and find solutions are described. A development programme for the Heysham-2 and Torness reactors iodine adsorber units was set up to identify a method of de-dusting granular charcoal and develop it for full-scale use, of assess the effect under conditions of high gas density of approach velocity on charcoal fines production and to establish the pressure drop characteristics of a packed granular bed and to develop an effective design of inlet gas diffuser manifold to ensure an acceptable velocity distribution. This has involved the construction of a small scale high pressure carbon dioxide rig and development of an air flow model. This work is described. (UK)

  6. Description of the advanced gas cooled type of reactor (AGR)

    Energy Technology Data Exchange (ETDEWEB)

    Nonboel, E. [Risoe National Lab., Roskilde (Denmark)

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: `Reactors in Nordic Surroundings`, which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs.

  7. Description of the advanced gas cooled type of reactor (AGR)

    International Nuclear Information System (INIS)

    Nonboel, E.

    1996-11-01

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR plant on the Lothian coastline of Scotland, 60 km east of Edinburgh, has been chosen as the reference plant and is described in some detail. Data on the other 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

  8. Identification of important phenomena under sodium fire accidents based on PIRT process with factor analysis in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Aoyagi, Mitsuhiro; Uchibori, Akihiro; Kikuchi, Shin; Takata, Takashi; Ohno, Shuji; Ohshima, Hiroyuki

    2016-01-01

    The PIRT (Phenomena Identification and Ranking Table) process is an effective method to identify key phenomena involved in safety issues in nuclear power plants. The present PIRT process is aimed to validate sodium fire analysis codes. Because a sodium fire accident in sodium-cooled fast reactor (SFR) involves complex phenomena, various figures of merit (FOMs) could exist in this PIRT process. In addition, importance evaluation of phenomena for each FOM should be implemented in an objective manner under the PIRT process. This paper describes the methodology for specification of FOMs, identification of associated phenomena and importance evaluation of each associated phenomenon in order to complete a ranking table of important phenomena involved in a sodium fire accident in an SFR. The FOMs were specified through factor analysis in this PIRT process. Physical parameters to be quantified by a sodium fire analysis code were identified by considering concerns resulting from sodium fire in the factor analysis. Associated phenomena were identified through the element- and sequence-based phenomena analyses as is often conducted in PIRT processes. Importance of each associated phenomenon was evaluated by considering the sequence-based analysis of associated phenomena correlated with the FOMs. Then, we complete the ranking table through the factor and phenomenon analyses. (author)

  9. Experimental facilities for gas-cooled reactor safety studies. Task group on Advanced Reactor Experimental Facilities (TAREF)

    International Nuclear Information System (INIS)

    2009-01-01

    In 2007, the NEA Committee on the Safety of Nuclear Installations (CSNI) completed a study on Nuclear Safety Research in OECD Countries: Support Facilities for Existing and Advanced Reactors (SFEAR) which focused on facilities suitable for current and advanced water reactor systems. In a subsequent collective opinion on the subject, the CSNI recommended to conduct a similar exercise for Generation IV reactor designs, aiming to develop a strategy for ' better preparing the CSNI to play a role in the planned extension of safety research beyond the needs set by current operating reactors'. In that context, the CSNI established the Task Group on Advanced Reactor Experimental Facilities (TAREF) in 2008 with the objective of providing an overview of facilities suitable for performing safety research relevant to gas-cooled reactors and sodium fast reactors. This report addresses gas-cooled reactors; a similar report covering sodium fast reactors is under preparation. The findings of the TAREF are expected to trigger internationally funded CSNI projects on relevant safety issues at the key facilities identified. Such CSNI-sponsored projects constitute a means for efficiently obtaining the necessary data through internationally co-ordinated research. This report provides an overview of experimental facilities that can be used to carry out nuclear safety research for gas-cooled reactors and identifies priorities for organizing international co-operative programmes at selected facilities. The information has been collected and analysed by a Task Group on Advanced Reactor Experimental Facilities (TAREF) as part of an ongoing initiative of the NEA Committee on the Safety of Nuclear Installations (CSNI) which aims to define and to implement a strategy for the efficient utilisation of facilities and resources for Generation IV reactor systems. (author)

  10. Minor actinide transmutation in a board type sodium cooled breed and burn reactor core

    International Nuclear Information System (INIS)

    Zheng, Meiyin; Tian, Wenxi; Zhang, Dalin; Qiu, Suizheng; Su, Guanghui

    2015-01-01

    Highlights: • A 1250 MWt board type sodium cooled breed and burn reactor core is further designed. • MCNP–ORIGEN coupled code MCORE is applied to perform neutronics and depletion calculation. • Transmutation efficiency and neutronic safety parameters are compared under different MA weight fraction. - Abstract: In this paper, a board type sodium cooled breed and burn reactor core is further designed and applied to perform minor actinide (MA) transmutation. MA is homogeneously loaded in all the fuel sub-assemblies with a weight fraction of 2.0 wt.%, 4.0 wt.%, 6.0 wt.%, 8.0 wt.%, 10.0 wt.% and 12.0 wt.%, respectively. The transmutation efficiency, transmutation amount, power density distribution, neutron fluence distribution and neutronic safety parameters, such as reactivity, Doppler feedback, void worth and delayed neutron fraction, are compared under different MA weight fraction. Neutronics and depletion calculations are performed based on the self-developed MCNP–ORIGEN coupled code with the ENDF/B-VII data library. In the breed and burn reactor core, a number of breeding sub-assemblies are arranged in the inner core in a board type way (scatter load) to breed, and a number of absorbing sub-assemblies are arranged in the inner side of the outer core to absorb neutrons and reduce power density in this area. All the fuel sub-assemblies (ignition and breeding sub-assemblies) are shuffled from outside in. The core reached asymptotically steady state after about 22 years, and the average and maximum discharged burn-up were about 17.0% and 35.3%, respectively. The transmutation amount increased linearly with the MA weight fraction, while the transmutation rate parabolically varied with the MA weight fraction. Power density in ignition sub-assembly positions increased with the MA weight fraction, while decreased in breeding sub-assembly positions. Neutron fluence decreased with the increase of MA weight fraction. Generally speaking, the core reactivity and void

  11. Fertile assembly for a fast neutron nuclear reactor cooled by liquid sodium, with regulation of the cooling rate

    International Nuclear Information System (INIS)

    Pradal, L.; Berte, M.; Chiarelli, C.

    1985-01-01

    The assembly has a casing in which are arranged the fertile elements, the liquid sodium flowing through the casing along these elements. It includes several apertured diaphragms transverse to the rods to regulate the liquid sodium flow rate. At least one diaphragm, in its central part around its aperture, of a material soluble in liquid sodium, such as copper. The invention applies, more particularly, to fast neutron nuclear reactor having a heterogeneous core. The coolant flow can increase with time to match the increased power generated by the fertile assembly along its life [fr

  12. Work Domain Analysis of a Predecessor Sodium-cooled Reactor as Baseline for AdvSMR Operational Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Ronald Farris; David Gertman; Jacques Hugo

    2014-03-01

    This report presents the results of the Work Domain Analysis for the Experimental Breeder Reactor (EBR-II). This is part of the phase of the research designed to incorporate Cognitive Work Analysis in the development of a framework for the formalization of an Operational Concept (OpsCon) for Advanced Small Modular Reactors (AdvSMRs). For a new AdvSMR design, information obtained through Cognitive Work Analysis, combined with human performance criteria, can and should be used in during the operational phase of a plant to assess the crew performance aspects associated with identified AdvSMR operational concepts. The main objective of this phase was to develop an analytical and descriptive framework that will help systems and human factors engineers to understand the design and operational requirements of the emerging generation of small, advanced, multi-modular reactors. Using EBR-II as a predecessor to emerging sodium-cooled reactor designs required the application of a method suitable to the structured and systematic analysis of the plant to assist in identifying key features of the work associated with it and to clarify the operational and other constraints. The analysis included the identification and description of operating scenarios that were considered characteristic of this type of nuclear power plant. This is an invaluable aspect of Operational Concept development since it typically reveals aspects of future plant configurations that will have an impact on operations. These include, for example, the effect of core design, different coolants, reactor-to-power conversion unit ratios, modular plant layout, modular versus central control rooms, plant siting, and many more. Multi-modular plants in particular are expected to have a significant impact on overall OpsCon in general, and human performance in particular. To support unconventional modes of operation, the modern control room of a multi-module plant would typically require advanced HSIs that would

  13. Reflector and Protections in a Sodium-cooled Fast Reactor: Modelling and Optimization

    Science.gov (United States)

    Blanchet, David; Fontaine, Bruno

    2017-09-01

    The ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration) is a Generation IV nuclear reactor concept under development in France [1]. In this frame, studies are underway to optimize radial reflectors and protections. Considering radial protections made in natural boron carbide, this study is conducted to assess the neutronic performances of the MgO as the reference choice for reflector material, in comparison with other possible materials including a more conventional stainless steel. The analysis is based upon a simplified 1-D and 2-D deterministic modelling of the reactor, providing simplified interfaces between core, reflector and protections. Such models allow examining detailed reaction rate distributions; they also provide physical insights into local spectral effects occurring at the Core-Reflector and at the Reflector-Protection interfaces.

  14. Impact of reducing sodium void worth on the severe accident response of metallic-fueled sodium-cooled reactors

    International Nuclear Information System (INIS)

    Wigeland, R.A.; Turski, R.B.; Pizzica, P.A.

    1994-01-01

    Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept

  15. Sodium-cooled fast reactor (SFR) fuel assembly design with graphite-moderating rods to reduce the sodium void reactivity coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Won, Jong Hyuck; Cho, Nam Zin, E-mail: nzcho@kaist.ac.kr; Park, Hae Min; Jeong, Yong Hoon, E-mail: jeongyh@kaist.ac.kr

    2014-12-15

    Highlights: • The graphite rod-inserted SFR fuel assembly is proposed to achieve low sodium void reactivity. • The neutronics/thermal-hydraulics analyses are performed for the proposed SFR cores. • The sodium void reactivity is improved about 960–1030 pcm compared to reference design. - Abstract: The concept of a graphite-moderating rod-inserted sodium-cooled fast reactor (SFR) fuel assembly is proposed in this study to achieve a low sodium void reactivity coefficient. Using this concept, two types of SFR cores are analyzed; the proposed SFR type 1 core has new SFR fuel assemblies at the inner/mid core regions while the proposed SFR type 2 core has a B{sub 4}C absorber sandwich in the middle of the active core region as well as new SFR fuel assemblies at the inner/mid core regions. For the proposed SFR core designs, neutronics and thermal-hydraulic analyses are performed using the DIF3D, REBUS3, and the MATRA-LMR codes. In the neutronics analysis, the sodium void reactivity coefficient is obtained in various void situations. The two types of proposed core designs reduce the sodium void reactivity coefficient by about 960–1030 pcm compared to the reference design. However, the TRU enrichment for the proposed SFR core designs is increased. In the thermal hydraulic analysis, the temperature distributions are calculated for the two types of proposed core designs and the mass flow rate is optimized to satisfy the design constraints for the highest power generating assembly. The results of this study indicate that the proposed SFR assembly design concept, which adopts graphite-moderating rods which are inserted into the fuel assembly, can feasibly minimize the sodium void reactivity coefficient. Single TRU enrichment and an identical fuel slug diameter throughout the SFR core are also achieved because the radial power peak can be flattened by varying the number of moderating rods in each core region.

  16. Objective Provision Trees of Reactivity Control Safety Function for Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kang, Bongsuk; Yang, Huichang; Suh, Namduk

    2014-01-01

    The purpose of this OPT is first to assure the DiD design during the licensing of Sf, but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). Based on the definition of Defense-in-Depth (DiD) levels and safety functions for KALIMER Sodium-Cooled Fast Reactor (SFR), suggested in the reference and, Objective Provision Trees (OPTs) of reactivity control function for level 1, 2, 3 and 4 DiD were developed and suggested in this paper. The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of Prototype Gen-IV Sf (PGSFR) is not mature yet, the OPT is developed for KALIMER design. Developed level 1 to 4 OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defense-in-depth evaluation frame for the regulatory reviews for the licensing process. In the next stage of this study, other safety function will be researched and findings can be suggested as recommendations for the safety improvement

  17. Status of conceptual safety design study of Japanese sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kubo, Shigenobu; Kurisaka, Kenichi; Niwa, Hajime; Shimakawa, Yoshio

    2005-01-01

    In this paper, the current conceptual safety design and related evaluation of Japanese Sodium-cooled Fast Reactor which is studied in the framework of the Feasibility Study (FS) on commercialized Fast Reactor Cycle Systems in Japan are described. The purpose of the safety design is to establish a feasible safety concept of FBR which aims at a sustainable energy source of the next generations. The safety targets and the safety design principle are set aiming at realizing worldwide acceptability of the safety level. The basic safety design concept, which can meet the safety targets, was formulated taking along with the defense-in-depth philosophy as the basic safety design principle. In order to cope with wide range of energy and resource demands, there are some various designs both of oxide and metal fuel for JSFR. Some analytical results of typical design basis events, design extension conditions and core damage frequency estimation show the feasibility of the safety design concept for them. (author)

  18. Compact sodium cooled nuclear power plant with fast core (KNK II- Karlsruhe), Safety Report

    International Nuclear Information System (INIS)

    1977-09-01

    After the operation of the KNK plant with a thermal core (KNK I), the installation of a fast core (KNK II) had been realized. The planning of the core and the necessary reconstruction work was done by INTERATOM. Owner and customer was the Nuclear Research Center Karlsruhe (KfK), while the operating company was the Kernkraftwerk-Betriebsgesellschaft mbH (KBG) Karlsruhe. The main goals of the KNK II project and its special experimental test program were to gather experience for the construction, the licensing and operation of future larger plants, to develop and to test fuel and absorber assemblies and to further develop the sodium technology and the associated components. The present safety report consists of three parts. Part 1 contains the description of the nuclear plant. Hereby, the reactor and its components, the handling facilities, the instrumentation with the plant protection, the design of the plant including the reactor core and the nominal operation processes are described. Part 2 contains the safety related investigation and measures. This concerns the reactivity accidents, local cooling perturbations, radiological consequences with the surveillance measures and the justification of the choice of structural materials. Part three finally is the appendix with the figures, showing the different buildings, the reactor and its components, the heat transfer systems and the different auxiliary facilities [de

  19. Objective Provision Trees of Reactivity Control Safety Function for Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Bongsuk; Yang, Huichang [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    The purpose of this OPT is first to assure the DiD design during the licensing of Sf, but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). Based on the definition of Defense-in-Depth (DiD) levels and safety functions for KALIMER Sodium-Cooled Fast Reactor (SFR), suggested in the reference and, Objective Provision Trees (OPTs) of reactivity control function for level 1, 2, 3 and 4 DiD were developed and suggested in this paper. The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of Prototype Gen-IV Sf (PGSFR) is not mature yet, the OPT is developed for KALIMER design. Developed level 1 to 4 OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defense-in-depth evaluation frame for the regulatory reviews for the licensing process. In the next stage of this study, other safety function will be researched and findings can be suggested as recommendations for the safety improvement.

  20. Uncertainty analysis of infinite homogeneous lead and sodium cooled fast reactors at beginning of life

    International Nuclear Information System (INIS)

    Vanhanen, R.

    2015-01-01

    The objective of the present work is to estimate breeding ratio, radiation damage rate and minor actinide transmutation rate of infinite homogeneous lead and sodium cooled fast reactors. Uncertainty analysis is performed taking into account uncertainty in nuclear data and composition of the reactors. We use the recently released ENDF/B-VII.1 nuclear data library and restrict the work to the beginning of reactor life. We work under multigroup approximation. The Bondarenko method is used to acquire effective cross sections for the homogeneous reactor. Modeling error and numerical error are estimated. The adjoint sensitivity analysis is performed to calculate generalized adjoint fluxes for the responses. The generalized adjoint fluxes are used to calculate first order sensitivities of the responses to model parameters. The acquired sensitivities are used to propagate uncertainties in the input data to find out uncertainties in the responses. We show that the uncertainty in model parameters is the dominant source of uncertainty, followed by modeling error, input data precision and numerical error. The uncertainty due to composition of the reactor is low. We identify main sources of uncertainty and note that the low-fidelity evaluation of 16 O is problematic due to lack of correlation between total and elastic reactions

  1. Safety approach and R and D program for future french sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Beils, Stephane; Carluec, Bernard; Devictor, Nicolas; Fiorini, Gian Luigi; Sauvage, Jean Francois

    2011-01-01

    This paper presents briefly the safety approach as well as the R and D program that is underway to support the deployment of future French Sodium-Cooled fast Reactors (SFRs): A) Safety objectives and principles for future reactors. The content of the first section reflects the works of AREVA, CEA, and EDF concerning the safety orientations for the future reactors. The availability of such orientations and requirements for the SFRs has to allow introducing and managing the process that will lead to the detailed definition of the safety approach, to the selection of the corresponding safety options, and to the identification and motivation of the supporting R and D. B) Strategy and roadmap in support of the R and D for future SFRs. This section describes the R and D program led jointly by CEA, EDF, and AREVA, which has been developed with the objectives to be able to preliminarily define, by 2012, the safety orientations for the future SFRs, and to deduce from them the characteristics of the ASTRID prototype. (author)

  2. Fabrication of uranium alloy fuel slug for sodium-cooled fast reactor by injection casting

    International Nuclear Information System (INIS)

    Jong Hwan Kim; Hoon Song; Ki Hwan Kim; Chan Bock Lee

    2014-01-01

    Metal fuel slugs of U-Zr alloys for a sodium-cooled fast reactor (SFR) have been fabricated using an injection casting method. However, casting alloys containing volatile radioactive constituents such as Am can cause problems in a conventional injection casting method. Therefore, in this study, several injection-casting methods were applied to evaluate the volatility of the metal-fuel elements and control the transport of volatile elements. Mn was selected as a volatile surrogate alloy since it possesses a total vapor pressure equivalent to that of minor actinide-bearing fuels for SFRs. U-10 wt% Zr and U-10 wt% Zr-5 wt% Mn metal fuels were prepared, and the casting processes were evaluated. The casting soundness of the fuel slugs was characterized by gamma-ray radiography and immersion density measurements. Inductively coupled plasma atomic emission spectroscopy was used to determine the chemical composition of fuel slugs. Fuel losses after casting were also evaluated according to the casting conditions. (author)

  3. Definition of a Robust Supervisory Control Scheme for Sodium-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ponciroli, R.; Passerini, S.; Vilim, R. B.

    2016-04-17

    In this work, an innovative control approach for metal-fueled Sodium-cooled Fast Reactors is proposed. With respect to the classical approach adopted for base-load Nuclear Power Plants, an alternative control strategy for operating the reactor at different power levels by respecting the system physical constraints is presented. In order to achieve a higher operational flexibility along with ensuring that the implemented control loops do not influence the system inherent passive safety features, a dedicated supervisory control scheme for the dynamic definition of the corresponding set-points to be supplied to the PID controllers is designed. In particular, the traditional approach based on the adoption of tabulated lookup tables for the set-point definition is found not to be robust enough when failures of the implemented SISO (Single Input Single Output) actuators occur. Therefore, a feedback algorithm based on the Reference Governor approach, which allows for the optimization of reference signals according to the system operating conditions, is proposed.

  4. Uncertainty analysis of infinite homogeneous lead and sodium cooled fast reactors at beginning of life

    Energy Technology Data Exchange (ETDEWEB)

    Vanhanen, R., E-mail: risto.vanhanen@aalto.fi

    2015-03-15

    The objective of the present work is to estimate breeding ratio, radiation damage rate and minor actinide transmutation rate of infinite homogeneous lead and sodium cooled fast reactors. Uncertainty analysis is performed taking into account uncertainty in nuclear data and composition of the reactors. We use the recently released ENDF/B-VII.1 nuclear data library and restrict the work to the beginning of reactor life. We work under multigroup approximation. The Bondarenko method is used to acquire effective cross sections for the homogeneous reactor. Modeling error and numerical error are estimated. The adjoint sensitivity analysis is performed to calculate generalized adjoint fluxes for the responses. The generalized adjoint fluxes are used to calculate first order sensitivities of the responses to model parameters. The acquired sensitivities are used to propagate uncertainties in the input data to find out uncertainties in the responses. We show that the uncertainty in model parameters is the dominant source of uncertainty, followed by modeling error, input data precision and numerical error. The uncertainty due to composition of the reactor is low. We identify main sources of uncertainty and note that the low-fidelity evaluation of {sup 16}O is problematic due to lack of correlation between total and elastic reactions.

  5. Conceptual Design for BOP of the Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yoo, Tae Geun; Kim, Seong O; Kim, Eui Kwang; Seong, Seung Hwan

    2010-01-01

    The heavy dependence on nuclear power eventually raise the issues of an efficient utilization of uranium resources, which Korea presently imports from abroad, end of a spent fuel storage. From the viewpoint that sodium-cooled fast Reactors (SFR s ) have the potential of an enhanced safety by utilizing inherent safety characteristics, trans-uranics (TRU) reduction and resolving the spent fuel storage problems through a proliferation-resistant actinide recycling. SFR s are sure to be most promising nuclear power operation. The Korea Atomic Energy Research Institute (KAERI) has been developing SFR design technologies since 1997. And nowadays, the preliminary heat balance of the demonstration SFR is calculated. However, in order to verify design condition of the NSSS, it is necessary to set the heat balance and the conceptual design for BOP of the SFR as a part of the SFR design technique development business. Moreover, in order to confirm whether the heat balance can actually appropriate via the turbine characteristic, it is required to carry out the performance analysis of the turbine cycle. For that, the main purposes of this study are; 1) to derivate the conceptual design for BOP, 2) to analyze the performance of the turbine cycle, 3) to derivate the main consideration for BOP design

  6. Large electro-magnetic pump design for application in the ASTRID sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Laffont, Guy; Rey, Frédéric; Aizawa, Rie; Suziki, Tetsu

    2013-01-01

    Conclusion: • Use of a LEMP motivated by several advantages in terms of the reactor design, operation and maintenance. • Collaboration agreement between the CEA and TOSHIBA Corporation came into force in April 2012 to carry out a joint work program on the ASTRID EMP design and development. • Preliminary LEMP calculations carried out by the CEA and TOSHIBA are in good agreement and provide a good confidence in the feasibility of the annular LEMP for the ASTRID intermediate sodium loop. • Theoretical and experimental investigations are currently underway at the CEA with the aim to improve the numerical tools. • In parallel, the ASTRID EMP conceptual design studies are ongoing at TOSHIBA (thermal and thermo-mechanical analyses to demonstrate the LEMP self-cooling, structural analysis of the casing, the supporting legs and the mechanical interfaces, definition of the power supply unit, instrumentation and remote control procedure). • This program is aiming at consolidating the ASTRID EMP conceptual design report and to support the design option choice for the ASTRID basic design

  7. Development of extreme rainfall PRA methodology for sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

    2016-01-01

    The objective of this study is to develop a probabilistic risk assessment (PRA) methodology for extreme rainfall with focusing on decay heat removal system of a sodium-cooled fast reactor. For the extreme rainfall, annual excess probability depending on the hazard intensity was statistically estimated based on meteorological data. To identify core damage sequence, event trees were developed by assuming scenarios that structures, systems and components (SSCs) important to safety are flooded with rainwater coming into the buildings through gaps in the doors and the SSCs fail when the level of rainwater on the ground or on the roof of the building becomes higher than thresholds of doors on first floor or on the roof during the rainfall. To estimate the failure probability of the SSCs, the level of water rise was estimated by comparing the difference between precipitation and drainage capacity. By combining annual excess probability and the failure probability of SSCs, the event trees led to quantification of core damage frequency, and therefore the PRA methodology for rainfall was developed. (author)

  8. Thermohydraulic relationships for advanced water cooled reactors and the role of the IAEA

    International Nuclear Information System (INIS)

    Badulescu, A.; Groeneveld, D.C.

    2000-01-01

    Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1998. It was included into the IAEA's Programme following endorsement in 1995 by the International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote International Information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water-cooled reactors. (authors)

  9. Advanced adsorption cooling cum desalination cycle: A thermodynamic framework

    KAUST Repository

    Chakraborty, Anutosh; Thu, Kyaw; Ng, K. C.

    2011-01-01

    We have developed a thermodynamic framework to calculate adsorption cooling cum desalination cycle performances as a function of pore widths and pore volumes of highly porous adsorbents, which are formulated from the rigor of thermodynamic property

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

    National Research Council Canada - National Science Library

    Can, Levent

    2006-01-01

    .... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

  11. Economizer Based Data Center Liquid Cooling with Advanced Metal Interfaces

    Energy Technology Data Exchange (ETDEWEB)

    Timothy Chainer

    2012-11-30

    A new chiller-less data center liquid cooling system utilizing the outside air environment has been shown to achieve up to 90% reduction in cooling energy compared to traditional chiller based data center cooling systems. The system removes heat from Volume servers inside a Sealed Rack and transports the heat using a liquid loop to an Outdoor Heat Exchanger which rejects the heat to the outdoor ambient environment. The servers in the rack are cooled using a hybrid cooling system by removing the majority of the heat generated by the processors and memory by direct thermal conduction using coldplates and the heat generated by the remaining components using forced air convection to an air- to- liquid heat exchanger inside the Sealed Rack. The anticipated benefits of such energy-centric configurations are significant energy savings at the data center level. When compared to a traditional 10 MW data center, which typically uses 25% of its total data center energy consumption for cooling this technology could potentially enable a cost savings of up to $800,000-$2,200,000/year (assuming electricity costs of 4 to 11 cents per kilowatt-hour) through the reduction in electrical energy usage.

  12. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  13. Design, in-sodium testing and performance evaluation of annular linear induction pump for a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Nashine, B.K.; Rao, B.P.C.

    2014-01-01

    Highlights: • Derivation of applicable design equations. • Design of an annular induction pump based on these equations. • Testing of the designed pump in a sodium test facility. • Performance evaluation of the designed pump. - Abstract: Annular linear induction pumps (ALIPs) are used for pumping electrically conducting liquid metals. These pumps find wide application in fast reactors since the coolant in fast reactors is liquid sodium which a good conductor of electricity. The design of these pumps is usually done using equivalent circuit approach in combination with numerical simulation models. The equivalent circuit of ALIP is similar to that of an induction motor. This paper presents the derivation of equivalent circuit parameters using first principle approach. Sodium testing of designed ALIP using the equivalent circuit approach is also described and experimental results of the testing are presented. Comparison between experimental and analytical calculations has also been carried out. Some of the reasons for variation have also been listed in this paper

  14. Sodium-cooled Fast Reactor Cores using Uranium-Free Metallic Fuels for Maximizing TRU Support Ratio

    International Nuclear Information System (INIS)

    You, WuSeung; Hong, Ser Gi

    2014-01-01

    The depleted uranium plays important roles in the SFR burner cores because it substantially contributes to the inherent safety of the core through the negative Doppler coefficient and large delayed neutron. However, the use of depleted uranium as a diluent nuclide leads to a limited value of TRU support ratio due to the generation of TRUs through the breeding. In this paper, we designed sodium cooled fast reactor (SFR) cores having uranium-free fuels 3,4 for maximization of TRU consumption rate. However, the uranium-free fuelled burner cores can be penalized by unacceptably small values of the Doppler coefficient and small delayed neutron fraction. In this work, metallic fuels of TRU-(W or Ni)-Zr are considered to improve the performances of the uranium-free cores. The objective of this work is to consistently compare the neutronic performances of uranium-free sodium cooled fast reactor cores having TRU-Zr metallic fuels added with Ni or W and also to clarify what are the problematic features to be resolved. In this paper, a consistent comparative study of 400MWe sodium cooled burner cores having uranium-based fuels and uranium-free fuels was done to analyze the relative core neutronic features. Also, we proposed a uranium-free metallic fuel based on Nickel. From the results, it is found that tungsten-based uranium-free metallic fuel gives large negative Doppler coefficient due to high resonance of tungsten isotopes but this core has large sodium void worth and small effective delayed neutron fraction while the nickel-based uranium-free metallic fuelled core has less negative Doppler coefficient but smaller sodium void worth and larger effective delayed neutron fraction than the tungsten-based one. On the other hand, the core having TRU-Zr has very high burnup reactivity swing which may be problematic in compensating it using control rods and the least negative Doppler coefficient

  15. Radiation heat transfer through the gas of a sodium cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Pradel, P.; Frachet, S.; Petit, D.

    1984-04-01

    Analysis based on results from the COCA test campaign and Germinal mockup of Super Phenix upper shuttings, of the heat transfers and radiation attenuation due to sodium aerosols between the free surface of sodium and the upper shuttings

  16. Numerical study of the underexpanded nitrogen jets submerged into liquid sodium in the frame of Sodium-cooled Fast Reactor (SFRs)

    International Nuclear Information System (INIS)

    Chen, F.; Allou, A.; Parisse, J.D.

    2017-01-01

    The study of the consequences of a gas leakage in the secondary/ tertiary heat exchangers is one of the essential points in the safety analysis of Sodium-cooled Fast nuclear Reactors (SFRs). This work is in the frame of the technology of the Compact plates Sodium-Gas heat Exchangers (ECSG) which is an alternative to conventional steam Rankine cycles. The overpressure of the tertiary nitrogen loop causes the formation of underexpanded gas jets submerged in the liquid sodium. In order to establish a safety evaluation, it would be an asset to be able to estimate the leakage. The gas leak detection by the acoustic method based on the bubbles field has been proposed. It requires then a delicate knowledge of the bubble field. This work contributes to development a numerical tool and its validation to model the transport and the production of bubbles in the downstream of underexpanded gas jets. The code CANOP modeling bi-phasic compressible flow is investigated under the actual condition of the underexpanded nitrogen jets submerged in the liquid sodium in an ECSG channel. Expensive computational cost is limited by using an Adaptive Mesh Refinement. (author)

  17. Sodium

    Science.gov (United States)

    Table salt is a combination of two minerals - sodium and chloride Your body needs some sodium to work properly. It helps with the function ... in your body. Your kidneys control how much sodium is in your body. If you have too ...

  18. Development of probabilistic risk assessment methodology against extreme snow for sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yamano, Hidemasa, E-mail: yamano.hidemasa@jaea.go.jp; Nishino, Hiroyuki; Kurisaka, Kenichi

    2016-11-15

    Highlights: • Snow PRA methodology was developed. • Snow hazard category was defined as the combination of daily snowfall depth (speed) and snowfall duration. • Failure probability models of snow removal action, manual operation of the air cooler dampers and the access route were developed. • Snow PRA showed less than 10{sup −6}/reactor-year of core damage frequency. - Abstract: This paper describes snow probabilistic risk assessment (PRA) methodology development through external hazard and event sequence evaluations mainly in terms of decay heat removal (DHR) function of a sodium-cooled fast reactor (SFR). Using recent 50-year weather data at a typical Japanese SFR site, snow hazard categories were set for the combination of daily snowfall depth (snowfall speed) and snowfall duration which can be calculated by dividing the snow depth by the snowfall speed. For each snow hazard category, the event sequence was evaluated by event trees which consist of several headings representing the loss of DHR. Snow removal action and manual operation of the air cooler dampers were introduced into the event trees as accident managements. Access route failure probability model was also developed for the quantification of the event tree. In this paper, the snow PRA showed less than 10{sup −6}/reactor-year of core damage frequency. The dominant snow hazard category was the combination of 1–2 m/day of snowfall speed and 0.5–0.75 day of snowfall duration. Importance and sensitivity analyses indicated a high risk contribution of the securing of the access routes.

  19. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Verma, V., E-mail: vasudha.verma@physics.uu.se [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Barbot, L.; Filliatre, P. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Hellesen, C. [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden); Jammes, C. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-lez-Durance (France); Svärd, S. Jacobsson [Division of Applied Nuclear Physics, Uppsala University, Box 516, SE-75120 Uppsala (Sweden)

    2017-07-11

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment. - Highlights: • Studied possibility of using SPNDs as in-core detectors in SFRs. • Study done to detect local power profile changes when reactor is at nominal power. • SPND with a Pt-emitter gives measurable prompt current of the order of 600 nA/m. • Dominant proportion of prompt response is maintained throughout the operation. • Detector signal gives dynamic information on the power fluctuations.

  20. New Sodium Cooled Long-Life Cores with Axially Multi-Driver Regions

    International Nuclear Information System (INIS)

    Hyun, Hae Ri; Hong, Ser Gi

    2014-01-01

    In this concept of long-life core (they are sometimes called B-B (Breed and Burn)), tall blanket is placed above the relatively short driver fuel. In the initial stage of burning, the power by fission is mostly generated in the driver region and it moves into the blanket region. The power and flux distributions that are highly peaked in the axial direction propagates slowly from the driver into the blanket region. This concept of long-life core fully utilizes the breeding of blanket in the fast spectra and it can achieve very high burnup of fuel. In this work, we introduce new sodium cooled longlife cores rating 600MWe (1800MWt). In these cores, the driver regions are heterogeneously placed into blanket region so as to achieve stabilized and less peaked axial power distribution as depletion proceeds. At present, our study is focused on only two axial driver regions but this concept can be easily extended onto the multi-driver region concept. The cores designed in this paper have two axial driver regions so as to have stabilized and less peaked axial power distributions as depletion proceeds. The results of the core design and analyses show that the cores have very long-lives longer than -49EFPYs and high discharge burnup higher than 200GWD/kg. Additionally, we considered a long-life core having no blanket. As expected, it was shown that these cores have stabilized and less peaked axial power distribution as the fuel depletes. However, the study shows that the cores having two driver regions still show high initial peaking of the axial power distributions and the core can be optimized by changing the driver fuel height

  1. Safeguards Considerations for the Design of a Future Fast Neutron Sodium Cooled Reactor

    International Nuclear Information System (INIS)

    Cazalet, J.; Raymond, P.; Masson, M.; Saturnin, A.

    2015-01-01

    Incorporating safeguards at an early stage of a reactor design is a way to increase the effectiveness and efficiency of safeguards measures minimizing the possibilities of misuse of the plant or nuclear material diversion. It also reduces the impact on the construction and operation cost. At the preliminary phase, the design will integrate: confinement, containment, surveillance features and non-destructive assay equipment. Taking into account these requirements will help the operator in the approval of the plant at the design phase by national and international authorities in charge of Nuclear Material accounting and safeguards. A large amount of work has been made by the GEN IV International Forum to assess the proliferation resistance of nuclear systems. The IAEA has developed guidelines on ''Safeguards by design'' describing reference requirements for future nuclear facilities. Based on these studies, this communication details implementation of safeguards in the design of a sodium cooled fast neutron reactor (SFR) currently studied in France. Specificities are the use of MOX fuel with high concentration of plutonium and the potential capacity of breeding. A great attention should be paid to avoid diversion of nuclear material contained in fresh or irradiated fuel. Scenarios of reactor misuse are analyzed. The identification of diversion pathways and requirements for nuclear material accountancy, leads to an approach of safeguards, specific to SFR: Material Balance Areas (MBA) and some key measurement points (KMP) are characterized. Specific instrumentation assay helping in the identification and/or characterization of fuel elements and the inventory of nuclear material is described. As concerns the fuel cycle, the safeguards of the reprocessing unit will be progressively increased through the development of materials monitoring and the implementation of these measures at strategic locations of buildings, thus providing real-time information

  2. Development of probabilistic risk assessment methodology against extreme snow for sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi

    2016-01-01

    Highlights: • Snow PRA methodology was developed. • Snow hazard category was defined as the combination of daily snowfall depth (speed) and snowfall duration. • Failure probability models of snow removal action, manual operation of the air cooler dampers and the access route were developed. • Snow PRA showed less than 10"−"6/reactor-year of core damage frequency. - Abstract: This paper describes snow probabilistic risk assessment (PRA) methodology development through external hazard and event sequence evaluations mainly in terms of decay heat removal (DHR) function of a sodium-cooled fast reactor (SFR). Using recent 50-year weather data at a typical Japanese SFR site, snow hazard categories were set for the combination of daily snowfall depth (snowfall speed) and snowfall duration which can be calculated by dividing the snow depth by the snowfall speed. For each snow hazard category, the event sequence was evaluated by event trees which consist of several headings representing the loss of DHR. Snow removal action and manual operation of the air cooler dampers were introduced into the event trees as accident managements. Access route failure probability model was also developed for the quantification of the event tree. In this paper, the snow PRA showed less than 10"−"6/reactor-year of core damage frequency. The dominant snow hazard category was the combination of 1–2 m/day of snowfall speed and 0.5–0.75 day of snowfall duration. Importance and sensitivity analyses indicated a high risk contribution of the securing of the access routes.

  3. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

    2017-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

  4. EXCURS: a computing programme for analysis of core transient behaviour in a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Saito, Shinzo

    1977-09-01

    In the code EXCURS developed for core transient behaviour calculation of a sodium-cooled fast reactor, a one-channel model is used to represent thermal behaviour of the reactor core. Calculations are made for three different channels; i.e. average, hot and hottest. In the average channel the power density and coolant velocity are equal to the mean values of the whole core. In the hot channel, a maximum power density of the core and a specific coolant velocity are introduced. In the hottest channel, engineering hot channel factors are considered to the hot channel. A one-point neutron kinetics equation with six delayed neutron groups is used to calculate the time-dependent power behaviour. Externally introduced reactivity effect and control rod movement in the case of a scram are taken into account. In the feedback effects evaluated on the basis of the average channel temperatures are considered Doppler effect, fuel axial expansion, cladding expansion, coolant expansion and structure expansion. The decay heat after reactor scram is also considered. Heat balance is taken in each cross section, neglecting the axial heat transfer except for the coolant region. Temperature dependence of the physical properties of materials is considered by second-order polynomials approximation, and also the fuel melting process. Each channel can be divided into a maximum of 20 regions in both radially and axially. The reactor core transient behaviour due to reactivity insertion or loss-of-coolant flow can be studied by EXCURS. The calculated results are plotted optionally by connected code EXPLOT. (auth.)

  5. Evaluation of a Sodium–Water Reaction Event Caused by Steam Generator Tubes Break in the Prototype Generation IV Sodium-cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    Sang June Ahn

    2016-08-01

    Full Text Available The prototype generation IV sodium-cooled fast reactor (PGSFR has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium–water reaction (SWR occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS and the safety of the primary heat-transfer system (PHTS. In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium–water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

  6. Developments and application of neutron noise diagnostics of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Zylbersztejn, F.

    2013-01-01

    The Sodium cooled Fast Reactor (SFR) is one of the six reactor types selected by the Generation-IV international forum (GIF), and the building of an industrial prototype is planned in France. The safety standard of the future SFR has to be equivalent to the EPR's. The general improvement of the safety of the new reactor goes through the examination of all the potentially harmful scenarios and both the study and monitoring of early signs. The mechanical deformations of the core can have harmful consequences in sodium fast reactors, such as unexpected power variations due to the reactivity increase in case of core compaction, or the excessive deterioration of the mechanical structures. The monitoring of such phenomena and of their potential early signs is then needed. The monitoring of such phenomena can be done with neutron detectors placed inside and outside the tank. This PhD thesis deals with the study of the neutron noise generated by the periodic deformation of the SFR core, restricted to the so-called core compaction or core flowering phenomenon, a deformation consisting in the variation of the inter-assembly sodium width by a radial bending the assemblies (the assemblies in SFR are held by the base). The PhD thesis has been performed within collaboration between CEA (France) and Chalmers Institute of Technology (Sweden). The work realized during the thesis led to the publication of 3 articles as first author and another as second author. This work has embraced the following topics: A state of the art of the monitoring of the core deformation phenomenon by interpretation of the noise measurements in SFR has been done. The PHENIX reactor multi physics measurements database has been scrutinized to provide an interpretation of the neutron noise bringing out mechanical vibration phenomena. An important conclusion was that the lack of theoretical knowledge about the neutron noise induced by the vibration phenomenon and the ill positioning of the neutron detectors

  7. Indiana State University Graduates to Advanced Plastic Cooling Towers

    Science.gov (United States)

    Sullivan, Ed

    2012-01-01

    Perhaps more than many other industries, today's universities and colleges are beset by dramatically rising costs on every front. One of the areas where overhead can be contained or reduced is in the operation of the chilled water systems that support air conditioning throughout college campuses, specifically the cooling towers. Like many…

  8. On the use of a moderation layer to improve the safety behavior in sodium cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merk, Bruno, E-mail: b.merk@fzd.de [Institute of Safety Research, Helmholtz-Zentrum Dresden-Rossendorf (Germany); Fridman, Emil; Weiss, Frank-Peter [Institute of Safety Research, Helmholtz-Zentrum Dresden-Rossendorf (Germany)

    2011-05-15

    Research highlights: > Using a moderation layer can reduce the sodium void effect in a SFR. > Inserting the moderation layer improves the Doppler effect significantly. > The uniform layer distribution avoids effects on power and burnup distribution. > Hydride containing material like uranium-zirconium hydride is most efficient. - Abstract: This work shows the effect of the use of moderating layers on the sodium void effect in sodium cooled fast breeder reactors. The moderating layers consisting of either boron carbide B{sub 4}C or uranium-zirconium hydride UZrH cause a strong reduction of the sodium void effect. Additionally these layers improve the fuel temperature effect and the coolant effect of the system. The use of the UZrH is significantly more effective for the reduction of the sodium void effect as well as for the improvement of the fuel temperature and the coolant effect. All changes cause by the insertion of the UZrH layer cause a significantly increased stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides.

  9. Development of GRIF-SM: The code for analysis of beyond design basis accidents in sodium cooled reactors

    International Nuclear Information System (INIS)

    Chvetsov, I.; Kouznetsov, I.; Volkov, A.

    2000-01-01

    GRIF-SM code was developed at the IPPE fast reactor department in 1992 for the analysis of transients in sodium cooled fast reactors under severe accident conditions. This code provides solution of transient hydrodynamics and heat transfer equations taking into account possibility of coolant boiling, fuel and steel melting, reactor kinetics and reactivity feedback due to variations of the core components temperature, density and dimensions. As a result of calculation, transient distribution of the coolant velocity and density was determined as well as temperatures of the fuel pins, reactor core and primary circuit as a whole. Development of the code during further 6 years period was aimed at the modification of the models describing thermal hydraulic characteristics of the reactor, and in particular in detailed description of the sodium boiling process. The GRIF-SM code was carefully validated against FZK experimental data on steady state sodium boiling in the electrically heated tube; transient sodium boiling in the 7-pin bundle; transient sodium boiling in the 37-pin bundle under flow redaction simulating ULOF accident. To show the code capabilities some results of code application for beyond design basis accident analysis on BN-800-type reactor are presented. (author)

  10. Design study of an IHX support structure for a POOL-TYPE Sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Park, Chang Gyu; Kim, Jong Bum; Lee, Jae Han

    2009-01-01

    The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity

  11. Optimization of material and production to develop fluoroelastomer inflatable seals for sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.i [Indira Gandhi Centre for Atomic Research (IGCAR), Department of Atomic Energy (DAE), Kalpakkam, Tamilnadu 603102 (India); Raj, Baldev, E-mail: dir@igcar.gov.i [Indira Gandhi Centre for Atomic Research (IGCAR), Department of Atomic Energy (DAE), Kalpakkam, Tamilnadu 603102 (India)

    2011-03-15

    Research highlights: Production of thin fluoroelastomer profiles by cold feed extrusion and continuous cure involving microwave and hot air heating. Use of peroxide curing in air during production. Use of fluoroelastomers based on advanced polymer architecture (APA) for the production of profiles. Use of the profiles in inflatable seals for critical application of Prototype Fast Breeder Reactor. Tailoring of material formulation by synchronized optimization of material and production technologies to ensure that the produced seal ensures significant gains in terms of performance and safety in reactor under synergistic influences of temperature, radiation, air and sodium aerosol. - Abstract: The feasibility of producing thin-walled fluoroelastomer profiles under continuous, atmospheric-pressure vulcanization conditions in air has been demonstrated by successful manufacture of {approx}2 m diameter test inflatable seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) using a 50/50 blend formulation of Viton GBL-200S/600S based on advanced polymer architecture (APA). A commercial cold feed screw extruder with 90 mm diameter screw was used along with continuous cure by microwave (2.45 GHz) and hot air heating (190 {sup o}C) at a line speed of 1 m/min to produce the seals. The blend formulation promises significant improvement in the performance and safety of the seals. This article depicts the relevant characteristics of the original inflatable seal compound that was used as reference to achieve the objectives through synchronized optimization of material and production technologies. The production trials are outlined and the blend formulation used with minor factory modifications to produce the test seals is reported. Progressive refinements of the original, Viton A-401C based compound to the blend formulation is presented along with an assessment of potential performance gains. Possible uses of the reported formulation and production technique for other large

  12. Optimization of material and production to develop fluoroelastomer inflatable seals for sodium cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Sinha, N.K.; Raj, Baldev

    2011-01-01

    Research highlights: → Production of thin fluoroelastomer profiles by cold feed extrusion and continuous cure involving microwave and hot air heating. → Use of peroxide curing in air during production. → Use of fluoroelastomers based on advanced polymer architecture (APA) for the production of profiles. → Use of the profiles in inflatable seals for critical application of Prototype Fast Breeder Reactor. → Tailoring of material formulation by synchronized optimization of material and production technologies to ensure that the produced seal ensures significant gains in terms of performance and safety in reactor under synergistic influences of temperature, radiation, air and sodium aerosol. - Abstract: The feasibility of producing thin-walled fluoroelastomer profiles under continuous, atmospheric-pressure vulcanization conditions in air has been demonstrated by successful manufacture of ∼2 m diameter test inflatable seals for the 500 MWe, Prototype Fast Breeder Reactor (PFBR) using a 50/50 blend formulation of Viton GBL-200S/600S based on advanced polymer architecture (APA). A commercial cold feed screw extruder with 90 mm diameter screw was used along with continuous cure by microwave (2.45 GHz) and hot air heating (190 o C) at a line speed of 1 m/min to produce the seals. The blend formulation promises significant improvement in the performance and safety of the seals. This article depicts the relevant characteristics of the original inflatable seal compound that was used as reference to achieve the objectives through synchronized optimization of material and production technologies. The production trials are outlined and the blend formulation used with minor factory modifications to produce the test seals is reported. Progressive refinements of the original, Viton A-401C based compound to the blend formulation is presented along with an assessment of potential performance gains. Possible uses of the reported formulation and production technique for

  13. The report of inspection and repair technology of sodium cooled reactors

    International Nuclear Information System (INIS)

    Kisohara, Naoyuki; Uchita, Masato; Konomura, Mamoru

    2002-12-01

    Sodium is the most promising candidate of an FBR coolant because of its excellent properties such as high thermal conductivity. Whereas, sodium reacts with water/air and its opaqueness makes it difficult to inspect sodium components. These weaknesses of sodium affect not only plant safety but also plant availability (economy). To overcome these sodium weak points, the appropriate countermeasure must be adopted to commercialized FBR plants. This report describes the working group activities for sodium/water reaction of steam generators (SG), in-service inspection for sodium components and sodium leak due to sodium components boundary failure. The prospect of each countermeasure is discussed in the viewpoint of the commercialized FBR plants. 1) Sodium/water reaction. The principle of the countermeasure for sodium/water reaction accidents was organized in the viewpoint of economy (the investment of SG and the plant availability). The countermeasures to restrain failure propagation were investigated for a large-sized SG. Preliminary analysis revealed the possibility of minimizing tubes failure propagation by improving the leak detection system and the blow down system. Detailed failure propagation analysis will be required and the early water leak detection system and rapid blow system must be evaluated to realize its performance. 2) In-service inspection (ISI and R). The viewpoint of the commercialized plant's ISI and R was organized by comparing with the prototype reactor's ISI and R method. We also investigated short-term ISI and R method without sodium draining to prevent the degrading of the plant availability, however, it is difficult to realize the with the present technology. Hereafter, the ISI and R of the commercialized plants must be defined by considering its characteristics. 3) Sodium leak from the components. This report organized the basic countermeasure policy for primary and secondary sodium leak accidents. Double-wall structure of sodium piping was

  14. Advances in electron cooling in heavy-ion storage rings

    International Nuclear Information System (INIS)

    Danared, H.

    1994-01-01

    The efficiency of electron cooling can be improved by reducing the temperature of the electrons. If the magnetic field at the location of the electron gun is stronger than in the region where the electrons interact with the ions, and the field gradient is adiabatic with respect to the cyclotron motion of the electrons, the resulting expansion of the electron beam reduces its transverse temperature by a factor equal to the ratio between the two fields. A ten times expanded electron beam was introduced in the CRYRING electron cooler in the summer of 1993, and similar arrangements have since then been made at the TSR ring in Heidelberg and at ASTRID in Aarhus. The reduction of the transverse electron temperature has increased cooling rates with large factors, and improves the energy resolution and increases count rates when the cooler is used as an electron target for ion-electron recombination experiments

  15. Advanced adsorption cooling cum desalination cycle: A thermodynamic framework

    KAUST Repository

    Chakraborty, Anutosh

    2011-01-01

    We have developed a thermodynamic framework to calculate adsorption cooling cum desalination cycle performances as a function of pore widths and pore volumes of highly porous adsorbents, which are formulated from the rigor of thermodynamic property surfaces of adsorbent-adsorbate system and the adsorption interaction potential between them. Employing the proposed formulations, the coefficient of performance (COP) and overall performance ratio (OPR) of adsorption cycle are computed for various pore widths of solid adsorbents. These results are compared with experimental data for verifying the proposed thermodynamic formulations. It is found from the present analysis that the COP and OPR of adsorption cooling cum desalination cycle is influenced by (i) the physical characteristics of adsorbents, (ii) characteristics energy and (iii) the surface-structural heterogeneity factor of adsorbent-water system. The present study confirms that there exists a special type of adsorbents having optimal physical characteristics that allows us to obtain the best performance.

  16. Advanced turbine cooling, heat transfer, and aerodynamic studies

    Energy Technology Data Exchange (ETDEWEB)

    Je-Chin Han; Schobeiri, M.T. [Texas A& M Univ., College Station, TX (United States)

    1995-10-01

    The contractual work is in three parts: Part I - Effect of rotation on enhanced cooling passage heat transfer, Part II - Effect on Thermal Barrier Coating (TBC) spallation on surface heat transfer, and Part III - Effect of surface roughness and trailing edge ejection on turbine efficiency under unsteady flow conditions. Each section of this paper has been divided into three parts to individually accommodate each part. Part III is further divided into Parts IIIa and IIIb.

  17. Effect of horizontal flow on the cooling of the moderator brick in the advanced gas-cooled reactor

    International Nuclear Information System (INIS)

    Ganesan, P.; He, S.; Hamad, F.; Gotts, J.

    2011-01-01

    The paper reports an investigation of the effect of the horizontal cross flow on the temperature of the moderator brick in UK Advanced Gas-cooled Reactor (AGR) using computational fluid dynamics (CFD) with a conjugate heat transfer model for the solid and fluid. The commercial software package of ANSYS Fluent is used for this purpose. The CFD model comprises the full axial length of one-half of a typical fuel channel (assuming symmetry) and part of neighbouring channels on either side. Two sets of simulations have been carried out, namely, one with cross flow and one without cross flow. The effect of cross flow has subsequently been derived by comparing the results from the two groups of simulations. The study shows that a small cross flow can have a significant effect on the cooling of the graphite brick, causing the peak temperature of the brick to reduce significantly. Two mechanisms are identified to be responsible for this. Firstly, the small cross flow causes a significant redistribution of the main axial downward flow and this leads to an enhancement of heat transfer in some of the small clearances, and an impairment in others although overall, the enhancement is dominant leading to a better cooling. Secondly, the cross flow makes effective use of the small clearances between the key/keyway connections which increases the effective heat transfer area, hence increasing the cooling. Under the conditions of no cross flow, these areas remain largely inactive in heat transfer. The study shows that the cooling of the moderator is significantly enhanced by the cross flow perpendicular to the main cooling flow. (author)

  18. 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)

  19. Gas Cooled Fast Reactors: Recent advances and prospects

    International Nuclear Information System (INIS)

    Poette, C.; Guedeney, P.; Stainsby, R.; Mikityuk, K.; Knol, S.

    2013-01-01

    Gas Cooled Fast Reactors: Conclusion - GFR: an attractive longer term option allowing to combine Fast spectrum & Helium coolant benefits; • Innovative SiC fuel cladding solutions were found; • A first design confirming the encouraging potential of the reactor system Design improvements are nevertheless recommended and interesting tracks have been identified (core & system design, DHR system); • The GFR requires large R&D needs to confirm its potential (fuel & core materials, specific Helium technology); • ALLEGRO prototype studies are the first step and are drawing the R&D priorities

  20. Low grade heat driven adsorption system for cooling and power generation using advanced adsorbent materials

    International Nuclear Information System (INIS)

    Al-Mousawi, Fadhel Noraldeen; Al-Dadah, Raya; Mahmoud, Saad

    2016-01-01

    Highlights: • Adsorption system based on water and advanced physical adsorbents has the potential of producing cooling and power. • Adding an expander to physisorption system enhances efficiency by up to 11%. • MIL101Cr MOF can produce 95 W/kg and 1357 W/kg of specific power and cooling. • AQSOA Z02 can produce 73 W/kg and 640 W/kg of specific power and cooling. - Abstract: Globally there is abundance of low grade heat sources (around 150 °C) from renewables like solar energy or from industrial waste heat. The exploitation of such low grade heat sources will reduce fossil fuel consumption and CO_2 emissions. Adsorption technology offers the potential of using such low grade heat to generate cooling and power. In this work, the effect of using advanced adsorbent materials like AQSOA-Z02 (SAPO-34) zeolite and MIL101Cr Metal Organic Framework (MOF) at various operating conditions on power and cooling performance compared to that of commonly used silica-gel was investigated using water as refrigerant. A mathematical model for a two bed adsorption cooling cycle has been developed with the cycle modified to produce power by incorporating an expander between the desorber and the condenser. Results show that it is possible to produce power and cooling at the same time without affecting the cooling output. Results also show that for all adsorbents used as the heat source temperature increases, the cooling effect and power generated increase. As for increasing the cold bed temperature, this will decrease the cooling effect and power output except for SAPO-34 which shows slightly increasing trend of cooling and power output. As the condenser cooling temperature increases, the cooling effect and power output will decrease while for the chilled water temperature, the cooling load and power generated increased as the temperature increased. The maximum values of average specific power generation (SP), specific cooling power (SCP) and cycle efficiency are 73 W

  1. Calculation of the neutron noise induced by periodic deformations of a large sodium-cooled fast reactor core

    International Nuclear Information System (INIS)

    Zylbersztejn, F.; Tran, H.N.; Pazsit, I.; Filliatre, P.; Jammes, C.

    2014-01-01

    The subject of this paper is the calculation of the neutron noise induced by small-amplitude stationary radial variations of the core size (core expansion/compaction, also called core flowering) of a large sodium-cooled fast reactor. The calculations were performed on a realistic model of the European Sodium Fast Reactor (ESFR) core with a thermal output of 3600 MW(thermal), using a multigroup neutron noise simulator. The multigroup cross sections and their fluctuations that represent the core geometry changes for the neutron noise calculations were generated by the code ERANOS. The space and energy dependences of the noise source represented by the core expansion/compaction and the induced neutron noise are calculated and discussed. (authors)

  2. Study of thermophysical and thermohydraulic properties of sodium for fast sodium cooled reactors; Estudio de las propiedades termofisicas y termohidraulicas del sodio para reactores rapidos enfriados por sodio

    Energy Technology Data Exchange (ETDEWEB)

    Vega R, A. K.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Gomez T, A. M., E-mail: a.karen.vr@gmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The importance of liquid sodium lies in its use as a coolant for fast reactors, but why should liquid metal be used as a coolant instead of water? Water is difficult to use as a coolant for a fast nuclear reactor because its acts as a neutron moderator, that is, stop the fast neutrons and converts them to thermal neutrons. Nuclear reactors such as the Pressurized Water Reactor or the Boiling Water Reactor are thermal reactors, which mean they need thermal neutrons for their operation. However, is necessary for fast reactors to conserve as much fast neutrons, so that the liquid metal coolants that do have this capability are implemented. Sodium does not need to be pressurized, its low melting point and its high boiling point, higher than the operating temperature of the reactor, make it an adequate coolant, also has a high thermal conductivity, which is necessary to transfer thermal energy and its viscosity is close to that of the water, which indicates that is an easily transportable liquid and does not corrode the steel parts of the reactor. This paper presents a brief state of the art of the rapid nuclear reactors that operated and currently operate, as well as projects in the door in some countries; types of nuclear reactors which are cooled by liquid sodium and their operation; the mathematical models for obtaining the properties of liquid sodium in a range of 393 to 1673 Kelvin degrees and a pressure atmosphere. Finally a program is presented in FORTRAN named Thermo-Sodium for the calculation of the properties, which requires as input data the Kelvin temperature in which the liquid sodium is found and provides at the user the thermo-physical and thermo-hydraulic properties for that data temperature. Additional to this the user is asked the Reynolds number and the hydraulic diameter in case of knowing them, and in this way the program will provide the value of the convective coefficient and that of the dimensionless numbers: Nusselt, Prandtl and Peclet. (Author)

  3. Modeling of the acoustic boiling noise of sodium during an assembly blockage in sodium-cooled reactors

    International Nuclear Information System (INIS)

    Vanderhaegen, M.

    2013-01-01

    In the framework of the fourth generation of nuclear reactors safety requirements, the acoustic boiling detection is studied to detect subassembly blockages. Boiling, that might occur during subassembly blockages and that can lead to clad failure, generates hydrodynamic noise that can be related to the two-phase flow. A bubble dynamics study shows that the sound source during subassembly boiling is condensation. This particular phenomenon generates most noise as a high subcooling is present in the subassembly and because of the high thermal diffusivity of sodium. This result leads to an estimate of the form of the acoustic spectrum that will be filtered and amplified during propagation inside the liquid. And even though it is unlikely that bubbles will be present inside the subassembly, due to the very gradual temperature profile at the wall and due to the geometry that leads to a strong confinement of the vapor, the historical bubble dynamics approach gives some insight in previous measurements. Additionally, some hypotheses can be disproved. These theoretical ideas are validated with a small water experiment, yet it also shows that a simple experience in sodium doesn't lead to a better knowledge of the acoustic source. A theoretical analysis also revealed that a realistic experiment with a simulant fluid, such as water or mercury, isn't representative. A similar conclusion is obtained when studying cavitation as a simulant acoustic source. As such, the acoustic detection of boiling, in comparison with other detection systems, isn't sufficiently developed yet to be applied as a reactor protective system. (author) [fr

  4. A Review of PSA Technology Applications according to the Development of Sodium-cooled Fast Reactors in the World

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Lee, Yong Bum; Jung, Hae Yong; Kim, Sang Ji; Hahn, Do Hee; Yang, Joon Eon

    2008-12-01

    The international nuclear societies request to perform Probabilistic Safety Assessment (PSA) according to the development of Gen IV Sodium-cooled Fast Reactors (SFR). One of the major tasks of the PSA is to identify various sequences of events which could lead to the release of radioactivity. However, due to the limited operating and SFR PSA experiences, it will be difficult to derive and to quantify core damage frequency for SFR under development in Korea, so called KALIMER. Hence, in this report, the foreign PSA results, such as USA and Japan, are analyzed based on the obtained documents. Finally an approach on how to perform PSA for KALIMER is suggested

  5. Nuclear Power Station Kalkar, 300 MWe Prototype Nuclear Power Plant with Fast Sodium Cooled Reactor (SNR-300), Plant description

    International Nuclear Information System (INIS)

    1984-06-01

    The nuclear power station Kalkar (SNR-300) is a prototype with a sodium cooled fast reactor and a thermal power of 762 MW. The present plant description has been made available in parallel to the licensing procedure for the reactor plant and its core Mark-Ia as supplementary information for the public. The report gives a detailed description of the whole plant including the prevention measures against the impact of external and plant internal events. The radioactive materials within the reactor cooling system and the irradiation protection and surveillance measures are outlined. Finally, the operation of the plant is described with the start-up procedures, power operation, shutdown phases with decay heat removal and handling procedures

  6. Methods for studying fuel management in advanced gas cooled reactors

    International Nuclear Information System (INIS)

    Buckler, A.N.; Griggs, C.F.; Tyror, J.G.

    1971-07-01

    The methods used for studying fuel and absorber management problems in AGRs are described. The basis of the method is the use of ARGOSY lattice data in reactor calculations performed at successive time steps. These reactor calculations may be quite crude but for advanced design calculations a detailed channel-by-channel representation of the whole core is required. The main emphasis of the paper is in describing such an advanced approach - the ODYSSEUS-6 code. This code evaluates reactor power distributions as a function of time and uses the information to select refuelling moves and determine controller positions. (author)

  7. ASTRID: Advanced Sodium Technological Reactor for Industrial Demonstration

    International Nuclear Information System (INIS)

    Vasile, A.

    2012-01-01

    Conclusions: • R&D results [CEA-AREVA-EDF] obtained from 2007 to 2009 have contributed to ASTRID mid 2010 choice of options; • ASTRID has the objective to demonstrate at the industrial scale progress in the identified domains of SFR weakness (safety, operability, economy). and to perform transmutation demonstrations; • A lot of improvements are related to safety; • The first very important milestone is 2012 (June 2006 French Act on wastes management): – ASTRID pre-conceptual design studies: 2010-2012; – First investment cost evaluation; – First safety Authorities advice on the orientations for ASTRID safety; • With the ASTRID program funded by the French government, France has the opportunity to develop a GEN IV Sodium Fast Reactor

  8. A directly cooled grating substrate for ALS [Advanced Light Source] undulator beam lines

    International Nuclear Information System (INIS)

    DiGennaro, R.; Swain, T.

    1989-08-01

    Design analyses using finite element methods are presented for thermal distortion of water-cooled diffraction grating substrates for a potential application at the LBL Advanced Light Source, demonstrating that refinements in cooling channel configuration and heat flux distribution can significantly reduce optical surface distortion with high heat loads. Using an existing grating substrate design, sensitivity of tangential slope errors due to thermal distortion is evaluated for a variety of thermal boundary conditions, including coolant flow rate and heat transfer film coefficients, surface illumination area and heat distribution profile, and location of the convection cooling surfaces adjacent to the heated region. 1 ref., 5 figs., 2 tabs

  9. Advances in passive cooling design and performance analysis

    International Nuclear Information System (INIS)

    Woodcock, J.

    1994-01-01

    The Third International Conference on Containment Design and Operation continues the trend of rapidly extending the state of the art in containment methodology, joining other conferences, OECD-sponsored International Standard Problem exercises, and vendor licensing submittals. Methodology developed for use on plants with passive features is under increasing scrutiny for advanced designs, since the passive features are often the only deviation from existing operating base of the past 30 years of commercial nuclear power. This session, 'Containment Passive Safety Systems Design and Operation,' offers papers on a wide range of topics, with authors from six organizations from around the world, dealing with general passive containments, Westinghouse AP600, large (>1400 MWe) passive plants, and the AECL advanced CANDU reactor. This level and variety of participation underscores the high interest and accelerated methods development associated with advanced passive containment heat removal. The papers presented in this session demonstrate that significant contributions are being made to the advancement of technology necessary for building a new generation of safer, more economical nuclear plants. (author)

  10. Fire protection at the Fast Flux Test Facility (a sodium cooled test reactor)

    International Nuclear Information System (INIS)

    Bell, J.R.

    1980-01-01

    For purposes of this presentation, fire protection at the FFTF is subdivided into two catagories; protection for non-sodium areas and protection for areas containing sodium. Fire protection systems and philosophies for non-sodium areas at the FFTF are very similar to those used at conventional power plants being constructed throughout the country. They follow, essentially, the NRC rules and guidelines and ANSI 59.4 Generic Requirements for Light Water Nuclear Power Plant Fire Protection. The FFTF with its support facilities have their own water system comprised of a looped 8'' and 10'' underground distribution system, three 1500 GPM fire pumps and three ground level storage tanks totaling 736,000 gallons with 420,000 reserved for fire protection. Fire hydrants are enclosed with hose houses outfitted for use by the Emergency Response Team (ERT). Fire prevention systems for sodium areas of the FFTF are also described

  11. Seismic snubber reduction on advanced gas-cooled reactor pipework

    International Nuclear Information System (INIS)

    Kennedy, P.A.; Harkin, N.J.

    1989-01-01

    Recent advances in pipework dynamic analysis procedures have enabled a more realistic approach to be taken to the design of pipework under earthquake loadings. In particular, it is proving possible to reduce the number of seismic snubbers employed to limit pipework displacements. This paper presents the background to, and outcome of, a snubber optimisation study performed for the main steam pipework system at Torness Nuclear Power Station. (author)

  12. Progress in development and design aspects of advanced water cooled reactors

    International Nuclear Information System (INIS)

    1992-12-01

    The objective of the Technical Committee Meeting (TCM) was to provide an international forum for technical specialists to review and discuss technology developments and design work for advanced water cooled reactors, safety approaches and features of current water cooled reactors and to identify, understand and describe advanced features for safety and operational improvements. The TCM was attended by 92 participants representing 18 countries and two international organizations and included 40 presentations by authors of 14 countries and one international organization. A separate abstract was prepared for each of these presentations. Refs, figs, tabs

  13. Future nuclear systems, Astrid, an option for the fourth generation: preparing the future of nuclear energy, sustainably optimising resources, defining technological options, sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Ter Minassian, Vahe

    2016-01-01

    Energy independence and security of supplies, improved safety standards, sustainably optimised material management, minimal waste production - all without greenhouse gas emissions. These are the Generation IV International Forum specifications for nuclear energy of the future. The CEA is responsible for designing Astrid, an integrated technology demonstrator for the 4. generation of sodium-cooled fast reactors, in accordance with the French Sustainable Nuclear Materials and Waste Management Act of June 28, 2006, and funded as part of the Investments for the Future programme enacted by the French parliament in 2010. Energy management - a vital need and a factor of economic growth - is a major challenge for the world of tomorrow. The nuclear industry has significant advantages in this regard, although it faces safety, resource sustainability, and waste management issues that must be met through continuing technological innovation. Fast reactors are also of interest to the nuclear industry because their recycling capability would solve a number of problems related to the stockpiles of uranium and plutonium. After the resumption of R and D work with EDF and AREVA in 2006, the Astrid design studies began in 2010. The CEA, as owner and contracting authority for this programme, is now in a position to define the broad outlines of the demonstrator 4. generation reactor that could be commissioned during the next decade. A sodium-cooled fast reactor (SFR) operates in the same way as a conventional nuclear reactor: fission reactions in the atoms of fuel in the core generate heat, which is conveyed to a turbine generator to produce electricity. In the context of 4. generation technology, SFRs represent an innovative solution for optimising the use of raw materials as well as for enhancing safety. Here are a few ideas advanced by the CEA. (authors)

  14. Development of Core Heat Removal Objective Provision Trees for Sodium-Cooled Fast Reactor Defense-in-Depth Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Huichang; Kang, Bongsuk; Lee, Youngho [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    Based on the definition of Defense-in-Depth levels and safety functions for KALIMER sodium-cooled fast reactor, suggested in the reference and, OPTs for level 1, 2, and 3 defense-in-depth and core heat removal safety function, were developed and suggested in this paper. The purpose of this OPT is first to assure the defensein-depth design during the licensing of Sodium-Cooled Fast Reactors (SFR), but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of PGSFR (Prototype Gen-IV SFR) is not mature yet, the OPT is developed for KALIMER design. Developed OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defensein-depth evaluation frame for the regulatory reviews for the licensing process. At this moment, the identified provisions have both aspects as requirements and design features already adopted in KALIMER design. In the next stage of this study, derived provisions to be adopted will be compared with the actual design features and findings can be suggested as recommendations for the safety improvement.

  15. Two neural network based strategies for the detection of a total instantaneous blockage of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Martinez-Martinez, Sinuhe; Messai, Nadhir; Jeannot, Jean-Philippe; Nuzillard, Danielle

    2015-01-01

    The total instantaneous blockage (TIB) of an assembly in the core of a sodium-cooled fast reactor (SFR) is investigated. Such incident could appear as an abnormal rise in temperature on the assemblies neighbouring the blockage. Its detection relies on a dataset of temperature measurements of the assemblies making up the core of the French Phenix Nuclear Reactor. The data are provided by the French Commission of Atomic and Alternatives Energies (CEA). Here, two strategies are proposed depending on whether the sensor measurement of the suspected assembly is reliable or not. The proposed methodology implements a time-lagged feed-forward neural (TLFFN) Network in order to predict the one-step-ahead temperature of a given assembly. The incident is declared if the difference between the predicted process and the actual one exceeds a threshold. In these simulated conditions, the method is efficient to detect small gradients as expected in reality. - Highlights: • We study the total instantaneous blockage (TIB) of a sodium-cooled fast reactor. • The TIB symptom is simulated as an abrupt rise on temperature (0.1–1 °C/s). • The goal is to improve the early detection of the incident. • Two strategies laying on neural networks are proposed. • TIB is detected in 3 s for 1 °C/s and 18–21 s for 0.1 °C/s

  16. Variable electricity and steam from salt, helium and sodium cooled base-load reactors with gas turbines and heat storage - 15115

    International Nuclear Information System (INIS)

    Forsberg, C.; McDaniel, P.; Zohuri, B.

    2015-01-01

    Advances in utility natural-gas-fired air-Brayton combed cycle technology is creating the option of coupling salt-, helium-, and sodium-cooled nuclear reactors to Nuclear air-Brayton Combined Cycle (NACC) power systems. NACC may enable a zero-carbon electricity grid and improve nuclear power economics by enabling variable electricity output with base-load nuclear reactor operations. Variable electricity output enables selling more electricity at times of high prices that increases plant revenue. Peak power is achieved using stored heat or auxiliary fuel (natural gas, bio-fuels, hydrogen). A typical NACC cycle includes air compression, heating compressed air using nuclear heat and a heat exchanger, sending air through a turbine to produce electricity, reheating compressed air, sending air through a second turbine, and exhausting to a heat recovery steam generator (HRSG). In the HRSG, warm air produces steam that is used to produce added electricity. For peak power production, auxiliary heat (natural gas, stored heat) is added before the air enters the second turbine to raise air temperatures and power output. Like all combined cycle plants, water cooling requirements are dramatically reduced relative to other power cycles because much of the heat rejection is in the form of hot air. (authors)

  17. Innovating analytical spectroscopies for the improvement of liquid sodium cooled fast neutron reactors safety

    International Nuclear Information System (INIS)

    Maury, C.

    2012-01-01

    In the context of the project of sodium fast reactor ASTRID, CEA is currently developing new analytical techniques to monitor the chemical purity of liquid sodium. Indeed, incidental situations occurring in the reactor, such as fuel clad failures, leakages in the steam generator or in the coolant pumps, and accelerated corrosion, might release several elements in the sodium. Analytical techniques based on laser ablation and emission spectroscopy are well suited for this application. They do not require any sample preparation, and can perform direct on-line analysis. Amongst them, Laser-Induced Breakdown Spectroscopy (LIBS) and Laser-Ablation coupled to Laser-Induced Fluorescence (LA-LIF) have been selected for this study. The objective of this work was to characterize the sensitivity of those two techniques for the detection of impurities in liquid sodium. Their limits of detection were calculated for model analytes using calibration lines. Then results were theoretically extrapolated to other analytes of interest. This study shows the feasibility of the detection of steel corrosion products in liquid sodium. However, the LIBS technique is more robust and easier to implement, and would therefore be more suited to nuclear conditions. (author) [fr

  18. Accident alarm in steam generators in sodium cooled fast reactor power plants. II

    International Nuclear Information System (INIS)

    Matal, O.; Martoch, J.; Taraba, O.; Hanke, V.

    1978-01-01

    Conditions were simulated in the economizer of a steam generator of water leaks in sodium at a sodium flow of O.62x10 -3 to 1.24x10 -3 m 3 /s and a sodium temperature of 320 to 380 degC by injecting water at a pressure of 6 to 10 MPa which roughly corresponds to conditions in an economizer of an actual steam generator with leaks within the limits of 0.01 to 0.3 g/s. The leak was recorded by acoustic detectors at all observed sodium flow rates and temperatures. The mean signal-to-noise ratio was in all cases greater than 2. At the assumed 25 dB noise level of the real steam generator of micromodular design it may be assumed that using existing acoustic detectors with waveguides a 0.02 g/s leak of water into sodium may be detected. The measurements showed that the technical standard of the equipment is at least as good as that of the flowmeter system of accident monitoring. (J.B.)

  19. A moderation layer to improve the safety behavior of sodium cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Merk, B.; Weiß, F.P., E-mail: b.merk@fzd.de [Forschungszentrum Dresden-Rossendorf, Institut für Sicherheitsforschung, Dresden (germany)

    2011-07-01

    The nature of the sodium void effect in an infinite lattice is discussed and for a reduction of the effect the insertion of moderating material is proposed. The effect of three different moderating layers on the sodium void defect and the feedback effects is investigated. Especially the uranium zirconium hydride UzrH layer causes a strong reduction of the sodium void effect. Additionally, this layer improves the fuel temperature effect and the coolant effect of the system significantly. All changes caused by the insertion of the UZrH layer lead to a significant increase in stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides. (author)

  20. A moderation layer to improve the safety behavior of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Merk, B.; Weiß, F.P.

    2011-01-01

    The nature of the sodium void effect in an infinite lattice is discussed and for a reduction of the effect the insertion of moderating material is proposed. The effect of three different moderating layers on the sodium void defect and the feedback effects is investigated. Especially the uranium zirconium hydride UzrH layer causes a strong reduction of the sodium void effect. Additionally, this layer improves the fuel temperature effect and the coolant effect of the system significantly. All changes caused by the insertion of the UZrH layer lead to a significant increase in stability of the fast reactor system against transients. The moderating layers have only a small influence on the breeding effect and on the production of minor actinides. (author)

  1. Characterization of a sodium-cooled fast reactor in an MHR-SFR synergy for TRU transmutation

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Kim, Yonghee; Venneri, Francesco

    2008-01-01

    In the task of destroying the light water reactor (LWR) transuranics (TRUs), we consider the concept of a synergistic combination of a deep-burn (DB) gas-cooled reactor followed by a sodium-cooled fast reactor (SFR), as an alternative way to the direct feeding of the LWR TRUs to the SFR. In the synergy concept, TRUs from LWR are first deeply incinerated in a graphite-moderated DB-MHR (modular helium reactor) and then the spent fuels of DB-MHR are recycled into the closed-cycle SFR. The DB-MHR core is 100% TRU-loaded and a deep-burning (50-65%) is achieved in a safe manner (as discussed in our previous work). In this analysis, the SFR fuel cycle is closed with a pyro-processing technology to minimize the waste stream to a final repository. Neutronic characteristics of the SFR core in the MHR-SFR synergy have been evaluated from the core physics point of view. Also, we have compared core characteristics of the synergy SFR with those of a stand-alone SFR transuranic burner. For a consistent comparison, the two SFRs are designed to have the same TRU consumption rate of ∼250 kg/GW EFPY that corresponds to the TRU discharge rate from three 600 MW DB-MHRs. The results of our work show that the synergy SFR, fed with TRUs from DB-MHR, has a much smaller burnup reactivity swing, a slightly greater delayed neutron fraction (both positive features) but also a higher sodium void worth and a less negative Doppler coefficients than the conventional SFR, fed with TRUs directly from the LWRs. In addition, several design measures have been considered to reduce the sodium void worth in the synergy SFR core

  2. System design study of a membrane reforming hydrogen production plant using a small sized sodium cooled reactor

    International Nuclear Information System (INIS)

    Chikazawa, Y.; Konomura, M.; Hori, T.; Sato, H.; Uchida, S.

    2004-01-01

    In this study, a membrane reforming hydrogen production plant using a small sized sodium cooled reactor was designed as one of promising concepts. In the membrane reformer, methane and steam are reformed into carbon dioxide and hydrogen with sodium heat at a temperature 500 deg-C. In the equilibrium condition, steam reforming proceeds with catalyst at a temperature more than 800 deg-C. Using membrane reformers, the steam reforming temperature can be decreased from 800 to 500 deg-C because the hydrogen separation membrane removes hydrogen selectively from catalyst area and the partial pressure of hydrogen is kept much lower than equilibrium condition. In this study, a hydrogen and electric co-production plant has been designed. The reactor thermal output is 375 MW and 25% of the thermal output is used for hydrogen production (70000 Nm 3 /h). The hydrogen production cost is estimated to 21 yen/Nm 3 but it is still higher than the economical goal (17 yen/Nm 3 ). The major reason of the high cost comes from the large size of hydrogen separation reformers because of the limit of hydrogen separation efficiency of palladium membrane. A new highly efficient hydrogen separation membrane is needed to reduce the cost of hydrogen production using membrane reformers. There is possibility of multi-tube failure in the membrane reformers. In future study, a design of measures against tube failure and elemental experiments of reaction between sodium and reforming gas will be needed. (authors)

  3. Advanced multistage turbine blade aerodynamics, performance, cooling, and heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Fleeter, S.; Lawless, P.B. [Purdue Univ., West Lafayette, IN (United States)

    1995-10-01

    The gas turbine has the potential for power production at the highest possible efficiency. The challenge is to ensure that gas turbines operate at the optimum efficiency so as to use the least fuel and produce minimum emissions. A key component to meeting this challenge is the turbine. Turbine performance, both aerodynamics and heat transfer, is one of the barrier advanced gas turbine development technologies. This is a result of the complex, highly three-dimensional and unsteady flow phenomena in the turbine. Improved turbine aerodynamic performance has been achieved with three-dimensional highly-loaded airfoil designs, accomplished utilizing Euler or Navier-Stokes Computational Fluid Dynamics (CFD) codes. These design codes consider steady flow through isolated blade rows. Thus they do not account for unsteady flow effects. However, unsteady flow effects have a significant impact on performance. Also, CFD codes predict the complete flow field. The experimental verification of these codes has traditionally been accomplished with point data - not corresponding plane field measurements. Thus, although advanced CFD predictions of the highly complex and three-dimensional turbine flow fields are available, corresponding data are not. To improve the design capability for high temperature turbines, a detailed understanding of the highly unsteady and three-dimensional flow through multi-stage turbines is necessary. Thus, unique data are required which quantify the unsteady three-dimensional flow through multi-stage turbine blade rows, including the effect of the film coolant flow. This requires experiments in appropriate research facilities in which complete flow field data, not only point measurements, are obtained and analyzed. Also, as design CFD codes do not account for unsteady flow effects, the next logical challenge and the current thrust in CFD code development is multiple-stage analyses that account for the interactions between neighboring blade rows.

  4. Fuel cycles and advanced core designs for the Gas-Cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Simon, R.H.; Hamilton, C.J.; Hunter, R.S.

    1982-01-01

    Studies indicate that a 1200 MW(e) Gas-Cooled Fast Breeder Reactor could achieve compound system doubling times of under ten years when using advanced oxide or carbide fuels. In addition, when thorium is used in the breeding blankets, enough U-233 can be generated in each GCFR to supply several advanced converter reactors with fissionable material and this symbiotic relationship could provide energy for the world for centuries. (author)

  5. Thiazolidinediones and Edema: Recent Advances in the Pathogenesis of Thiazolidinediones-Induced Renal Sodium Retention

    Directory of Open Access Journals (Sweden)

    Shoko Horita

    2015-01-01

    Full Text Available Thiazolidinediones (TZDs are one of the major classes of antidiabetic drugs that are used widely. TZDs improve insulin resistance by activating peroxisome proliferator-activated receptor gamma (PPARγ and ameliorate diabetic and other nephropathies, at least, in experimental animals. However, TZDs have side effects, such as edema, congestive heart failure, and bone fracture, and may increase bladder cancer risk. Edema and heart failure, which both probably originate from renal sodium retention, are of great importance because these side effects make it difficult to continue the use of TZDs. However, the pathogenesis of edema remains a matter of controversy. Initially, upregulation of the epithelial sodium channel (ENaC in the collecting ducts by TZDs was thought to be the primary cause of edema. However, the results of other studies do not support this view. Recent data suggest the involvement of transporters in the proximal tubule, such as sodium-bicarbonate cotransporter and sodium-proton exchanger. Other studies have suggested that sodium-potassium-chloride cotransporter 2 in the thick ascending limb of Henle and aquaporins are also possible targets for TZDs. This paper will discuss the recent advances in the pathogenesis of TZD-induced sodium reabsorption in the renal tubules and edema.

  6. Thiazolidinediones and Edema: Recent Advances in the Pathogenesis of Thiazolidinediones-Induced Renal Sodium Retention.

    Science.gov (United States)

    Horita, Shoko; Nakamura, Motonobu; Satoh, Nobuhiko; Suzuki, Masashi; Seki, George

    2015-01-01

    Thiazolidinediones (TZDs) are one of the major classes of antidiabetic drugs that are used widely. TZDs improve insulin resistance by activating peroxisome proliferator-activated receptor gamma (PPARγ) and ameliorate diabetic and other nephropathies, at least, in experimental animals. However, TZDs have side effects, such as edema, congestive heart failure, and bone fracture, and may increase bladder cancer risk. Edema and heart failure, which both probably originate from renal sodium retention, are of great importance because these side effects make it difficult to continue the use of TZDs. However, the pathogenesis of edema remains a matter of controversy. Initially, upregulation of the epithelial sodium channel (ENaC) in the collecting ducts by TZDs was thought to be the primary cause of edema. However, the results of other studies do not support this view. Recent data suggest the involvement of transporters in the proximal tubule, such as sodium-bicarbonate cotransporter and sodium-proton exchanger. Other studies have suggested that sodium-potassium-chloride cotransporter 2 in the thick ascending limb of Henle and aquaporins are also possible targets for TZDs. This paper will discuss the recent advances in the pathogenesis of TZD-induced sodium reabsorption in the renal tubules and edema.

  7. The Advancement of Cool Roof Standards in China from 2010 to 2015

    Energy Technology Data Exchange (ETDEWEB)

    Ge, Jing [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Levinson, Ronnen M. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

    2016-11-01

    Since the initiation of the U.S.-China Clean Energy Research Center-Building Energy Efficiency (CERC-BEE) cool roof research collaboration between the Lawrence Berkeley National Laboratory Heat Island Group and Chinese institutions in 2010, new cool surface credits (insulation trade- offs) have been adopted in Chinese building energy efficiency standards, industry standards, and green building standards. JGJ 75-2012: Design Standard for Energy Efficiency of Residential Buildings in Hot Summer and Warm Winter Zone became the first national level standard to provide cool surface credits. GB/T 50378-2014: Assessment Standard for Green Building is the first national level green building standard that offers points for heat island mitigation. JGJ/T 359-2015: Technical Specification for Application of Architectural Reflective Thermal Insulation Coating is the first industry standard that offers cool coating credits for both public and residential buildings in all hot-summer climates (Hot Summer/Cold Winter, Hot Summer/Warm Winter). As of December 2015, eight provinces or municipalities in hot-summer regions have credited cool surfaces credits in their residential and/or public building design standards; five other provinces or municipalities in hot-summer regions recommend, but do not credit, the use of cool surfaces in their building design standards. Cool surfaces could be further advanced in China by including cool roof credits for residential and public building energy efficiency standards in all hot-summer regions; developing a standardized process for natural exposure and aged-property rating of cool roofing products; and adapting the U.S.-developed laboratory aging process for roofing materials to replicate solar reflectance changes induced by natural exposure in China.

  8. A neutronics study for improving the safety and performance parameters of a 3600 MWth Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Sun, Kaichao; Krepel, Jiri; Mikityuk, Konstantin; Chawla, Rakesh

    2013-01-01

    Highlights: ► The potential for neutronics design optimization is assessed for a large SFR core. ► Both beginning-of-life and equilibrium fuel cycle conditions are considered. ► The sodium void effect is decomposed via a neutron balance based methodology. ► The optimized core options adopt an appropriate sodium plenum design to reduce the void effect. ► The introduction of moderator pins is considered for enhancing the Doppler effect. - Abstract: The Sodium-cooled Fast Reactor (SFR) is one of the most promising Generation IV systems with many performance advantages, but has one dominating neutronics drawback – a positive sodium void reactivity. The starting point for the present study is an SFR core design considered in the Collaborative Project on the European Sodium-cooled Fast Reactor (CP-ESFR). The aim is to analyze, for this reference core, four safety and performance parameters from the viewpoint of four different optimization options, and to propose possible optimized core designs. In doing so, the study focuses not only on the beginning-of-life state of the core, but also on the beginning of equilibrium closed fuel cycle. The four studied optimization options are: (a) introducing an upper sodium plenum and boron layer, (b) varying the core height-to-diameter (H/D) ratio, (c) introducing moderator pins into the fuel assembly, and (d) modifying the initial plutonium content. The sensitivity of the void reactivity, Doppler constant, nominal reactivity and breeding gain has been evaluated. In particular, the void reactivity, which is the most crucial safety parameter for the SFR, has been decomposed into its reaction-wise, isotope-wise and energy-group-wise components using a methodology based on the neutron balance equation. Extended voiding in the upper sodium plenum region – in conjunction with the effect of a boron layer introduced above the plenum – is found to be particularly effective in the void effect reduction while, at the same time

  9. Proposals for in-service inspection and monitoring of selected components located within or part of the primary containment of sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Bolt, P.R.

    1976-01-01

    Design and operational experience of CEGB gas cooled reactors and certain overseas reactor plant is reviewed in relation to in-service inspection and monitoring capabilities. Design guidelines and preliminary proposals are given for in-service inspection and monitoring of selected components located within or part of the primary containment of sodium cooled fast reactors. Specific comments are made on the items of further design and development work believed to be necessary

  10. IAEA activities in technology development for advanced water-cooled nuclear power plants

    International Nuclear Information System (INIS)

    Juhn, Poong Eil; Kupitz, Juergen; Cleveland, John; Lyon, Robert; Park, Je Won

    2003-01-01

    As part of its Nuclear Power Programme, the IAEA conducts activities that support international information exchange, co-operative research and technology assessments and advancements with the goal of improving the reliability, safety and economics of advanced water-cooled nuclear power plants. These activities are conducted based on the advice, and with the support, of the IAEA Department of Nuclear Energy's Technical Working Groups on Advanced Technologies for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs). Assessments of projected electricity generation costs for new nuclear plants have shown that design organizations are challenged to develop advanced designs with lower capital costs and short construction times, and sizes, including not only large evolutionary plants but also small and medium size plants, appropriate to grid capacity and owner financial investment capability. To achieve competitive costs, both proven means and new approaches should be implemented. The IAEA conducts activities in technology development that support achievement of improved economics of water-cooled nuclear power plants (NPPs). These include fostering information sharing and cooperative research in thermo-hydraulics code validation; examination of natural circulation phenomena, modelling and the reliability of passive systems that utilize natural circulation; establishment of a thermo-physical properties data base; improved inspection and diagnostic techniques for pressure tubes of HWRs; and collection and balanced reporting from recent construction and commissioning experiences with evolutionary water-cooled NPPs. The IAEA also periodically publishes Status Reports on global development of advanced designs. (author)

  11. Effects of duct configuration on flow and temperature structure in sodium-cooled 19-rod simulated LMFBR fuel bundles with helical wire-wrap spacers

    International Nuclear Information System (INIS)

    Wantland, J.L.; Fontana, M.H.; Gnadt, P.A.; Hanus, N.; MacPherson, R.E.; Smith, C.M.

    1976-01-01

    Thermal-hydrodynamic testing of sodium-cooled 19-rod simulated LMFBR fuel bundles is being conducted at the O ak Ridge National Laboratory in the Fuel Failure Mockup (FFM), an engineering-scale high-temperature sodium facility which provides prototypic flows, temperatures and power densities. Electrically heated bundles have been tested with two scalloped and two hexagonal duct configurations. Peripheral helical flows, attributed to the spacers, have been observed with strengths dependent upon the evenness and relative sizes of the peripheral flow areas. Diametral sodium temperature profiles are more uniform with smaller peripheral flow areas

  12. Study on external reactor vessel cooling capacity for advanced large size PWR

    International Nuclear Information System (INIS)

    Jin Di; Liu Xiaojing; Cheng Xu; Li Fei

    2014-01-01

    External reactor vessel cooling (ERVC) is widely adopted as a part of in- vessel retention (IVR) in severe accident management strategies. In this paper, some flow parameters and boundary conditions, eg., inlet and outlet area, water inlet temperature, heating power of the lower head, the annular gap size at the position of the lower head and flooding water level, were considered to qualitatively study the effect of them on natural circulation capacity of the external reactor vessel cooling for an advanced large size PWR by using RELAP5 code. And the calculation results provide some basis of analysis for the structure design and the following transient response behavior of the system. (authors)

  13. sodium

    International Development Research Centre (IDRC) Digital Library (Canada)

    Les initiatives de réduction de la consommation de sel qui visent l'ensemble de la population et qui ciblent la teneur en sodium des aliments et sensibilisent les consommateurs sont susceptibles de réduire la consommation de sel dans toutes les couches de la population et d'améliorer la santé cardiovasculaire. Ce projet a ...

  14. Future work in the DeBeNeLux research centres on the sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Goedkoop, J.A.

    1976-01-01

    The general objectives as they now apply over the world in the further development of the sodium cooled fast reactor are to realize a reactor and the associated fuel cycle, that will ensure a good fuel utilization; secondly, as long as we live in a more or less free market economy, such a system will only be acceptable if it is competitive, which means that the difference in investment cost between the fast reactor and the presently used light water reactors has to be brought down; thirdly, to justify the investment the system should work reliably; finally the developments in reactor design should not be at the expense of reactor safety. The pursuit of these objectives during the coming years will require the DeBeNeLuX laboratories to do work in a number of fields. (Auth.)

  15. Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

    Energy Technology Data Exchange (ETDEWEB)

    Kuk, Seoung Woo, E-mail: swkuk@kaeri.re.kr [Next Generation Fuel Development Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock [Next Generation Fuel Development Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Youn, Young-Sang [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Kim, Jong-Yun [Nuclear Chemistry Research Division, Korea Atomic Energy Research Institute, Daedeok-daero 989-111, Yuseong-gu, Daejeon, 34057 (Korea, Republic of); Radiochemistry & Nuclear Nonproliferation, University of Science & Technology, Gajeong-ro 217, Yuseong-gu, Daejeon, 34113 (Korea, Republic of)

    2017-04-01

    Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

  16. Experimental determination of temperature fields in sodium-cooled rod bundles with hexagonal rod arrangement and grid spacers

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.; Kolodziej, M.

    1977-01-01

    Three-dimensional temperature fields in the claddings of sodium cooled rods were determined experimentally under representative nominal operating conditions for a SNR typical 19-rod bundle model provided with spark-eroded spacers. These experiments are required to verify thermohydraulic computer programs which will provide the output data for strength calculations of the high loaded cladding tubes. In this work the essentials are reported of the measured circumferential distributions of wall temperatures of peripheral rods. In addition the sub-channel temperatures measured over the bundle cross section are indicated, they are required to sustain codes for the global thermohydraulic design of core elements. The most important results are: 1) The whole fuel element is located within the thermal entrance length. 2) High azimuthal temperature differences were measured in the claddings of peripheral rods, which are strongly influenced by the distance between the rod and the shroud, especially for the corner rod. 3) With decreasing Pe-number ( [de

  17. THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

    Directory of Open Access Journals (Sweden)

    MEHMET E. KORKMAZ

    2014-06-01

    Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

  18. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-01

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean

  19. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  20. Sodium-immersed self-cooled electromagnetic pump design and development of a large-scale coil for high temperature

    International Nuclear Information System (INIS)

    Oto, Akihiro; Naohara, Nobuyuki; Ishida, Masayoshi; Katsuki, Kenji; Kumazawa, Ryouji

    1995-01-01

    A sodium-immersed, self-cooled electromagnetic (EM) pump was recently studied as a prospective innovative technology to simplify a fast breeder reactor plant system. The EM pump for a primary pump, a pump type, was designed, and the structural concept and the system performance were clarified. For the flow control method, a constant voltage/frequency method was preferable from the point of view of pump performance and efficiency. The insulation life was tested on a large-scale coil at high temperature as part of the development of a large-capacity EM pump. Mechanical and electrical damage were not observed, and the insulation performance was quite good. The insulation system could also be applied to large-scale coils

  1. Three-dimensional tsunami analysis for the plot plan of a sodium-cooled fast reactor plant

    International Nuclear Information System (INIS)

    Hayakawa, Satoshi; Watanabe, Osamu; Itoh, Kei; Yamamoto, Tomohiko

    2013-01-01

    As the practical evaluation method of the effect of tsunami on buildings, the formula of tsunami force has been used. However, it cannot be applied to complex geometry of buildings. In this study, to analyze the effect of tsunami on the buildings of sodium-cooled fast reactor plant more accurately, three-dimensional tsunami analysis was performed. In the analysis, VOF (Volume of Fluid) method was used to capture free surface of tsunami. At the beginning, it was confirmed that the tsunami experiment results was reproduced by VOF method accurately. Next, the three-dimensional tsunami analysis was performed with VOF method to evaluate the flow field around the buildings of the plant from the beginning of the tsunami until the backwash of that. (author)

  2. Thermodynamic Data to Model the Interaction Between Coolant and Fuel in Gen IV Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Dinsdale, Alan; Gisby, John; Davies, Hugh; Konings, Rudy; Benes, Ondrej

    2013-06-01

    Understanding the behaviour of nuclear fuels in various environments is vital to the design and safe operation of nuclear reactors. While this is true if the reactor is operating within its design specification, it is even more so if accidents occur and the fuel is exposed to unexpected temperatures, pressures or chemical environments. It is clearly hazardous and costly to explore all such scenarios experimentally and therefore it is necessary to undertake modelling where possible using well-grounded theoretical approaches. This paper will show examples of where calculations of chemical and phase equilibria have been applied successfully to the long term storage of nuclear waste, phase formation during core meltdown and prediction of fission product release into the atmosphere. It will also highlight the development of thermodynamic data carried out during the European Metrology Research Project Metrofission required to model the potential interaction between the coolant, nuclear fuel, containment materials and atmosphere of a sodium cooled fast reactor. (authors)

  3. Transport of radioactive corrosion products in primary system of sodium-cooled fast breeder reactor 'MONJU'

    International Nuclear Information System (INIS)

    Matuo, Youichirou; Hasegawa, Masanori; Maegawa, Yoshiharu; Miyahara, Shinya

    2011-01-01

    Radioactive corrosion products (CP) are primary cause of personal radiation exposure during maintenance work at FBR plants with no breached fuel. The PSYCHE code has been developed based on the Solution-Precipitation model for analysis of CP transfer behavior. We predicted and analyzed the CP solution and precipitation behavior of MONJU to evaluate the applicability of the PSYCHE code to MONJU, using the parameters verified in the calculations for JOYO. From the calculation result pertaining to the MONJU system, distribution of 54 Mn deposited in the primary cooling system over 20 years of operation is predicted to be approximately 7 times larger than that of 60 Co. In particular, predictions show a notable tendency for 54 Mn precipitation to be distributed in the primary pump and cold-leg. The calculated distribution of 54 Mn and 60 Co in the primary cooling system of MONJU agreed with tendencies of measured distribution of JOYO. (author)

  4. A statistical analysis on failure-to open/close probability of pneumatic valve in sodium cooling systems

    International Nuclear Information System (INIS)

    Kurisaka, Kenichi

    1999-11-01

    The objective of this study is to develop fundamental data for examination on efficiency of preventive maintenance and surveillance test from the standpoint of failure probability. In this study, as a major standby component, a pneumatic valve in sodium cooling systems was selected. A statistical analysis was made about a trend of valve in sodium cooling systems was selected. A statistical analysis was made about a trend of valve failure-to-open/close (FTOC) probability depending on number of demands ('n'), time since installation ('t') and standby time since last open/close action ('T'). The analysis is based on the field data of operating- and failure-experiences stored in the Component Reliability Database and Statistical Analysis System for LMFBR's (CORDS). In the analysis, the FTOC probability ('P') was expressed as follows: P=1-exp{-C-En-F/n-λT-aT(t-T/2)-AT 2 /2}. The functional parameters, 'C', 'E', 'F', 'λ', 'a' and 'A', were estimated with the maximum likelihood estimation method. As a result, the FTOC probability is almost expressed with the failure probability being derived from the failure rate under assumption of the Poisson distribution only when valve cycle (i.e. open-close-open cycle) exceeds about 100 days. When the valve cycle is shorter than about 100 days, the FTOC probability can be adequately estimated with the parameter model proposed in this study. The results obtained from this study may make it possible to derive an adequate frequency of surveillance test for a given target of the FTOC probability. (author)

  5. JSFR design progress related to development of safety design criteria for Generation IV sodium-cooled fast reactors. (1) Overview

    International Nuclear Information System (INIS)

    Kamide, Hideki; Ando, Masato; Ito, Takaya

    2015-01-01

    JAEA, JAPC and MFBR have been conducting design study for the Japan Sodium-cooled Fast Reactor (JSFR), which is a design concept aiming at future commercial use as sustainable electric power source. As the result of the design study and R and D activity related the innovative technologies incorporated in the design in the Fast Reactor Cycle Technology Development (FaCT) project up to 2010, basic design concept of JSFR was established and its development process to the commercialization including construction and operation of a demonstration version of JSFR was outlined. JSFR is a looptype next generation sodium-cooled fast reactor (SFR), which is aiming at achieving development targets of Generation IV reactors concerning sustainability, safety and reliability, economics and proliferation resistance and physical protection by introducing the innovative technologies such as shortened high-chromium steel piping. The output power is assumed for the design study as 1,500 MWe for the commercial version and 750 MWe for the demonstration version. In FaCT phase I up to 2010, in order to evaluate feasibility to achieve the development targets, the design study has been conducted on the main components and systems. Since 2011, in order to contribute to the development of safety design criteria (SDC) and safety design guideline (SDG), which include the lessons learned from the TEPCO's Fukushima Dai-ichi nuclear power plants accident, in the frame work of Generation IV International Forum (GIF), the design study is focusing on the design measures against severe external events such as earthquake and tsunami. At the same time, the design study is going into detail and paying much attention to the maintenance and repair to make surer its feasibility. This paper summarizes the design concept of the demonstration version of JSFR in which progress of design work was incorporated for the safety issues on SDC and SDG of a SFR. (author)

  6. Development of the evaluation methodology for the material relocation behavior in the core disruptive accident of sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Tobita, Yoshiharu; Kamiyama, Kenji; Tagami, Hirotaka; Matsuba, Ken-ichi; Suzuki, Tohru; Isozaki, Mikio; Yamano, Hidemasa; Morita, Koji; Guo, LianCheng; Zhang, Bin

    2016-01-01

    The in-vessel retention (IVR) of core disruptive accident (CDA) is of prime importance in enhancing safety characteristics of sodium-cooled fast reactors (SFRs). In the CDA of SFRs, molten core material relocates to the lower plenum of reactor vessel and may impose significant thermal load on the structures, resulting in the melt-through of the reactor vessel. In order to enable the assessment of this relocation process and prove that IVR of core material is the most probable consequence of the CDA in SFRs, a research program to develop the evaluation methodology for the material relocation behavior in the CDA of SFRs has been conducted. This program consists of three developmental studies, namely the development of the analysis method of molten material discharge from the core region, the development of evaluation methodology of molten material penetration into sodium pool, and the development of the simulation tool of debris bed behavior. The analysis method of molten material discharge was developed based on the computer code SIMMER-III since this code is designed to simulate the multi-phase, multi-component fluid dynamics with phase changes involved in the discharge process. Several experiments simulating the molten material discharge through duct using simulant materials were utilized as the basis of validation study of the physical models in this code. It was shown that SIMMER-III with improved physical models could simulate the molten material discharge behavior, including the momentum exchange with duct wall and thermal interaction with coolant. In order to develop an evaluation methodology of molten material penetration into sodium pool, a series of experiments simulating jet penetration behavior into sodium pool in SFR thermal condition were performed. These experiments revealed that the molten jet was fragmented in significantly shorter penetration length than the prediction by existing correlation for light water reactor conditions, due to the direct

  7. Mathematical modelling of performance of safety rod and its drive mechanism in sodium cooled fast reactor during scram action

    International Nuclear Information System (INIS)

    Rajan Babu, V.; Thanigaiyarasu, G.; Chellapandi, P.

    2014-01-01

    Highlights: • Mathematical modelling of dynamic behaviour of safety rod during scram action in fast reactor. • Effects of hydraulics, structural interaction and geometry on drop time of safety rod are understood. • Using simplified model, drop time can be assessed replacing detailed CFD analysis. • Sensitivities of the related parameters on drop time are understood. • Experimental validation qualifies the modelling and computer software developed. - Abstract: Performance of safety rod and its drive mechanism which are parts of shutdown systems in sodium cooled fast reactor (SFR) plays a major role in ensuring safe operation of the plant during all the design basis events. The safety rods are to be inserted into the core within a stipulated time during off-normal conditions of the reactor. Mathematical modelling of dynamic behaviour of a safety rod and its drive mechanism in a typical 500 MWe SFR during scram action is considered in the present study. A full-scale prototype system has undergone qualification tests in air, water and in sodium simulating the operating conditions in the reactor. In this paper, the salient features of the safety rod and its mechanism, details related to mathematical modelling and sensitivity of the parameters having influence on drop time are presented. The outcomes of the numerical analysis are compared with the experimental results. In this process, the mathematical model and the computer software developed are validated

  8. Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization

    Directory of Open Access Journals (Sweden)

    Franck Dechelette

    2014-01-01

    Full Text Available The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of three SFR fuel handling route fully described and compared to a reference FHR option. In those three FHR, two use a gas corridor to transfer spent and fresh fuel assembly and the third uses two casks with a sodium pot to evacuate and load an assembly in parallel. All of them are designed for the ASTRID reactor (1500 MWth but can be extrapolated to power reactors and are compatible with the mutualisation of one FHS coupled with two reactors. These three concepts are then intercompared and evaluated with the reference FHR according to four criteria: performances, risk assessment, investment cost, and qualification time. This analysis reveals that the “mixed way” FHR presents interesting solutions mainly in terms of design simplicity and time reduction. Therefore its study will be pursued for ASTRID as an alternative option.

  9. A two-dimensional model for transients calculations with phase changes in sodium cooled reactors

    International Nuclear Information System (INIS)

    Granziera, M.R.

    1981-01-01

    A computer code (NATOF2D) for the numerical simulation of situations where the radial non-uniformity in the sodium flow is an important factor, was developed. This computer code uses the two-fluid model, in which each phase is described by a complete set of mass conservation equations, energy equations and momentum equations. The experiment SLSF-P3A realized in the Engineering Test Reactor, Idaho, during the period of july to september of 1977, was simulated. (E.G.) [pt

  10. Problems specific to the piping of sodium-cooled fast breeder reactors

    International Nuclear Information System (INIS)

    Vrillon, B.; Befre, J.; Schaller, K.

    1975-01-01

    A certain number of specific problems arising in connection with the sodium pipes in fast neutron reactors, especially those of large diameter, are presented. The supporting system must be designed to achieve the best compromise among stresses due to weight and various stresses of thermal origin. Large-scale experimental studies carried out on actual elements of the intermediate circuit of the Phenix reactor showed that the circuits can withstand considerable deformation collapse of the walls without danger of leakage. Protection studies against earthquakes are mentionned [fr

  11. Experimental Flow Performance Evaluation of novel miniaturized Advanced Piezoelectric Dual Cooling Jet

    International Nuclear Information System (INIS)

    De Bock, H P J; Whalen, B P; Chamarthy, P; Jackson, J L

    2012-01-01

    In recent years, electronics systems have significantly reduced in size at maintained or increased functionality. This trend has led to an increased demand for smaller and more capable thermal management. However, miniaturization of conventional fan and heat sink cooling systems introduce significant size, weight and efficiency challenges. In this study the flow performance of a novel alternative thin form-factor cooling solution, the advanced piezoelectric dual cooling jet(DCJ), is evaluated. A DCJ is a system where two piezoelectric actuators are excited to produce air flow. The total height of the device is about 1mm. The design of the experimental method for evaluating the equivalent fan-curve of the DCJ device is described in detail. Experimental results in comparison to conventional fan solutions are provided. The DCJ is expected to be a good candidate for thermal management in next generation thin profile consumer electronics.

  12. Wave propagation simulation in the upper core of sodium-cooled fast reactors using a spectral-element method for heterogeneous media

    Science.gov (United States)

    Nagaso, Masaru; Komatitsch, Dimitri; Moysan, Joseph; Lhuillier, Christian

    2018-01-01

    ASTRID project, French sodium cooled nuclear reactor of 4th generation, is under development at the moment by Alternative Energies and Atomic Energy Commission (CEA). In this project, development of monitoring techniques for a nuclear reactor during operation are identified as a measure issue for enlarging the plant safety. Use of ultrasonic measurement techniques (e.g. thermometry, visualization of internal objects) are regarded as powerful inspection tools of sodium cooled fast reactors (SFR) including ASTRID due to opacity of liquid sodium. In side of a sodium cooling circuit, heterogeneity of medium occurs because of complex flow state especially in its operation and then the effects of this heterogeneity on an acoustic propagation is not negligible. Thus, it is necessary to carry out verification experiments for developments of component technologies, while such kind of experiments using liquid sodium may be relatively large-scale experiments. This is why numerical simulation methods are essential for preceding real experiments or filling up the limited number of experimental results. Though various numerical methods have been applied for a wave propagation in liquid sodium, we still do not have a method for verifying on three-dimensional heterogeneity. Moreover, in side of a reactor core being a complex acousto-elastic coupled region, it has also been difficult to simulate such problems with conventional methods. The objective of this study is to solve these 2 points by applying three-dimensional spectral element method. In this paper, our initial results on three-dimensional simulation study on heterogeneous medium (the first point) are shown. For heterogeneity of liquid sodium to be considered, four-dimensional temperature field (three spatial and one temporal dimension) calculated by computational fluid dynamics (CFD) with Large-Eddy Simulation was applied instead of using conventional method (i.e. Gaussian Random field). This three-dimensional numerical

  13. Leakage limits for inflatable seals of sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, N.K., E-mail: nksinha@igcar.gov.in; Raj, Baldev

    2014-01-15

    Highlights: • All possible types/modes of gas escape covered. • Limits include simultaneous contributions from bypass and permeation leakage modes. • Leakage of radioactive cover gas with fission products assumed. • Possibility of sodium frost deposition in sealed gap considered. • Cover gas activity decay during fuel handling and relative importance of types/modes of leakage considered for realistic results and simpler seal design. -- Abstract: Estimation and stipulation of allowable leakage for inflatable seals of 500 MWe Prototype Fast Breeder Reactor is depicted. Leakage limits are specified using a conservative approach, which assumes escape of radioactive cover gas with fission products across the seals in bypass and permeation modes and possibility of sodium frost deposition in sealed gaps because of permeation leakage of inflation gas. Procedures to arrive at the allowable leakages of argon cover gas (normal-operation/fuel-handling: 10{sup −3}/10{sup −2} scc/s/m length of seal) and argon inflation gas (10{sup −3} scc/s/m length of seal) is described.

  14. Applicability of RELAP5-3D for Thermal-Hydraulic Analyses of a Sodium-Cooled Actinide Burner Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    C. B. Davis

    2006-07-01

    The Actinide Burner Test Reactor (ABTR) is envisioned as a sodium-cooled, fast reactor that will burn the actinides generated in light water reactors to reduce nuclear waste and ease proliferation concerns. The RELAP5-3D computer code is being considered as the thermal-hydraulic system code to support the development of the ABTR. An evaluation was performed to determine the applicability of RELAP5-3D for the analysis of a sodium-cooled fast reactor. The applicability evaluation consisted of several steps, including identifying the important transients and phenomena expected in the ABTR, identifying the models and correlations that affect the code’s calculation of the important phenomena, and evaluating the applicability of the important models and correlations for calculating the important phenomena expected in the ABTR. The applicability evaluation identified code improvements and additional models needed to simulate the ABTR. The accuracy of the calculated thermodynamic and transport properties for sodium was also evaluated.

  15. Innovate pin design for Sphere-pac fuel in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Pouchon, Manuel A.; Niceno, Bojan; Krepel, Jiri

    2011-01-01

    The paper discusses a new fuel element type, which combines a particle fuel concept, the Sphere-pac, with a new pin design which features internal cooling. Particle fuels are auspicious when considering a closed fuel cycle, where minor actinide containing fuels must be fabricated. The principle advantage lies in their production simplicity with much less maintenance intensive mechanical devices. Furthermore the Sphere-pac is usually produced by a wet and therefore powder-less route. Therefore the implementation in a remotely controlled and heavily shielded environment becomes easier to realize. Besides the advantages in the production process, the Sphere-pac bears one important disadvantage: the lower thermal conductivity of the particle arrangement, and the therefore higher peak temperatures in the fuel. Consequently a new fuel design is suggested in this paper. It offers an internal cooling channel and therefore smaller maximal fuel distances to the coolant. As the concept is new, the most important aspects are studied; these are the neutronics, the temperature profile in the fuel plus thermal-hydraulics aspects. (author)

  16. Optimization of advanced gas-cooled reactor fuel performance by a stochastic method

    International Nuclear Information System (INIS)

    Parks, G.T.

    1987-01-01

    A brief description is presented of a model representing the in-core behaviour of a single advanced gas-cooled reactor fuel channel, developed specifically for optimization studies. The performances of the only suitable Numerical Algorithms Group (NAG) library package and a Metropolis algorithm routine on this problem are discussed and contrasted. It is concluded that, for the problem in question, the stochastic Metropolis algorithm has distinct advantages over the deterministic NAG routine. (author)

  17. A safety design approach for sodium cooled fast reactor core toward commercialization in Japan

    International Nuclear Information System (INIS)

    Kubo, Shigenobu

    2012-01-01

    JAEA’s safety approach for SFR core design is based on defence‐in‐depth concept, which includes DBAs and DECs (prevention and mitigation): • The reactor core is designed to have inherent reactivity feedback characteristics with negative power coefficient. • Operation temperature range is set sufficiently below the coolant boiling temperature so as to avoid coolant boiling against anticipated operational occurrences and DBAs. • If the plant state deviates from operational states, the safe reactor shutdown is achieved by automatic insertion of control rods. 2 active reactor shutdown systems are provided. • Failure of active reactor shutdown is assumed in a design extension condition . Passive shutdown capability is provided by SASS under such condition. • As a design extension condition, core disruptive accident is assumed. In order to prevent severe mechanical energy release which might cause containment function failure, core sodium void worth is limited below 6 dollars and molten fuel discharge capability is utilized by FAIDUS. (author)

  18. Passive acoustic leak detection for sodium cooled fast reactors using hidden Markov models

    Energy Technology Data Exchange (ETDEWEB)

    Riber Marklund, A. [CEA, Cadarache, DEN/DTN/STCP/LIET, Batiment 202, 13108 St Paul-lez-Durance, (France); Kishore, S. [Fast Reactor Technology Group of IGCAR, (India); Prakash, V. [Vibrations Diagnostics Division, Fast Reactor Technology Group of IGCAR, (India); Rajan, K.K. [Fast Reactor Technology Group and Engineering Services Group of IGCAR, (India)

    2015-07-01

    Acoustic leak detection for steam generators of sodium fast reactors have been an active research topic since the early 1970's and several methods have been tested over the years. Inspired by its success in the field of automatic speech recognition, we here apply hidden Markov models (HMM) in combination with Gaussian mixture models (GMM) to the problem. To achieve this, we propose a new feature calculation scheme, based on the temporal evolution of the power spectral density (PSD) of the signal. Using acoustic signals recorded during steam/water injection experiments done at the Indira Gandhi Centre for Atomic Research (IGCAR), the proposed method is tested. We perform parametric studies on the HMM+GMM model size and demonstrate that the proposed method a) performs well without a priori knowledge of injection noise, b) can incorporate several noise models and c) has an output distribution that simplifies false alarm rate control. (authors)

  19. Minor Actinide Recycle in Sodium Cooled Fast Reactors Using Heterogeneous Targets

    International Nuclear Information System (INIS)

    Bays, Samuel; Medvedev, Pavel; Pope, Michael; Ferrer, Rodolfo; Forget, Benoit; Asgari, Mehdi

    2009-01-01

    This paper investigates the plausible design of transmutation target assemblies for minor actinides (MA) in Sodium Fast Reactors (SFR). A heterogeneous recycling strategy is investigated, whereby after each reactor pass, un-burned MAs from the targets are blended with MAs produced by the driver fuel and additional MAs from Spent Nuclear Fuel (SNF). A design iteration methodology was adopted for customizing the core design, target assembly design and matrix composition design. The overall design was constrained against allowable peak or maximum in-core performances. While respecting these criteria, the overall design was adjusted to reduce the total number of assemblies fabricated per refueling cycle. It was found that an inert metal-hydride MA-Zr-Hx target matrix gave the highest transmutation efficiency, thus allowing for the least number of targets to be fabricated per reactor cycle.

  20. A design study of high breeding ratio sodium cooled metal fuel core without blanket fuels

    International Nuclear Information System (INIS)

    Kobayashi, Noboru; Ogawa, Takashi; Ohki, Shigeo; Mizuno, Tomoyasu; Ogata, Takanari

    2009-01-01

    The metal fuel core is superior to the mixed oxide fuel core because of its high breeding ratio and compact core size resulting from hard neutron spectrum and high heavy metal densities. Utilizing these characteristics, a conceptual design for a high breeding ratio was performed without blanket fuels. The design conditions were set so a sodium void worth of less than 8 $, a core height of less than 150 cm, the maximum cladding temperature of 650degC, and the maximum fuel pin bundle pressure drop of 0.4 MPa. The breeding ratio of the resultant core was 1.34 with 6wt% zirconium content fuel. Applying 3wt% zirconium content fuel enhanced the breeding ratio up to 1.40. (author)

  1. Materials problems related to the core catcher of sodium cooled reactors

    International Nuclear Information System (INIS)

    Goetzmann, O.

    1975-05-01

    There are in principal two possible solutions for the external core catcher as far as materials are concerned. 1) A barrier consisting of a material with a high melting point, 2) a tray of comparatively low melting material with a high solubility for the fuel. In case of the first concept one has to look for materials whose melting temperatures are above the temperature of the molten core. Based on metallurgical reasons it seems very likely that the molten core does not exceed a temperature in the range between 2,500 and 2,800 0 C. Due to the compatibility situation with the molten core only a few high melting oxides will be suitable as liner materials for a core catcher. In the second case basalt or concrete, if free of water and lime, are suitable materials. Graphite is a high melting material, however, due to its behaviour with the molten core it should be listed under the second group. By the reaction of graphite with the core materials the melt can be kept liquid down to temperatures of around 1,100 0 C. The evolution of CO by this reaction should be supportable. It is an endothermal reaction. Experiments on the behaviour of core catcher materials have shown that sodium is capable of penetrating into sintered bodies of UO 2 with densities of 90% TD at temperatures higher than 200 0 C. This may lead to the desintegration of these bodies. The exposure to moist air has not done much harm to UO 2 pellets of densities from 80 to 90% TD. Even after one year of exposure, swelling or desintegration could not be observed. Sodium is also capable of penetrating into bodies of synthetic carbon and graphite. Only well graphitized material will not be destroyed. (orig.) [de

  2. Numerical simulation for debris bed behavior in sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Tagami, Hirotaka; Tobita, Yoshiharu

    2014-01-01

    For safety analysis of SFR, it is necessary to evaluate behavior along with coolability of debris bed in lower plenum which is formed in severe accident. In order to analyze debris behavior, model for dense sediment particles behavior was proposed and installed in SFR safety analysis code SIMMER. SIMMER code could adequately reproduce experimental results simulating the self-leveling phenomena with appropriate model parameters for bed stiffness. In reactor condition, the self-leveling experiment for prototypical debris bed has not been performed. Additionally, the prototypical debris bed consists of non-spherical particles and it is difficult to quantify model parameters. This situation brings sensitivity analysis to investigate effect of model parameters on the self-leveling phenomena of prototypical debris bed in present paper. As initial condition for sensitivity analysis, simple mound-like debris bed in sodium-filled lower plenum in reactor vessel is considered. The bed consists of the mixture of fuel debris of 3,300 kg and steel debris of 1,570 kg. Decay heat is given to this fuel debris. The model parameter is chosen as sensitivity parameter. Sensitivity analysis shows that the model parameters can effect on intensity of self-leveling phenomena and eventual flatness of bed. In all analyses, however, coolant and sodium vapor break the debris bed at mainly center part of bed and the debris is relocated to outside of bed. Through this process, the initial debris bed is almost planarized before re-melting of debris. This result shows that the model parameters affect the self-leveling phenomena, but its effect in the safety analysis of SFRs is limited. (author)

  3. Development of prototype reactor maintenance. (2) Application to piping support of sodium-cooled reactor prototype

    International Nuclear Information System (INIS)

    Arai, Masanobu; Kunogi, Kosuke; Aizawa, Kosuke; Chikazawa, Yoshitaka; Takaya, Shigeru; Kubo, Shigenobu; Kotake, Shoji; Ito, Takaya; Yamaguchi, Akira

    2017-01-01

    A maintenance program on piping support of prototype fast breeder reactor Monju are studied. Based on degradation mechanism, snubbers in Monju primary cooling system showed lifetime more than the plant lifetime of 30 years by experiments conservatively. For the first step during construction, visual inspection on accessible all supports could be available. In that visual inspection, mounting conditions and damages of all accessible supports could be monitored. One of major features of the Monju primary piping system is large thermal expansion due to large temperature difference between maintenance and operation conditions. Thanks to that large thermal expansion, integrity of piping supports could be monitored by measuring piping displacement. When technologies of piping displacement monitoring are matured in Monju, visual inspection on piping support could be shifted to piping displacement monitoring. At that stage, the visual inspection could be limited only on representative supports. (author)

  4. Status of the design and safety project for the sodium-cooled fast reactor as a generation IV nuclear energy system

    International Nuclear Information System (INIS)

    Niwa, Hajime; Fiorini, Gian-Luigi; Sim, Yoon-Sub; Lennox, Tom; Cahalan, James E.

    2005-01-01

    The Design and Safety Project Management Board (DSPMB) was established under the Sodium Cooled Fast Reactor (SFR) System Steering Committee (SSC) in the Generation IV international Forum. The DSPMB will promote collaborative R and D activities on reactor core design, and safety assessment for candidate systems, and also integrate these results together with those from other PMBs such as advanced fuel and component to a whole fast reactor system in order to develop high performance systems that will satisfy the goals of Generation IV nuclear energy systems. The DSPMB has formulated the present R and D schedules for this purpose. Two SFR concepts were proposed: a loop-type system with primarily a MOX fuel core and a pool-type system with a metal fuel core. Study of innovative systems and their evaluation will also be included. The safety project will cover both the safety assessment of the design and the preparation of the methods/tools to be used for the assessment. After a rather short viability phase, the project will move to the performance phase for development of performance data and design optimization of conceptual designs. This paper describes the schedules, work packages and tasks for the collaborative studies of the member countries. (author)

  5. Advanced materials for sodium-beta alumina batteries: Status, challenges and perspectives

    Science.gov (United States)

    Lu, Xiaochuan; Xia, Guanguang; Lemmon, John P.; Yang, Zhenguo

    The increasing penetration of renewable energy and the trend toward clean, efficient transportation have spurred growing interests in sodium-beta alumina batteries that store electrical energy via sodium ion transport across a β″-Al 2O 3 solid electrolyte at elevated temperatures (typically 300-350 °C). Currently, the negative electrode or anode is metallic sodium in molten state during battery operation; the positive electrode or cathode can be molten sulfur (Na-S battery) or solid transition metal halides plus a liquid phase secondary electrolyte (e.g., ZEBRA battery). Since the groundbreaking works in the sodium-beta alumina batteries a few decades ago, encouraging progress has been achieved in improving battery performance, along with cost reduction. However, there remain issues that hinder broad applications and market penetration of the technologies. To better the Na-beta alumina technologies require further advancement in materials along with component and system design and engineering. This paper offers a comprehensive review on materials of electrodes and electrolytes for the Na-beta alumina batteries and discusses the challenges ahead for further technology improvement.

  6. Advanced materials for sodium-beta alumina batteries: Status, challenges and perspectives

    International Nuclear Information System (INIS)

    Lu, Xiaochuan; Xia, Guanguang; Lemmon, John P.; Yang, Zhenguo

    2010-01-01

    The increasing penetration of renewable energy and the trend toward clean, efficient transportation have spurred growing interests in sodium-beta alumina batteries that store electrical energy via sodium ion transport across a β''-Al 2 O 3 solid electrolyte at elevated temperatures (typically 300-350 C). Currently, the negative electrode or anode is metallic sodium in molten state during battery operation; the positive electrode or cathode can be molten sulfur (Na-S battery) or solid transition metal halides plus a liquid phase secondary electrolyte (e.g., ZEBRA battery). Since the groundbreaking works in the sodium-beta alumina batteries a few decades ago, encouraging progress has been achieved in improving battery performance, along with cost reduction. However, there remain issues that hinder broad applications and market penetration of the technologies. To better the Na-beta alumina technologies require further advancement in materials along with component and system design and engineering. This paper offers a comprehensive review on materials of electrodes and electrolytes for the Na-beta alumina batteries and discusses the challenges ahead for further technology improvement. (author)

  7. Passive vibro-acoustic detection of a sodium-water reaction in a steam generator of a sodium-cooled fast neutrons nuclear reactor by beam forming

    International Nuclear Information System (INIS)

    Moriot, Jeremy

    2013-01-01

    This thesis deals with a new method to detect a sodium-water reaction in a steam generator of a fast sodium-cooled nuclear reactor. More precisely, the objective is to detect a micro-leak of water (flow ≤ 1 g/s) in less than 10 seconds by measuring the external shell vibrations of the component. The strong background noise in operation makes impossible the use of a detection system based on a threshold overrun. A beam forming method applied to vibrations measured by a linear array of accelerometers is developed in this thesis to increase the signal-to-noise ratio and to detect and locate the leak in the steam generator. A numerical study is first realized. Two models are developed in order to simulate the signals measured by the accelerometers of the array. The performances of the beam forming are then studied in function of several parameters, such as the source location and frequency, the damping factor, the background noise considered. The first model consists in an infinite plate in contact with a heavy fluid, excited by an acoustic monopole located in this fluid. Analyzing the transverse displacements in the wavenumber domain is useful to establish a criterion to sample correctly the vibration field of the plate. A second model, more representative of the system is also proposed. In this model, an elastic infinite cylindrical shell, filled with a heavy fluid is considered. The finite dimensions in the radial and circumferential directions lead to a modal behavior of the system which impacts the beam forming. Finally, the method is tested on an experimental mock-up which consists in a cylindrical pipe made in stainless steel and filled with water connected to hydraulic circuit. The water flow speed can be controlled by varying the speed of the pump. The acoustic source is generated by a hydro-phone. The performances of the beam forming are studied for different water flow speeds and different amplitude and frequencies of the source. (author) [fr

  8. IAEA'S study on advanced applications of water cooled nuclear power plants

    International Nuclear Information System (INIS)

    Cleveland, J.; McDonald, A.; Rao, A.; )

    2008-01-01

    About one-fifth of the world's energy consumption is used for electricity generation, with nuclear power contributing approximately 15.2% of this electricity. However; most of the world's energy consumption is for heat and transportation. Nuclear energy has considerable potential to penetrate these energy sectors now served by fossil fuels that are characterized by price volatility and finite supply. Advanced applications of nuclear energy include seawater desalination, district heating, and heat for industrial processes. Nuclear energy also has potential to provide a near-term, greenhouse gas free, source of energy for transportation. These applications rely on a source of heat and electricity. Nuclear energy from water-cooled reactors, of course, is not unique in this sense. Indeed, higher temperature heat can be produced by burning natural gas and coal, or through the use of other nuclear technologies such as gas-cooled or liquid-metal-cooled reactors. Water-cooled reactors, however; are being deployed today while other reactor types have had considerably less operational and regulatory experience and will take still some time to be widely accepted in the market. Both seawater desalination and district heating with nuclear energy are well proven, and new seawater desalination projects using water-cooled reactors will soon be commissioned. Provision of process heat with nuclear energy can result in less dependence on fossil fuels and contribute to reductions of greenhouse gases. Importantly, because nuclear power produces base-load electricity at stable and predictable prices, it provides a greenhouse gas free source of electricity for transportation systems (trains and subways), and for electric and plug-in hybrid vehicles, and in the longer term nuclear energy could produce hydrogen for fuel cell vehicles, as well as for other components of a hydrogen economy. These advanced applications can play an important role in enhancing public acceptance of nuclear

  9. An assessment of methods of calculating Doppler effects in plutonium fuelled sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Butland, A.T.D.; Reddell, G.

    1979-01-01

    After a survey of the requirements, an assessment of UK methods and data is made on the basis of the following work. First, the analysis of the SEFOR Doppler experiments, carried out using the UK FGL5 fine group nuclear data library, the MURAL cell code and whole reactor diffusion theory calculations of the neutron flux. Second, the analysis of some Japanese FCA central sample perturbation measurements of structural material Doppler effects. Third, an assessment of the accuracy of Doppler predictions in a sodium voided core using results from Zebra 5 and BIZET, and theoretical studies of additional effects relevant to power reactors and not covered by the above analyses, including the following, the calculation of Doppler effects at high temperature, fuel cycle and burn-up effects, and the heterogeneity effects of large fuelled subassemblies in pin geometry. The importance of crystalline binding effects in the fuel are discussed as is the importance of reactor material boundaries in the calculation of resonance shielding effects. Some suggestions for further Doppler studies are made. (U.K.)

  10. Large Eddy Simulation of turbulent flow in wire wrapped fuel pin bundles cooled by sodium

    International Nuclear Information System (INIS)

    Saxena, Aakanksha; Cadiou, Thierry; Bieder, Ulrich; Viazzo, Stephane

    2013-06-01

    The objective of the study is to understand the thermal hydraulics in a core sub-assembly with liquid sodium as coolant by performing detailed numerical simulations. The passage for the coolant flow between the fuel rods is maintained by thin wires wrapped around the rods. The contact point between the fuel pin and the spacer wire is the region of creation of hot spots and a cyclic variation of temperature in hot spots can adversely affect the mechanical properties of the clad due to the phenomena like thermal stripping. The current status quo provides two different models to perform the numerical simulations, namely Reynolds Averaged Navier-Stokes (RANS) and Large Eddy Simulation (LES). The two models differ in the extent of modelling used to close the Navier-Stokes equations. LES is a filtered approach where the large scale of motions are explicitly resolved while the small scale motions are modelled whereas RANS is a time averaging approach where all scale of motions are modelled. Thus LES involves less modelling as compared to RANS and so the results are comparatively more accurate. An attempt has been made to use the LES model. The simulations have been performed using the code Trio-U (developed by CEA). The turbulent statistics of the flow and thermal quantities are calculated. Finally the goal is to obtain the frequency of temperature oscillations at the region of hot spots near the spacer wire. (authors)

  11. FAST and SAFE Passive Safety Devices for Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chihyung; Kim, In-Hyung; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The major factor is the impact of the neutron spectral hardening. The second factor that affects the CVR is reduced capture by the coolant when the coolant voiding occurs. To improve the CVR, many ideas and concepts have been proposed, which include introduction of an internal blanket, spectrum softening, or increasing the neutron leakage. These ideas may reduce the CVR, but they deteriorate the neutron economy. Another potential solution is to adopt a passive safety injection device such as the ARC (autonomous reactivity control) system, which is still under development. In this paper, two new concepts of passive safety devices are proposed. The devices are called FAST (Floating Absorber for Safety at Transient) and SAFE (Static Absorber Feedback Equipment). Their purpose is to enhance the negative reactivity feedback originating from the coolant in fast reactors. SAFE is derived to balance the positive reactivity feedback due to sodium coolant temperature increases. It has been demonstrated that SAFE allows a low-leakage SFR to achieve a self-shutdown and self-controllability even though the generic coolant temperature coefficient is quite positive and the coolant void reactivity can be largely managed by the new FAST device. It is concluded that both FAST and SAFE devices will improve substantially the fast reactor safety and they deserve more detailed investigations.

  12. Implementation Plan for Qualification of Sodium-Cooled Fast Reactor Technology Information

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This document identifies and discusses implementation elements that can be used to facilitate consistent and systematic evaluation processes relating to quality attributes of technical information (with focus on SFR technology) that will be used to support licensing of advanced reactor designs. Information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The approach for determining acceptability of test data, analysis, and/or other technical information is based on guidance provided in INL/EXT-15-35805, “Guidance on Evaluating Historic Technology Information for Use in Advanced Reactor Licensing.” The implementation plan can be adopted into a working procedure at each of the national laboratories performing data qualification, or by applicants seeking future license application for advanced reactor technology.

  13. Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    David, Dijo K.; Mangarjuna Rao, P., E-mail: pmr@igcar.gov.in; Nashine, B.K.; Selvaraj, P.; Chellapandi, P.

    2015-09-15

    Highlights: • PAHR in SFR under large core relocation to in-vessel core catcher is numerically analyzed. • A 1-D thermal conduction model and a 2-D axisymmetric CFD model are developed for turbulent natural convection phenomenon. • The side pool (cold pool) was found out to be instrumental in storing heat and dissipating it to the heat sink. • Single tray type in-vessel core catcher is found to be thermally effective under one-fourth core relocation. - Abstract: A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal

  14. CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Hung, T.C.; Dhir, V.K.; Chang, J.C.; Wang, S.K.

    2011-01-01

    Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 o C which is substantially lower than ∼627 o C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a

  15. Start-up fuel and power flattening of sodium-cooled candle core

    International Nuclear Information System (INIS)

    Takaki, Naoyuki; Sagawa, Yu; Umino, Akitake; Sekimoto, Hiroshi

    2013-01-01

    The hard neutron spectrum and unique power shape of CANDLE enable its distinctive performances such as achieving high burnup more than 30% and exempting necessity of both enrichment and reprocessing. On the other hand, they also cause several challenging problems. One is how the initial fuel can be prepared to start up the first CANDLE reactor because the equilibrium fuel composition that enables stable CANDLE burning is complex both in axial and radial directions. Another prominent problem is high radial power peaking factor that worsens averaged burnup, namely resource utilization factor in once-through mode and shorten the life time of structure materials. The purposes of this study are to solve these two problems. Several ideas for core configurations and startup fuel using single enrichment uranium and iron as a substitute of fission products are studied. As a result, it is found that low enriched uranium is applicable to ignite the core but all concepts examined here exceeded heat limits. Adjustment in enrichment and height of active and burnt zone is opened for future work. Sodium duct assemblies and thorium fuel assemblies loaded in the center region are studied as measures to reduce radial power peaking factor. Replacing 37 fuels by thorium fuel assemblies in the zeroth to third row provides well-balanced performance with flattened radial power distribution. The CANDLE core loaded with natural uranium in the outer and thorium in the center region achieved 35.6% of averaged burnup and 7.0 years of cladding life time owing to mitigated local fast neutron irradiation at the center. Using thorium with natural or depleted uranium in CANDLE reactor is also beneficial to diversifying fission resource and extending available term of fission energy without expansion of needs for enrichment and reprocessing

  16. Status of sodium cooled fast reactors with closed fuel cycle in India

    International Nuclear Information System (INIS)

    Raj, B.

    2007-01-01

    Fast reactors form the second stage of India's 3-stage nuclear power programme. The seed for India's fast reactor programme was sown through the construction of the Fast Breeder Test Reactor (FBTR) at IGCAR, Kalpakkam, that was commissioned in 1985. FBTR has operated with an unique, indigenously developed plutonium rich mixed carbide fuel, which has reached a burn up as high as 155 GWd/t without any fuel failure in the core. The sodium systems in the reactor have performed excellently. The availability of the reactor has been as high as 92% in the recent campaigns. The fuel discharged from FBTR up to 100 GWd/t has been reprocessed successfully. The experience gained in the construction, commissioning and operation of FBTR has provided the necessary confidence to launch a Prototype FBR of 500 MWe capacity (PFBR). This reactor will be fuelled by uranium, plutonium mixed oxide. The reactor construction started in 2003 and the reactor is scheduled to be commissioned by 2010. The design of the reactor has incorporated the worldwide operating experience from the FBRs and has addressed various safety issues reported in literature, besides introducing a number of innovative features which have reduced the unit energy cost and contributed to its enhanced safety. Simultaneous with the construction of the reactor, the fuel cycle of the reactor has been addressed in a comprehensive manner and construction of a fuel cycle facility has been initiated. Subsequent to the PFBR, 4 more reactors with identical design are proposed to be constructed. Various elements of reactor design are being carefully analysed with the aim of introducing innovative features towards further reduction in unit energy cost and enhancing safety in these reactors

  17. Benchmark for Neutronic Analysis of Sodium-cooled Fast Reactor Cores with Various Fuel Types and Core Sizes

    International Nuclear Information System (INIS)

    Stauff, N.E.; Kim, T.K.; Taiwo, T.A.; Buiron, L.; Rimpault, G.; Brun, E.; Lee, Y.K.; Pataki, I.; Kereszturi, A.; Tota, A.; Parisi, C.; Fridman, E.; Guilliard, N.; Kugo, T.; Sugino, K.; Uematsu, M.M.; Ponomarev, A.; Messaoudi, N.; Lin Tan, R.; Kozlowski, T.; Bernnat, W.; Blanchet, D.; Brun, E.; Buiron, L.; Fridman, E.; Guilliard, N.; Kereszturi, A.; Kim, T.K.; Kozlowski, T.; Kugo, T.; Lee, Y.K.; Lin Tan, R.; Messaoudi, N.; Parisi, C.; Pataki, I.; Ponomarev, A.; Rimpault, G.; Stauff, N.E.; Sugino, K.; Taiwo, T.A.; Tota, A.; Uematsu, M.M.; Monti, S.; Yamaji, A.; Nakahara, Y.; Gulliford, J.

    2016-01-01

    One of the foremost Generation IV International Forum (GIF) objectives is to design nuclear reactor cores that can passively avoid damage of the reactor when control rods fail to scram in response to postulated accident initiators (e.g. inadvertent reactivity insertion or loss of coolant flow). The analysis of such unprotected transients depends primarily on the physical properties of the fuel and the reactivity feedback coefficients of the core. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS), the Sodium Fast Reactor core Feed-back and Transient response (SFR-FT) Task Force was proposed to evaluate core performance characteristics of several Generation IV Sodium-cooled Fast Reactor (SFR) concepts. A set of four numerical benchmark cases was initially developed with different core sizes and fuel types in order to perform neutronic characterisation, evaluation of the feedback coefficients and transient calculations. Two 'large' SFR core designs were proposed by CEA: those generate 3 600 MW(th) and employ oxide and carbide fuel technologies. Two 'medium' SFR core designs proposed by ANL complete the set. These medium SFR cores generate 1 000 MW(th) and employ oxide and metallic fuel technologies. The present report summarises the results obtained by the WPRS for the neutronic characterisation benchmark exercise proposed. The benchmark definition is detailed in Chapter 2. Eleven institutions contributed to this benchmark: Argonne National Laboratory (ANL), Commissariat a l'energie atomique et aux energies alternatives (CEA of Cadarache), Commissariat a l'energie atomique et aux energies alternatives (CEA of Saclay), Centre for Energy Research (CER-EK), Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Helmholtz Zentrum Dresden Rossendorf (HZDR), Institute of Nuclear Technology and Energy Systems (IKE), Japan Atomic Energy Agency (JAEA), Karlsruhe Institute of Technology (KIT

  18. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  19. Study on integrated TRU multi-recycling in sodium cooled fast reactor CDFR

    International Nuclear Information System (INIS)

    Hu Yun; Xu Mi; Wang Kan

    2010-01-01

    In view of recently proposed closed fuel cycle strategy which would recycle the integrated transuranics (TRU) from PWR spent fuel in the fast reactors, the neutronics characteristics of TRU recycled in China Demonstration Fast Reactor (CDFR) are studied in this paper. The results show that loading integrated TRU to substitute pure Pu as driver fuel will mainly make the influence on sodium void worth and negligible effects on other parameters, and hence TRU recycling in CDFR is feasible from viewpoint of core neutronics. If TRU is multi-recycled, the variation of TRU composition depends on fuel types and the ratio of TRU and U when recycling. It is indicated that, when TRU is multi-recycled in CDFR with MOX fuel, the minor actinides (MA) fraction in TRU will firstly decrease to ∼7.24% (minimum) within 8 TRU recycle times and then slowly increase to ∼7.7% after 20 TRU recycle times; while when TRU is multi-recycled in CDFR with metal fuel (TRU-U-10Zr), the MA fraction in TRU will gradually approach to an equilibrium state with the MA fraction of ∼3.8%, demonstrating better MA transmutation effect in metal fuel core. No matter 7.7 or 3.8%, they are both lower than ∼10% in PWR spent fuel with burnup of 45 GWd/tU, which presents satisfying effect of MA amount controlling for TRU multi-recycling strategy. On the other hand, the corresponding recycling parameters such as TRU heat release and neutron emission rate are also much lower in metal fuel than those in MOX fuel. Moreover, TRU recycled in metal fuel will bring greater fissile Pu isotopes equilibrium fraction due to better breeding capability of metal fuel. Finally, it could be summarized that integrated TRU multi-recycling in fast reactor can make contributions to both breeding and transmutation, and such strategy is a prospective closed fuel cycle manner to achieve the object of effective control of cumulated MA amount and sustainable development of nuclear energy.

  20. Safety design features for current UK advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yellowlees, J. M.; Cobb, E. C. [Nuclear Power Co. (Risley) Ltd. (UK)

    1981-01-15

    The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed.

  1. The cryogenic cooling program in high-heat-load optics at the Advanced Photon Source

    International Nuclear Information System (INIS)

    Rogers, C.S.

    1993-07-01

    This paper describes some of the aspects of the cryogenic optics program at the Advanced Photon Source (APS). A liquid-nitrogen-cooled, high-vacuum, double crystal monochromator is being fabricated at Argonne National Laboratory (ANL). A pumping system capable of delivering a variable flow rate of up to 10 gallons per minute of pressurized liquid nitrogen and removing 5 kilowatts of x-ray power is also being constructed. This specialized pumping system and monochromator will be used to test the viability of cryogenically cooled, high-heat-load synchrotron optics. It has been determined that heat transfer enhancement will be required for optics used with APS insertion devices. An analysis of a porous-matrix-enhanced monochromator crystal is presented. For the particular case investigated, a heat transfer enhancement factor of 5 to 6 was calculated

  2. Safety design features for current UK advanced gas-cooled reactors

    International Nuclear Information System (INIS)

    Yellowlees, J.M.; Cobb, E.C.

    1981-01-01

    The nuclear power stations planned for Heysham II and Torness will each have twin 660 MW(e) Advanced Gas-cooled Reactors (AGR) based on the design of those which have been operating at Hinkley Point 'B' and Hunterston 'B' since 1976. This paper has described the way in which the shutdown and cooling systems for the Heysham II and Torness AGRs have been selected in order to meet current UK safety requirements. Fault tree analyses have been used to identify the credible fault sequences, the probabilities of which have been calculated. By this means the relative importance of the various protective systems has been established and redundancy and reliability requirements identified. This systematic approach has led to a balanced design giving protection over the complete spectrum of fault sequences. Current safety requirements for thermal reactors in the UK and particular requirements in the design of the Heysham II and Torness reactors are discussed

  3. Calculation of ex-core detector weighting functions for a sodium-cooled tru burner mockup using MCNP5

    International Nuclear Information System (INIS)

    Pham Nhu Viet Ha; Min Jae Lee; Sunghwan Yun; Sang Ji Kim

    2015-01-01

    Power regulation systems of fast reactors are based on the signals of excore detectors. The excore detector weighting functions, which establish correspondence between the core power distribution and detector signal, are very useful for detector response analyses, e.g., in rod drop experiments. This paper presents the calculation of the weighting functions for a TRU burner mockup of the Korean Prototype Generation-IV Sodium-cooled Fast Reactor (named BFS-76-1A) using the MCNP5 multi-group adjoint capability. For generation of the weighting functions, all fuel assemblies were considered and each of them was divided into ten horizontal layers. Then the weighting functions for individual fuel assembly horizontal layers, the assembly weighting functions, and the shape annealing functions at RCP (Reactor Critical Point) and at conditions under which a control rod group was fully inserted into the core while other control rods at RCP were determined and evaluated. The results indicate that the weighting functions can be considered relatively insensitive to the control rods position during the rod drop experiments and therefore those weighting values at RCP can be applied to the dynamic rod worth simulation for the BFS-76-1A. (author)

  4. Basic visualization experiments on eutectic reaction of boron carbide and stainless steel under sodium-cooled fast reactor conditions

    International Nuclear Information System (INIS)

    Yamano, Hidemasa; Suzuki, Tohru; Kamiyama, Kenji; Kudo, Isamu

    2016-01-01

    This paper describes basic visualization experiments on eutectic reaction and relocation of boron carbide (B 4 C) and stainless steel (SS) under a high temperature condition exceeding 1500degC as well as the importance of such behaviors in molten core during a core disruptive accident in a Generation-IV sodium-cooled fast reactor (750 MWe class) designed in Japan. At first, a reactivity history was calculated using an exact perturbation calculation tool taking into account expected behaviors. This calculation indicated the importance of a relocation behavior of the B 4 C-SS eutectic because its behavior has a large uncertainty in the reactivity history. To clarify this behavior, basic experiments were carried out by visualizing the reaction of a B 4 C pellet contacted with molten SS in a high temperature-heating furnace. The experiments have shown the eutectic reaction visualization as well as freezing and relocation of the B 4 C-SS eutectic in upper part of the solidified test piece due to the density separation. (author)

  5. Numerical Analysis on the Free Fall Motion of the Control Rod Assembly for the Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Se-Hong; Choi, Choengryul; Son, Sung-Man [ELSOLTEC, Yongin (Korea, Republic of); Kim, Jae-Yong; Yoon, Kyung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    On receiving the scram signal, the control rod assemblies are released to fall into the reactor core by its weight. Thus drop time and falling velocity of the control rod assembly must be estimated for the safety evaluation. However, because of its complex shape, it is difficult to estimate the drop time by theoretical method. In this study, numerical analysis has been carried out in order to estimate drop time and falling velocity of the control rod assembly to provide the underlying data for the design optimization. Numerical analysis has been carried out to estimate the drop time and falling velocity of the control rod assembly for sodium-cooled fast reactor. Before performing the numerical analysis for the control rod assembly, sphere dropping experiment has been carried out for verification of the CFD methodology. The result of the numerical analysis for the method verification is almost same as the result of the experiment. Falling velocity and drag force increase rapidly in the beginning. And then it goes to the stable state. When the piston head of the control rod assembly is inserted into the damper, the drag force increases instantaneously and the falling velocity decreases quickly. The falling velocity is reduced about 14 % by damper. The total drop time of the control rod assembly is about 1.47s. In the next study, the experiment for the control rod assembly will be carried out, and its result is going to be compared with the CFD analysis result.

  6. Uranium Enrichment Reduction in the Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR with PBO Reflector

    Directory of Open Access Journals (Sweden)

    Chihyung Kim

    2016-04-01

    Full Text Available The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

  7. Uranium enrichment reduction in the Prototype Gen-IV sodium-cooled fast reactor (PGSFR) with PBO reflector

    Energy Technology Data Exchange (ETDEWEB)

    Hartanto, Donny; Kim, Chi Hyung; Kim, Yong Hee [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology (KAIST), Daejeon (Korea, Republic of)

    2016-04-15

    The Korean Prototype Gen-IV sodium-cooled fast reactor (PGSFR) is supposed to be loaded with a relatively-costly low-enriched U fuel, while its envisaged transuranic fuels are not available for transmutation. In this work, the U-enrichment reduction by improving the neutron economy is pursued to save the fuel cost. To improve the neutron economy of the core, a new reflector material, PbO, has been introduced to replace the conventional HT9 reflector in the current PGSFR core. Two types of PbO reflectors are considered: one is the conventional pin-type and the other one is an inverted configuration. The inverted PbO reflector design is intended to maximize the PbO volume fraction in the reflector assembly. In addition, the core radial configuration is also modified to maximize the performance of the PbO reflector. For the baseline PGSFR core with several reflector options, the U enrichment requirement has been analyzed and the fuel depletion analysis is performed to derive the equilibrium cycle parameters. The linear reactivity model is used to determine the equilibrium cycle performances of the core. Impacts of the new PbO reflectors are characterized in terms of the cycle length, neutron leakage, radial power distribution, and operational fuel cost.

  8. Study on MAs transmutation of accelerator-driven system sodium-cooled fast reactor loaded with metallic fuel

    International Nuclear Information System (INIS)

    Han Song; Yang Yongwei

    2007-01-01

    Through the analysis of the effect of heavy metal actinides on the effective multiplication constant (k eff ) of the core in accelerator-driven system (ADS) sodium-cooled fast reactor loaded with metallic fuel, we gave the method for determining fuel components. the characteristics of minor actinides (MAs) transmutation was analyzed in detail. 3D burn-up code COUPLE, which couples MCNP4c3 and ORIGEN2, was applied to the neutron simulation and burn up calculation. The results of optimized scheme shows that adjusting the proportion of 239 Pu and maintaining the value during the burn-up cycle is an efficient method of designing k eff and keeping stable during the burn-up cycle. Spallation neutrons lead to the neutron spectrum harder at inner core than that at outer core. It is in favor of improving MA's fission cross sections and the capture-to-fission ratio. The total MAs transmutation support ratio 8.3 achieves excellent transmutation effect. For higher flux at inner core leads to obvious differences on transmutation efficiency,only disposing MAs at inner core is in favor of decreasing the loading mass and improving MAs transmutation effect. (authors)

  9. Drop performance test of conceptually designed control rod assembly for prototype generation IV sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Kyu; Lee, Jae Han; Kim, Hoe Woong; KIm, Sung Kyun; Kim, Jong Bum [Sodium-cooled Fast Reactor NSSS Design Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-06-15

    The control rod assembly controls reactor power by adjusting its position during normal operation and shuts down chain reactions by its free drop under scram conditions. Therefore, the drop performance of the control rod assembly is important for the safety of a nuclear reactor. In this study, the drop performance of the conceptually designed control rod assembly for the prototype generation IV sodium-cooled fast reactor that is being developed at the Korea Atomic Energy Research Institute as a next-generation nuclear reactor was experimentally investigated. For the performance test, the test facility and test procedure were established first, and several free drop performance tests of the control rod assembly under different flow rate conditions were then carried out. Moreover, performance tests under several types and magnitudes of seismic loading conditions were also conducted to investigate the effects of seismic loading on the drop performance of the control rod assembly. The drop time of the conceptually designed control rod assembly for 0% of the tentatively designed flow rate was measured to be 1.527 seconds, and this agrees well with the analytically calculated drop time. It was also observed that the effect of seismic loading on the drop time was not significant.

  10. Structural assessment of intermediate printed circuit heat exchanger for sodium-cooled fast reactor with supercritical CO2 cycle

    International Nuclear Information System (INIS)

    Lee, Youho; Lee, Jeong Ik

    2014-01-01

    Highlights: • We numerically model PCHE stress arising from pressure, and thermal loadings. • Stress levels are the highest around S-CO 2 channels, due to high pressure of S-CO 2 . • The conventional analytic models for PCHE underestimate actual stress levels. • Plasticity sufficiently lowers stress levels at channel tips. • PCHE for SFR-SCO 2 is anticipated to assure compliance with ASME design standards. - Abstract: Structural integrity of intermediate Printed Circuit Heat Exchanger (PCHE) for Sodium-cooled Fast Reactor (SFR) attached to Supercritical CO 2 (S-CO 2 ) is investigated. ANSYS-Mechanical was used to simulate stress fields of representative PCHE channels, with temperature fields imported from FLUENT simulation. Mechanical stress induced by pressure loading is found to be the primary source of stress. As plasticity sufficiently lowers local stress concentration at PCHE channel tips, PCHE type intermediate heat exchangers made of SS316 are anticipated to reliably assure compliance with design standards prescribed in the ASME standards, thanks to the structure temperature that is below the effective creep inducing point. The actual life time of PCHE for SFR-SCO 2 is likely to be affected by mechanical behavior change of SS316 with reactions with S-CO 2 and fatigue

  11. A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Roh, Gyu Hong; Choi, Hang Bok

    2006-06-15

    Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted.

  12. Comparison of neutron diffusion theory codes in two and three space dimensions using a sodium cooled fast reactor benchmark

    International Nuclear Information System (INIS)

    Butland, A.T.D.; Putney, J.; Sweet, D.W.

    1980-04-01

    This report describes work performed to compare two UK neutron diffusion theory codes, TIGAR and SNAP, with published results for eight other codes available abroad. Both mesh edge and mesh centred finite difference diffusion theory codes as well as one axial synthesis code are included in the comparison and a range of iteration procedures are used by them. Comparison is made of calculations for a model of the sodium cooled fast reactor SNR-300 in both triangular and rectangular geometry and for a range of spatial meshes, enabling extrapolations to infinite mesh to be made. Calculated values of the effective multiplication constant, keff, for all the codes, agree very well when extrapolated to infinite mesh, indicating that no significant errors arising from the finite difference approximation but independent of mesh spacing are present in the calculations. The variation of keff with mesh area is found to be linear for the small meshes considered here, with the gradients for the mesh centred and mesh edged codes being of opposite sign. The results obtained using the mesh centred codes TIGAR, SNAP and CITATION agree closely with one another for all the meshes considered; the mesh edge codes agree less closely. (author)

  13. A Neutronic Feasibility Study on the Recycling of an Oxide Fuel in Sodium-Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Roh, Gyu Hong; Choi, Hang Bok

    2006-06-01

    Neutronic feasibility was implemented for the recycling of a mixed oxide fuel in sodium-cooled fast reactors (SFR) through a thermal/mechanical dry process, which is recognized as one of the most proliferation- resistant recycling processes. In order to assess the applicability of a simple dry process which is not capable of completely removing all the fission products from a spent fuel, sensitivity calculations were performed for the reactor physics parameters with a dependency on the fission product removal rate of the recycled spent fuel. The equilibrium core calculations were performed by the REBUS-3 code for a BN-600 core without blanket fuels and a modified core with an increased fuel volume fraction. The reactor performance parameters such as the transuranic content, breeding ratio, peak linear power, burnup reactivity swing and reactivity coefficients were calculated for an equilibrium core under a fixed fuel management scheme. The results showed that a recycling of the oxide fuel in the SFR is feasible if the fission products are removed by more than 70% through the dry process as far as the material balance is concerned. However the physics analysis also showed that some of the physics design parameters are slightly deteriorated. The results of this study indicate that the recycling characteristics can be improved if the dry process can remove more fission products, and the reactor configuration is further optimized or the spent fuel composition is adjusted

  14. Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 61-Rod Bundle for a Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Hyungmo; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Jeong, Ji-Young; Lee, Hyeong-Yeon

    2015-01-01

    For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are crucial factor for the design code verification and validation. Wrapped wires make a cross flow in a circumference of the fuel rod, and this effect lets flow be mixed. Therefore the sub-channel analysis method is commonly used for thermal hydraulic analysis of a SFR, a wire wrapped sub-channel type. To measure flow mixing characteristics, a wire mesh sensing technique can be useful method. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, the recent reports that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. The subchannel flow characteristics analysis method is commonly used for the thermal hydraulic analysis of a SFR, a wire wrapped subchannel type. In this study, mixing experiments were conducted successfully at a hexagonally arrayed 61-pin wire-wrapped fuel rod bundle test section. Wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable

  15. Acceleration Test Method for Failure Prediction of the End Cap Contact Region of Sodium Cooled Fast Reactor Fuel Rod

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung-Kyu; Lee, Young-Ho; Lee, Hyun-Seung; Lee, Kang-Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-05-15

    This paper reports the results of an acceleration test to predict the contact-induced failure that could occur at the cylinder-to-hole joint for the fuel rod of a sodium-cooled fast reactor (SFR). To incorporate the fuel life of the SFR currently under development at KAERI (around 35,000 h), the acceleration test method of reliability engineering was adopted in this work. A finite element method was used to evaluate the flow-induced vibration frequency and amplitude for the test parameter values. Five specimens were tested. The failure criterion during the life of the SFR fuel was applied. The S-N curve of the HT-9, the material of concern, was used to obtain the acceleration factor. As a result, a test time of 16.5 h was obtained for each specimen. It was concluded that the B{sub 0.004} life would be guaranteed for the SFR fuel rods with 99% confidence if no failure was observed at any of the contact surfaces of the five specimens.

  16. Study of the core compaction effects and its monitoring in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Zylbersztejn, F.

    2012-01-01

    Conclusions: • On calculation of reactivity impacts of core compaction/flowering: → Upper bound of the reactivity coefficients for each type of deformation; → Uniform compaction model: significant reactivity impact; Circular symmetric model: small reactivity impact. • On the visibility of these phenomena by the neutron detectors: → The direct monitoring of the core compaction by neutron detector in the BCC is not possible. (the identification that the reactivity perturbations observed are due to variation of the core geometry). Perspectives of solutions: → Improved core design: reducing the effects. → Physical improvements: Steel resistance to deformations (irradiation, flexion); Direct devices: core constraint (prevents deformations). → Additional calculations: Considering more localized deformations; Advanced monitoring with neutron noise (in progress)

  17. Instrumentation and control of future sodium cooled fast reactors - Design improvements

    International Nuclear Information System (INIS)

    Madhusoodanan, K.; Sakthivel, M.; Chellapandi, P.

    2013-06-01

    India's fast reactor program started with the 40 MWt Fast Breeder Test Reactor. 500 MWe Prototype Fast Breeder Reactor (PFBR) is currently under construction at Kalpakkam. Safety of PFBR is enhanced by improved design features of I and C system. Since the design of Instrumentation and control (I and C) of PFBR, considerable improvements in terms of advancement in technology and indigenization has taken place. Further improvements in I and C is proposed for solving many of the difficulties faced during the design and construction phases of PFBR. Design improvements proposed are covered in this paper which will make the implementation and maintenance of I and C of future SFRs easier. (authors)

  18. Experimental investigations of heat transfer during sodium boiling in fuel assembly model in justification of advanced fast reactor safety

    International Nuclear Information System (INIS)

    Khafizov, R.R.; Poplavskij, V.M.; Rachkov, V.I.; Sorokin, A.P.; Ashurko, Yu.M.; Volkov, A.V.; Ivanov, E.F.; Privezentsev, V.V.

    2015-01-01

    The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m 2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling [ru

  19. Development and Performance of an Advanced Ejector Cooling System for a Sustainable Built Environment

    Directory of Open Access Journals (Sweden)

    Paulo ePereira

    2015-06-01

    Full Text Available Ejector refrigeration is a promising technology for the integration into solar driven cooling systems because of its relative simplicity and low initial cost. The major drawback of such a system is associated to its relatively low coefficient of performance (COP under variable operating conditions. In order to overcome this problem, an advanced ejector was developed that changes its geometrical features depending on the upstream and downstream conditions. This paper provides a short overview of the development process and results of a small cooling capacity (1.5 kW solar driven cooling system using a variable geometry ejector. During the design steps, a number of theoretical works have been carried out, including the selection of the working fluid, the determination of the geometrical requirements and prototype design. Based on the analysis, R600a was selected as working fluid. A prototype was constructed with two independent variable geometrical factors: the area ratio and the nozzle exit position. A test rig was also assembled in order to test the ejector performance under controlled laboratory conditions and to elaborate a control algorithm for the variable geometry. Ejector performance was assessed by calculation of cooling cycle COP, entrainment ratio and critical back pressure. The results show that for a condenser pressure of 3 bar, an 80% increase in the COP was obtained when compared to the performance of a fixed geometry ejector. Experimental COP values varied between 0.4 and 0.8, depending on operating conditions. Currently the cooling cycle is being integrated into a solar driven demonstration site for long term in situ assessment.

  20. Status of advanced technology and design for water cooled reactors: Heavy water reactors

    International Nuclear Information System (INIS)

    1989-07-01

    In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

  1. Status of advanced technology and design for water cooled reactors: Light water reactors

    International Nuclear Information System (INIS)

    1988-10-01

    Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

  2. Neutronics investigation of advanced self-cooled liquid blanket systems in helical reactor

    International Nuclear Information System (INIS)

    Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M.Z.

    2006-10-01

    Neutronics performances of advanced self-cooled liquid blanket systems have been investigated in design activity of the helical-type reactor FFHR2. In the present study, a new three-dimensional (3-D) neutronics calculation system has been developed for the helical-type reactor to enhance quick feedback between neutronics evaluation and design modification. Using this new calculation system, advanced Flibe-cooled and Li-cooled liquid blanket systems proposed for FFHR2 have been evaluated to make clear design issues to enhance neutronics performance. Based on calculated results, modification of the blanket dimensions and configuration have been attempted to achieve the adequate tritium breeding ability and neutron shielding performance in the helical reactor. The total tritium breeding ratios (TBRs) obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. Issues in neutron shielding performance have been investigated quantitatively using 3-D geometry of the helical blanket system, support structures, poloidal coils etc. Shielding performance of the helical coils against direct neutrons from core plasma would achieve design target by further optimization of shielding materials. However, suppression of the neutron streaming and reflection through the divertor pumping areas in the original design is important issue to protect the poloidal coils and helical coils, respectively. Investigation of the neutron wall loading indicated that the peaking factor of the neutron wall load distribution would be moderated by the toroidal and helical effect of the plasma distribution in the helical reactor. (author)

  3. Economic Viability of Metallic Sodium-Cooled Fast Reactor Fuel in Korea

    Directory of Open Access Journals (Sweden)

    S. K. Kim

    2013-01-01

    Full Text Available This paper evaluates whether SFR metallic nuclear fuel can be economical. To make this determination, the cost of SFCF (SFR fuel cycle facilities was estimated, and the break-even point of the manufacturing cost of SFR metallic nuclear fuel for direct disposal option was then calculated. As a result of the cost estimation, the levelized unit cost (LUC for SFCF was calculated to be 5,311 $/kgHM, and the break-even point was calculated to be $5,267/kgHM. Therefore, the cost difference between LUC and the break-even point is not only small but is also within the relevant range of the uncertainty level of Class 3 in accordance with a generic cost estimate classification matrix of AACE (the Association for the Advancement of Cost Engineering. This means it is very difficult to judge the economical feasibility of SFR metallic nuclear fuel because as of today there are no commercial facilities in Korea or the world. The economic feasibility of SFR metallic nuclear fuel, however, will be enhanced if the mass production of SFCF becomes possible in the future.

  4. Development of a CVD silica coating for UK advanced gas-cooled nuclear reactor fuel pins

    International Nuclear Information System (INIS)

    Bennett, M.J.; Houlton, M.R.; Moore, D.A.; Foster, A.I.; Swidzinski, M.A.M.

    1983-04-01

    Vapour deposited silica coatings could extend the life of the 20% Cr/25% Ni niobium stabilised (20/25/Nb) stainless steel fuel cladding of the UK advanced gas cooled reactors. A CVD coating process developed originally to be undertaken at atmospheric pressure has now been adapted for operation at reduced pressure. Trials on the LP CVD process have been pursued to the production scale using commercial equipment. The effectiveness of the LP CVD silica coatings in providing protection to 20/25/Nb steel surfaces against oxidation and carbonaceous deposition has been evaluated. (author)

  5. Advances in measuring techniques for turbine cooling test rigs - Status report

    Science.gov (United States)

    Pollack, F. G.

    1974-01-01

    Instrumentation development pertaining to turbine cooling research has resulted in the design and testing of several new systems. Pressure measurements on rotating components are being made with a rotating system incorporating ten miniature transducers and a slip-ring assembly. The system has been tested successfully up to speeds of 9000 rpm. An advanced system development combining pressure transducer and thermocouple signals is also underway. Thermocouple measurements on rotating components are transferred off the shaft by a 72-channel rotating data system. Thermocouple data channels are electronically processed on board and then removed from the shaft in the form of a digital serial train by one winding of a rotary transformer.

  6. An assessment of the low seismic risk of the inherently safe sodium advanced fast reactor (SAFR)

    International Nuclear Information System (INIS)

    Rutherford, P.D.

    1988-01-01

    A recent probabilistic risk assessment (PRA) of the sodium advanced fast reactor (SAFR) demonstrated the inherently low risk of advanced liquid-metal, pool-type fast reactors with inherent safety systems. As a result, it was recognized that external events, especially seismic events, may not only be a major contributor to risk (as shown in several LWR PRAs) but also may completely dominate the risk. Accordingly, a seismic risk assessment has been completed for SAFR, which resulted in a core damage frequency of 2 x 10 -7 /year and a large release frequency of 4 x 10 -9 /year. This paper reports that public health risk in terms of early fatality risk and latent fatality risk were also several orders of magnitude below the NRC safety goals and below recent LWR risks reported in NUREB/CR1150

  7. Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (5) Identification of dominant factors in ex-vessel accident sequences

    International Nuclear Information System (INIS)

    Ohno, Shuji; Seino, Hiroshi; Miyahara, Shinya

    2009-01-01

    The evaluation of accident progression outside of a reactor vessel (ex-vessel) and subsequent transfer behavior of radioactive materials is of great importance from the viewpoint of Level 2 PSA. Hence typical ex-vessel accident sequences in the JAEA Sodium-cooled Fast Reactor are qualitatively discussed in this paper and dominant behaviors or factors in the sequences are investigated through parametric calculations using the CONTAIN/LMR code. Scenarios to be focused on are, 1) sodium vapor leakage from the reactor vessel and 2) sodium-concrete reaction, which are both to be considered in the accident category of LOHRS (loss of heat removal system) and might be followed by an early containment failure due to the thermal effect of sodium combustion and hydrogen burning respectively. The calculated results clarify that the sodium vapor leak rate and the scale of sodium-concrete reaction are the important factors to dominate the ex-vessel accident progression. In addition to the understandings of the dominant factors, the analyzed results also provide the specific information such as pressure loading value to the containment and the timing of pressurization, which is indispensable as technical base in Level 2 PSA for developing event trees and for quantifying the accident consequences. (author)

  8. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    International Nuclear Information System (INIS)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook

    2016-01-01

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results

  9. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  10. Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

    Directory of Open Access Journals (Sweden)

    Guidez Joel

    2017-01-01

    In the case of sodium-cooled fast reactors (SFRs, the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction. From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

  11. Effect of meat ingredients (sodium nitrite and erythorbate) and processing (vacuum storage and packaging atmosphere) on germination and outgrowth of Clostridium perfringens spores in ham during abusive cooling.

    Science.gov (United States)

    Redondo-Solano, Mauricio; Valenzuela-Martinez, Carol; Cassada, David A; Snow, Daniel D; Juneja, Vijay K; Burson, Dennis E; Thippareddi, Harshavardhan

    2013-09-01

    The effect of nitrite and erythorbate on Clostridium perfringens spore germination and outgrowth in ham during abusive cooling (15 h) was evaluated. Ham was formulated with ground pork, NaNO2 (0, 50, 100, 150 or 200 ppm) and sodium erythorbate (0 or 547 ppm). Ten grams of meat (stored at 5 °C for 3 or 24 h after preparation) were transferred to a vacuum bag and inoculated with a three-strain C. perfringens spore cocktail to obtain an inoculum of ca. 2.5 log spores/g. The bags were vacuum-sealed, and the meat was heat treated (75 °C, 20 min) and cooled within 15 h from 54.4 to 7.2 °C. Residual nitrite was determined before and after heat treatment using ion chromatography with colorimetric detection. Cooling of ham (control) stored for 3 and 24 h, resulted in C. perfringens population increases of 1.46 and 4.20 log CFU/g, respectively. For samples that contained low NaNO2 concentrations and were stored for 3 h, C. perfringens populations of 5.22 and 2.83 log CFU/g were observed with or without sodium erythorbate, respectively. Residual nitrite was stable (p > 0.05) for both storage times. Meat processing ingredients (sodium nitrite and sodium erythorbate) and their concentrations, and storage time subsequent to preparation of meat (oxygen content) affect C. perfringens spore germination and outgrowth during abusive cooling of ham. Copyright © 2013 Elsevier Ltd. All rights reserved.

  12. Pre evaluation for heat balance of prototype sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Han, Ji Woong; Kim, De Hee; Yoon, Jung; Kim, Eui Kwang; Lee, Tae Ho

    2012-01-01

    Under the long term advanced SFR R and D plan, the design of prototype reactor has been carried out toward the construction of the prototype SFR plant by 2028. The R and D efforts in fluid system design will be focused on developing a prototype design of primary heat transport system(PHTS), intermediate heat transport system (IHTS), decay heat removal system(DHRS), steam generation system(SGS), and related auxiliary system design for a prototype reactor as shown in Fig. 1. In order to make progress system design, top tier requirements for prototype reactor related to design parameters of NSSS and BOP should be decided at first. The top tier requirement includes general design basis, capacity and characteristics of reactor, various requirements related to safety, performance, securities, economics, site, and etc.. Extensive discussion has been done within Korea Atomic Energy Research Institute(KAERI) for the decision of top tier requirements of the prototype reactor. The core outlet temperature, which should be described as top tier requirements, is one of the critical parameter for system design. The higher core exit temperature could contribute to increase the plant efficiency. However, it could also contribute to decrease the design margin for structure and safety. Therefore various operating strategies based on different core outlet temperatures should be examined and evaluated. For the prototype reactor two core outlet temperatures are taken into accounted. The lower temperature is for the operation condition and the higher temperature is for the system design and licensing process of the prototype reactor. In order to evaluate the operability of prototype reactor designed based on higher temperature, the heat balance calculations have been performed at different core outlet temperature conditions. The electrical power of prototype reactor was assumed to be 100MWe and reference operating conditions were decided based on existing available data. The

  13. Advanced Small-Safe Long-Life Lead Cooled Reactor Cores for Future Nuclear Energy

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jin Hyeong; Hong, Ser Gi [Kyung Hee University, Seoul (Korea, Republic of)

    2014-10-15

    One of the reasons for use of the lead or lead-bismuth alloy coolants is the high boiling temperature that avoids the possibility of coolant voiding. Also, these coolants are compatible with air, steam, and water. Therefore, intermediate coolant loop is not required as in the sodium cooled reactors 3. Lead is considered to be more attractive coolant than lead-bismuth alloy because of its higher availability, lower price, and much lower amount of polonium activity by factor of 104 relatively to lead. On the other hand, lead has higher melting temperature of 601K than that of lead-bismuth (398K), which narrows the operating temperature range and also leads to the possibility of freezing and blockage in fresh cores. Neutronically, the lead and lead-bismuth have very similar characteristics to each other. The lead-alloy coolants have lower moderating power and higher scattering without increasing moderation for neutrons below 0.5MeV, which reduces the leakage of the neutrons through the core and provides an excellent reflecting capability for neutrons. Due to the above features of lead or lead-alloy coolants, there have been lots of studies on the small lead cooled core designs. In this paper, small-safe long-life lead cooled reactor cores having high discharge burnup are designed and neutronically analyzed.. The cores considered in this work rates 110MWt (36.7MWe). In this work, the long-life with high discharge burnup was achieved by using thorium or depleted uranium blanket loaded in the central region of the core. Also, we considered a reference core having no blanket for the comparison. This paper provides the detailed neutronic analyses for these small long-life cores and the detailed analyses of the reactivity coefficients and the composition changes in blankets. The results of the core design and analyses show that our small long-life cores can be operated without refueling over their long-lives longer than 45EFPYs (Effective Full Power Year). In this work

  14. Coolant-fuel interaction in Sodium-cooled Fast Reactors: Structural investigations of The Na-An-O (An = U, Np, Pu) systems

    International Nuclear Information System (INIS)

    Smith, A.L.; Raison, P.E.; Bykov, D.M.; Konings, R.J.; Caciuffo, R.; Cheetham, A.K.

    2014-01-01

    Nuclear energy has the potential to provide Europe with a secure and sustainable electricity supply at a competitive price and to make a significant contribution to the reduction of greenhouse gases emissions. The interest for Sodium-cooled-Fast-spectrum Reactors (SFRs), when compared to Pressurized Water Reactors (PWRs), lies in their more efficient management of plutonium and other actinides as well as their ability to use almost all of the energy in the natural uranium versus 1% utilized in thermal spectrum systems. The high fuel efficiency of fast reactors could greatly dampen concerns about fuel supply. But these reactors have also several drawbacks when compared to PWRs (i.e sodium fire, Na reaction with O2 and H2O, interaction of sodium with oxide fuels). Their development at an industrial scale needs therefore an exhaustive safety assessment that comprises both experimental work and development of sophisticated modelling tools able to describe the reactor behaviour in normal or incidental conditions

  15. Study on flow-induced vibration of large-diameter pipings in a sodium-cooled fast reactor. Influence of elbow curvature on velocity fluctuation field

    International Nuclear Information System (INIS)

    Ono, Ayako; Kimura, Nobuyuki; Kamide, Hideki; Tobita, Akira

    2010-02-01

    The main cooling system of Japan Sodium-cooled Fast Reactor (JSFR) consists of two loops to reduce the plant construction cost. In the design of JSFR, sodium coolant velocity is beyond 9m/s in the primary hot leg pipe with large-diameter (1.3m). The maximum Reynolds number in the piping reaches 4.2x10 7 . The hot leg pipe having a 90 degree elbow with curvature ratio of r/D=1.0, so-called 'short elbow', which enables a compact reactor vessel. In sodium cooled fast reactors, the system pressure is so low that thickness of pipings in the cooling system is thinner than that in LWRs. Under such a system condition in the cooling system, the flow-induced vibration (FIV) is concerned at the short elbow. The evaluation of the structural integrity of pipings in JSFR should be conducted based on a mechanistic approach of FIV at the elbow. It is significant to obtain the knowledge of the fluctuation intensity and spectra of velocity and pressure fluctuations in order to grasp the mechanism of the FIV. In this study, water experiments were conducted. Two types of 1/8 scaled elbows with different curvature ratio, r/D=1.0, 1.5, were used to investigate the influence of curvature on velocity fluctuation at the elbow. The velocity fields in the elbows were measured using a high speed PIV method. Unsteady behavior of secondary flow at the elbow outlet and separation flow at the inner wall of elbow were observed in the two types of elbows. It was found that the growth of secondary flow correlated with the flow fluctuation near the inside wall of the elbow. (author)

  16. Advanced Refrigerant-Based Cooling Technologies for Information and Communication Infrastructure (ARCTIC)

    Energy Technology Data Exchange (ETDEWEB)

    Salamon, Todd

    2012-12-13

    efficiency and carbon footprint reduction for our nation's Information and Communications Technology (ICT) infrastructure. The specific objectives of the ARCTIC project focused in the following three areas: i) advanced research innovations that dramatically enhance the ability to deal with ever-increasing device heat densities and footprint reduction by bringing the liquid cooling much closer to the actual heat sources; ii) manufacturing optimization of key components; and iii) ensuring rapid market acceptance by reducing cost, thoroughly understanding system-level performance, and developing viable commercialization strategies. The project involved participants with expertise in all aspects of commercialization, including research & development, manufacturing, sales & marketing and end users. The team was lead by Alcatel-Lucent, and included subcontractors Modine and USHose.

  17. Dietary sodium bicarbonate, cool temperatures, and feed withdrawal: impact on arterial and venous blood-gas values in broilers.

    Science.gov (United States)

    Wideman, R F; Hooge, D M; Cummings, K R

    2003-04-01

    Sodium bicarbonate (NaHCO3) has been used successfully in mammals and birds to alleviate pulmonary hypertension. Experiment 1 was designed to provide measurements of arterial and venous blood-gas values from unanesthetized male broilers subjected to a cool temperature (16 degrees C) challenge and fed either a control diet or the same diet alkalinized by dilution with 1% NaHCO3. The incidences of pulmonary hypertension syndrome (PHS, ascites) for broilers fed the control or bicarbonate diets were 15.5 and 10.5%, respectively (P = 0.36, NS). Non-ascitic broilers fed the control diet were heavier than those fed the bicarbonate diet on d 49 (2,671 vs. 2,484 g, respectively); however, other comparisons failed to reveal diet-related differences in heart weight, pulse oximetry values, electrocardiogram amplitudes, or blood-gas values (P > 0.05). When the data were resorted into categories based on right:total ventricular weight ratios (RV:TV) indicative of normal (RV:TV or = 0.28) pulmonary arterial pressures, broilers with elevated RV:TV ratios had poorly oxygenated arterial blood that was more acidic, had high partial pressure of CO2 (PCO2), and had higher HCO3 concentrations when compared with broilers with normal RV:TV ratios. Experiment 2 was conducted to determine if metabolic variations associated with differences in feed intake or environmental temperature potentially could mask an impact of diet composition on blood-gas values. Male broilers maintained at thermoneutral temperature (24 degrees C) either received feed ad libitum or had the feed withdrawn > or = 12 h prior to blood sampling. Broilers fed ad libitum had lower venous saturation of hemoglobin with O2, higher venous PCO2, and higher arterial HCO3 concentrations than broilers subjected to feed withdrawal. Broilers in experiment 2 fed ad libitum and exposed to cool temperatures (16 degrees C) had lower arterial partial pressure of O2 and higher venous PCO2 than broilers fed ad libitum and maintained at 24

  18. Measurement of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Sandalls, F.J.

    1978-03-01

    Sulphur is an important element in some food chains and the release of radioactive sulphur to the environment must be closely controlled if the chemical form is such that it is available or potentially available for entering food chains. The presence of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor warranted a study to assess the quantity and chemical form of the radioactive sulphur in order to estimate the magnitude of the potential environmental hazard which might arise from the release of coolant gas from Civil Advanced Gas-Cooled Reactors. A combination of gas chromatographic and radiochemical analyses revealed carbonyl sulphide to be the only sulphur-35 compound present in the coolant gas of the Windscale Reactor. The concentration of carbonyl sulphide was found to lie in the range 40 to 100 x 10 -9 parts by volume and the sulphur-35 specific activity was about 20 mCi per gramme. The analytical techniques are described in detail. The sulphur-35 appears to be derived from the sulphur and chlorine impurities in the graphite. A method for the preparation of carbonyl sulphide labelled with sulphur-35 is described. (author)

  19. Investigation of velocity distribution in an inner subchannel of wire wrapped fuel pin bundle of sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Nishimura, Masahiro; Kamide, Hideki; Ohshima, Hiroyuki; Kobayashi, Jun; Sato, Hiroyuki

    2011-01-01

    A sodium cooled fast reactor is designed to attain a high burn-up of core fuel in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity via change of flow area in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the detail of flow velocity distribution in a wire wrapped pin bundle. In this study, water experiments were carried out to investigate the detailed velocity distribution in a subchannel of nominal pin geometry as the first step. These basic data are not only useful for understanding of pin bundle thermal hydraulics but also a code validation. A wire-wrapped 3-pin bundle water model was applied to investigate the detailed velocity distribution in the subchannel which is surrounded by 3 pins with wrapping wire. The test section consists of an irregular hexagonal acrylic duct tube and three pins made of fluorinated resin pins which has nearly the same refractive index with that of water and a high light transmission rate. This enables to visualize the central subchannel through the pins. The velocity distribution in the central subchannel with the wrapping wire was measured by PIV (Particle Image Velocimetry) through a side wall of the duct tube. Typical flow velocity conditions in the pin bundle were 0.36m/s (Re=2,700) and 1.6m/s (Re=13,500). Influence of the wrapping wire on the velocity distributions in vertical and horizontal directions was confirmed. A clockwise swirl flow around the wire was found in subchannel. Significant differences were not recognized between the two cases of Re=2,700 and 13,500 concerning flow patterns. (author)

  20. Measurements of Flow Mixing at Subchannels in a Wire-Wrapped 37-Rod Bundle for a Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Hyungmo; Bae, Hwang; Chang, Seok-Kyu; Choi, Sun Rock; Lee, Dong Won; Ko, Yung Joo; Choi, Hae Seob; Euh, Dong-Jin; Lee, Hyeong-Yeon

    2014-01-01

    For a safety analysis in a core thermal design of a sodium-cooled fast reactor (SFR), flow mixing characteristics at subchannels in a wire-wrapped rod bundle are very important. Wrapped wires make a cross flow in a around the fuel rod) of the fuel rod, and this effect lets flow be mixed. Experimental results of flow mixing can be meaningful for verification and validation of thermal mixing correlation in a reactor core thermo-hydraulic design code. A wire mesh sensing technique can be useful method for measuring of flow mixing characteristics. A wire mesh sensor has been traditionally used to measure the void fraction of a two-phase flow field, i.e. gas and liquid. However, it has been recently reported that the wire mesh sensor can be used successfully to recognize the flow field in liquid phase by injecting a tracing liquid with a different level of electric conductivity. This can be powerfully adapted to recognize flow mixing characteristics by wrapped wires in SFR core thermal design. In this work, we conducted the flow mixing experiments using a custom designed wire mesh sensor. To verify and validate computer codes for the SFR core thermal design, mixing experiments were conducted at a hexagonally arrayed 37-pin wire-wrapped fuel rod bundle test section. The well-designed wire mesh sensor was used to measure flow mixing characteristics. The developed post-processing method has its own merits, and flow mixing results were reasonable. In addition, by uncertainty analysis, the system errors and the random error were estimated in experiments. Therefore, the present results and methods can be used for design code verification and validation

  1. Upper limits to americium concentration in large sized sodium-cooled fast reactors loaded with metallic fuel

    International Nuclear Information System (INIS)

    Zhang, Youpeng; Wallenius, Janne

    2014-01-01

    Highlights: • The americium transmutation capability of Integral Fast Reactor was investigated. • The impact from americium introduction was parameterized by applying SERPENT Monte Carlo calculations. • Higher americium content in metallic fuel leads to a power penalty, preserving consistent safety margins. - Abstract: Transient analysis of a large sized sodium-cooled reactor loaded with metallic fuel modified by different fractions of americium have been performed. Unprotected loss-of-offsite power, unprotected loss-of-flow and unprotected transient-over-power accidents were simulated with the SAS4A/SASSYS code based on the geometrical model of an IFR with power rating of 2500 MW th , using safety parameters obtained with the SERPENT Monte Carlo code. The Ti-modified austenitic D9 steel, having higher creep rupture strength, was considered as the cladding and structural material apart from the ferritic/martensitic HT9 steel. For the reference case of U–12Pu–1Am–10Zr fuel at EOEC, the margin to fuel melt during a design basis condition UTOP is about 50 K for a maximum linear rating of 30 kW/m. In order to maintain a margin of 50 K to fuel failure, the linear power rating has to be reduced by ∼3% and 6% for 2 wt.% and 3 wt.% Am introduction into the fuel respectively. Hence, an Am concentration of 2–3 wt.% in the fuel would lead to a power penalty of 3–6%, permitting a consumption rate of 3.0–5.1 kg Am/TW h th . This consumption rate is significantly higher than the one previously obtained for oxide fuelled SFRs

  2. Radio-contaminant behaviour in the cover-gas space and the containment building of a sodium-cooled fast reactor in accident conditions

    International Nuclear Information System (INIS)

    Mathe, Emmanuel

    2014-01-01

    In the context of the Generation IV initiative, the consequences of a severe-accident (SA) in a sodium-cooled fast reactor must be studied. A SFR (Sodium cooled Fast Reactor) severe accident involves the disruption of the core by super-criticality involving the destruction of a certain number of fuel assemblies. Subsequently the interaction between hot fuel and liquid sodium can lead to a vapor explosion which could create a breach in the primary system. Some contaminated liquid sodium would thus be ejected into the containment building. In this situation, the evaluation of potential releases to the environment (the source term) must forecast the quantity and the chemical speciation of the radio-contaminants likely to be released from the containment building. One critical risk of a SA is the production of contaminated aerosols in the containment building by spray ejection of primary-system sodium. Being pyrophoric, the sodium droplets react with oxygen first oxidizing then burning, with significant heat of combustion. As well as evaluating the consequences of a pressure rise inside the containment, the evolution of the sodium must be assessed since not only is it activated and contaminated but, in oxide form, very toxic. Ultimately, the aerosols are the main radiological risk acting as the vector for radionuclide transport to the environment in the event of a problem with the confinement. These aerosols could evolve and interact with the FP (Fissile Products) and these interactions could modify the physical and chemical nature of the PF. We model a large part of the events that occur during a SA inside a SFR from the sodium spray fire to the reaction between sodium aerosols and PF (iodine). At first, we develop a numerical model (NATRAC) that simulates the sodium spray fire, calculates the temperature and the pressure inside the containment as well as the mass of aerosols produced during this kind of fire. The simulation has been validated with different

  3. Development of advanced methods for signal processing in the monitoring of sodium-cooled reactors

    International Nuclear Information System (INIS)

    Schleisiek, K.; Aberle, J.; Massier, H.; Scherer, K.P.; Vaeth, W.; Leder, H.J.; Schade, H.J.

    1987-01-01

    Selected examples (acoustic boiling detection, pattern recognition method, identification of fuel element vibrations, diagnosis system for KNK II) are used to demonstrate the benefits of up-to-date information technology in the monitoring of nuclear facilities. The methods used range from intelligent frequency analysis to AI methods like pattern recognition and expert systems. (DG) [de

  4. The concept of the sodium cooled small fast reactor 4S and the analyses of the loss of flow events

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Ueda, Nobuyuki; Koga, Tomonari; Matsumiya, Hisato

    2007-01-01

    CRIEPI has been developing the 4S reactor (Super Safe, Small and Simple reactor) for application in dispersed energy supply and multipurpose use, in conjunction with Toshiba Corporation. The 4S is sodium cooled fast reactor and their electrical output has two options of 10MWe and 50MWe. In this paper, 10MWe 4S (4S-10M) was proposed. 4S-10M has some unique features. It employs a burn-up control system with annular reflector in place of the control rod that requires the frequent maintenance service. The core life time of the 4S-10M is 30 years and the fuel transport is not required during core life time. All temperature feedback coefficients are negative during core life time. In the latest design for 4S-10M, a pool and tall type reactor design was selected to reduce the construction cost. Two types of decay heat removal system (Reactor Vessel Auxiliary Cooling System; RVACS, Intermediate Reactor Auxiliary Cooling System; IRACS) using natural convection power were adopted. It is necessary to confirm that these two heat removal system can operate appropriately. The transition analyses were executed by the CERES code to evaluate the design feasibility and the thermal hydraulic characteristics of the 4S-10M. CERES is a multi-dimensional plant dynamics simulation code for liquid metal reactors developed by the CRIEPI. CERES can perform simulations ranging from forced circulation (full/partial power operation) to natural circulation. Components (pumps, IHXs, SGs, pipings, etc.) of the reactor are modeled as one-dimensional. Multi-dimensional plena are connected to such components. Two loss-of-flow accident sequences are considered. In the first case, it is assumed that the primary and the secondary pump were stopped by the total station black out. The reactor shut down system was assumed to be success. This sequence is referred to as the protected loss-of-flow accident (PLOF). In the second case, it is assumed that the reactor shut down systems fail to operate and the

  5. Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh, E-mail: mukeshd@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Chakravarty, Aranyak [School of Nuclear Studies and Application, Jadavpur University, Kolkata 700032 (India); Nayak, A.K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Prasad, Hari; Gopika, V. [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2014-10-15

    Highlights: • The paper deals with the reliability assessment of Passive Containment Cooling System of Advanced Heavy Water Reactor. • Assessment of Passive System ReliAbility (APSRA) methodology is used for reliability assessment. • Performance assessment of the PCCS is initially performed during a postulated design basis LOCA. • The parameters affecting the system performance are then identified and considered for further analysis. • The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. - Abstract: Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components.

  6. Reliability assessment of Passive Containment Cooling System of an Advanced Reactor using APSRA methodology

    International Nuclear Information System (INIS)

    Kumar, Mukesh; Chakravarty, Aranyak; Nayak, A.K.; Prasad, Hari; Gopika, V.

    2014-01-01

    Highlights: • The paper deals with the reliability assessment of Passive Containment Cooling System of Advanced Heavy Water Reactor. • Assessment of Passive System ReliAbility (APSRA) methodology is used for reliability assessment. • Performance assessment of the PCCS is initially performed during a postulated design basis LOCA. • The parameters affecting the system performance are then identified and considered for further analysis. • The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. - Abstract: Passive Systems are increasingly playing a prominent role in the advanced nuclear reactor systems and are being utilised in normal operations as well as safety systems of the reactors following an accident. The Passive Containment Cooling System (PCCS) is one of the several passive safety features in an Advanced Reactor (AHWR). In this paper, the APSRA methodology has been employed for reliability evaluation of the PCCS of AHWR. Performance assessment of the PCCS is initially performed during a postulated design basis LOCA using the best-estimate code RELAP5/Mod 3.2. The parameters affecting the system performance are then identified and considered for further analysis. Based on some pre-determined failure criterion, the failure surface for the system is predicted using the best-estimate code taking into account the deviations of the identified parameters from their nominal states as well as the model uncertainties inherent to the best estimate code. Root diagnosis is then carried out to determine the various failure causes, which occurs mainly due to malfunctioning of mechanical components. The failure probabilities of the various components are assessed through a classical PSA treatment using generic data. The reliability of the PCCS is then evaluated from the probability of availability of these components

  7. Hydrogen Sulfide Prevents Advanced Glycation End-Products Induced Activation of the Epithelial Sodium Channel

    Directory of Open Access Journals (Sweden)

    Qiushi Wang

    2015-01-01

    Full Text Available Advanced glycation end-products (AGEs are complex and heterogeneous compounds implicated in diabetes. Sodium reabsorption through the epithelial sodium channel (ENaC at the distal nephron plays an important role in diabetic hypertension. Here, we report that H2S antagonizes AGEs-induced ENaC activation in A6 cells. ENaC open probability (PO in A6 cells was significantly increased by exogenous AGEs and that this AGEs-induced ENaC activity was abolished by NaHS (a donor of H2S and TEMPOL. Incubating A6 cells with the catalase inhibitor 3-aminotriazole (3-AT mimicked the effects of AGEs on ENaC activity, but did not induce any additive effect. We found that the expression levels of catalase were significantly reduced by AGEs and both AGEs and 3-AT facilitated ROS uptake in A6 cells, which were significantly inhibited by NaHS. The specific PTEN and PI3K inhibitors, BPV(pic  and LY294002, influence ENaC activity in AGEs-pretreated A6 cells. Moreover, after removal of AGEs from AGEs-pretreated A6 cells for 72 hours, ENaC PO remained at a high level, suggesting that an AGEs-related “metabolic memory” may be involved in sodium homeostasis. Our data, for the first time, show that H2S prevents AGEs-induced ENaC activation by targeting the ROS/PI3K/PTEN pathway.

  8. Advanced surrogate model and sensitivity analysis methods for sodium fast reactor accident assessment

    International Nuclear Information System (INIS)

    Marrel, A.; Marie, N.; De Lozzo, M.

    2015-01-01

    Within the framework of the generation IV Sodium Fast Reactors, the safety in case of severe accidents is assessed. From this statement, CEA has developed a new physical tool to model the accident initiated by the Total Instantaneous Blockage (TIB) of a sub-assembly. This TIB simulator depends on many uncertain input parameters. This paper aims at proposing a global methodology combining several advanced statistical techniques in order to perform a global sensitivity analysis of this TIB simulator. The objective is to identify the most influential uncertain inputs for the various TIB outputs involved in the safety analysis. The proposed statistical methodology combining several advanced statistical techniques enables to take into account the constraints on the TIB simulator outputs (positivity constraints) and to deal simultaneously with various outputs. To do this, a space-filling design is used and the corresponding TIB model simulations are performed. Based on this learning sample, an efficient constrained Gaussian process metamodel is fitted on each TIB model outputs. Then, using the metamodels, classical sensitivity analyses are made for each TIB output. Multivariate global sensitivity analyses based on aggregated indices are also performed, providing additional valuable information. Main conclusions on the influence of each uncertain input are derived. - Highlights: • Physical-statistical tool for Sodium Fast Reactors TIB accident. • 27 uncertain parameters (core state, lack of physical knowledge) are highlighted. • Constrained Gaussian process efficiently predicts TIB outputs (safety criteria). • Multivariate sensitivity analyses reveal that three inputs are mainly influential. • The type of corium propagation (thermal or hydrodynamic) is the most influential

  9. Advances in liquid metal cooled ADS systems, and useful results for the design of IFMIF

    International Nuclear Information System (INIS)

    Massaut, V.; Debruyn, D.; Decreton, M.

    2007-01-01

    Full text of publication follows: Liquid metal cooled Accelerator Driven Systems (ADS) have a lot of design commonalities with the design of IFMIF. The use of a powerful accelerator and a liquid metal spallation source makes it similar to the main features of the IFMIF irradiator. Developments in the field of liquid metal ADS can thus be very useful for the design phase of IFMIF, and synergy between both domains should be enhanced to avoid dubbing work already done. The liquid metal ADS facilities are developed for testing materials under high fast (> 1 MeV) neutron flux, and also for studying the transmutation of actinides as foreseen in the P and T (Partitioning and Transmutation) strategy of future fission industry. The ADS are mostly constituted of a sub-critical fission fuel assembly matrix, a spallation source (situated at the centre of the fuel arrangement) and a powerful accelerator targeting the spallation source. In liquid metal ADS, the spallation source is a liquid metal (like Pb-Bi) which is actively cooled to remove the power generated by the particle beam, spallation reactions and neutrons. Based on an advanced ADS design (e.g. the MYRRHA/XT-ADS facility), the paper shows the various topics which are common for both facilities (ADS and IFMIF) and highlights their respective specificities, leading to focused R and D activities. This would certainly cover the common aspects related to high power accelerators, liquid metal targets and beam-target coupling. But problems of safety, radioprotection, source heating and cooling, neutrons shielding, etc... lead also to common features and developments. Results already obtained for the ADS development will illustrate this synergy. This paper will therefore allow to take profit of recent developments in both fission and fusion programs and enhance the collaboration among the R and D teams in both domains. (authors)

  10. Advances in liquid metal cooled ADS systems, and useful results for the design of IFMIF

    Energy Technology Data Exchange (ETDEWEB)

    Massaut, V.; Debruyn, D. [SCK CEN, Mol (Belgium); Decreton, M. [Ghent Univ., Dept. of Applied Physics (Belgium)

    2007-07-01

    Full text of publication follows: Liquid metal cooled Accelerator Driven Systems (ADS) have a lot of design commonalities with the design of IFMIF. The use of a powerful accelerator and a liquid metal spallation source makes it similar to the main features of the IFMIF irradiator. Developments in the field of liquid metal ADS can thus be very useful for the design phase of IFMIF, and synergy between both domains should be enhanced to avoid dubbing work already done. The liquid metal ADS facilities are developed for testing materials under high fast (> 1 MeV) neutron flux, and also for studying the transmutation of actinides as foreseen in the P and T (Partitioning and Transmutation) strategy of future fission industry. The ADS are mostly constituted of a sub-critical fission fuel assembly matrix, a spallation source (situated at the centre of the fuel arrangement) and a powerful accelerator targeting the spallation source. In liquid metal ADS, the spallation source is a liquid metal (like Pb-Bi) which is actively cooled to remove the power generated by the particle beam, spallation reactions and neutrons. Based on an advanced ADS design (e.g. the MYRRHA/XT-ADS facility), the paper shows the various topics which are common for both facilities (ADS and IFMIF) and highlights their respective specificities, leading to focused R and D activities. This would certainly cover the common aspects related to high power accelerators, liquid metal targets and beam-target coupling. But problems of safety, radioprotection, source heating and cooling, neutrons shielding, etc... lead also to common features and developments. Results already obtained for the ADS development will illustrate this synergy. This paper will therefore allow to take profit of recent developments in both fission and fusion programs and enhance the collaboration among the R and D teams in both domains. (authors)

  11. Robust remote-pumping sodium laser for advanced LIDAR and guide star applications

    Science.gov (United States)

    Ernstberger, Bernhard; Enderlein, Martin; Friedenauer, Axel; Schwerdt, Robin; Wei, Daoping; Karpov, Vladimir; Leisching, Patrick; Clements, Wallace R. L.; Kaenders, Wilhelm G.

    2015-10-01

    The performance of large ground-based optical telescopes is limited due to wavefront distortions induced by atmospheric turbulence. Adaptive optics systems using natural guide stars with sufficient brightness provide a practical way for correcting the wavefront errors by means of deformable mirrors. Unfortunately, the sky coverage of bright stars is poor and therefore the concept of laser guide stars was invented, creating an artificial star by exciting resonance fluorescence from the mesospheric sodium layer about 90 km above the earth's surface. Until now, mainly dye lasers or sumfrequency mixing of solid state lasers were used to generate laser guide stars. However, these kinds of lasers require a stationary laser clean room for operation and are extremely demanding in maintenance. Under a development contract with the European Southern Observatory (ESO) and W. M. Keck Observatory (WMKO), TOPTICA Photonics AG and its partner MPB Communications have finalized the development of a next-generation sodium guide star laser system which is available now as a commercial off-the-shelf product. The laser is based on a narrow-band diode laser, Raman fiber amplifier (RFA) technology and resonant second-harmonic generation (SHG), thus highly reliable and simple to operate and maintain. It emits > 22 W of narrow-linewidth (≈ 5 MHz) continuous-wave radiation at sodium resonance and includes a re-pumping scheme for boosting sodium return flux. Due to the SHG resonator acting as spatial mode filter and polarizer, the output is diffraction-limited with RMS wavefront error concept of line-replaceable units (LRU). A comprehensive system software, as well as an intuitive service GUI, allow for remote control and error tracking down to at least the LRU level. In case of a failure, any LRU can be easily replaced. With these fiber-based guide star lasers, TOPTICA for the first time offers a fully engineered, off-the-shelf guide star laser system for groundbased optical telescopes

  12. Microscopical examination of carbon deposits formed in the Windscale advanced gas cooled reactor

    International Nuclear Information System (INIS)

    Livesey, D.J.; Chatwin, W.H.; Pearce, J.H.

    1980-12-01

    Methods are described of sampling and examining carbon deposits on fuel cladding in the Windscale advanced gas-cooled reactor. Deposition is observed on fuel cladding in both the reactor core and experimental loops in carbon dioxide coolants containing various amounts of carbon monoxide and methane. Deposit distribution over the cladding surface indicated that nucleation is dependent on local surface conditions. Microscopical examination showed that deposit thickness increases by carbon filament growth into the coolant gas stream and that the process can be markedly influenced by metallic impurities. There is evidence that nickel can play a particularly significant role in deposition in loop experiments but similar effects have not been observed in the reactor core. (author)

  13. The rate of diffusion into advanced gas cooled reactor moderator bricks: an equivalent cylinder model

    International Nuclear Information System (INIS)

    Kyte, W.S.

    1980-01-01

    The graphite moderator bricks which make up the moderator of an advanced gas-cooled nuclear reactor (AGR) are of many different and complex shapes. Many physico-chemical processes that occur within these porous bricks include a diffusional step and thus to model these processes it is necessary to solve the diffusion equation (with chemical reaction) in a porous medium of complex shape. A finite element technique is applied to calculating the rate at which nitrogen diffuses into and out of the porous moderator graphite during operation of a shutdown procedure for an AGR. However, the finite element method suffers from several disadvantages that undermine its general usefulness for calculating rates of diffusion in AGR moderator cores. A model which overcomes some of these disadvantages is presented (the equivalent cylinder model) and it is shown that this gives good results for a variety of different boundary and initial conditions

  14. Multi-criteria methodology to design a sodium-cooled carbide-fueled Gen-IV reactor

    International Nuclear Information System (INIS)

    Stauff, N.

    2011-01-01

    Compared with earlier plant designs (Phenix, Super-Phenix, EFR), Gen IV Sodium-cooled Fast Reactor requires improved economics while meeting safety and non-proliferation criteria. Mixed Oxide (U-Pu)O 2 fuels are considered as the reference fuels due to their important and satisfactory feedback experience. However, innovative carbide (U-Pu)C fuels can be considered as serious competitors for a prospective SFR fleet since carbide-fueled SFRs can offer another type of optimization which might overtake on some aspects the oxide fuel technology. The goal of this thesis is to reveal the potentials of carbide by designing an optimum carbide-fueled SFR with competitive features and a naturally safe behavior during transients. For a French nuclear fleet, a 1500 MW(e) break-even core is considered. To do so, a multi-physic approach was developed taking into account neutronics, fuel thermo-mechanics and thermal-hydraulic at a pre-design stage. Simplified modeling with the calculation of global neutronic feedback coefficients and a quasi-static evaluation was developed to estimate the behavior of a core during overpower transients, loss of flow and/or loss of heat removal transients. The breakthrough of this approach is to provide the designer with an overall view of the iterative process, emphasizing the well-suited innovations and the most efficient directions that can improve the SFR design project.This methodology was used to design a core that benefits from the favorable features of carbide fuels. The core developed is a large carbide-fueled SFR with high power density, low fissile inventory, break-even capability and forgiving behaviors during the un-scrammed transients studied that should prevent using expensive mitigate systems. However, the core-peak burnup is unlikely to significantly exceed 100 MWd/kg because of the large swelling of the carbide fuel leading to quick pellet-clad mechanical interaction and the low creep capacity of carbide. Moderate linear power fuel

  15. Energy reduction for a dual circuit cooling water system using advanced regulatory control

    International Nuclear Information System (INIS)

    Muller, C.J.; Craig, I.K.

    2016-01-01

    Highlights: • Potentially reduce energy required by a dual circuit cooling water system by 30%. • Accomplished using an advanced regulatory control and switching strategy. • No formal process model is required. • Can be implemented on control system hardware commonly used in industry. - Abstract: Various process utilities are used in the petrochemical industry as auxiliary variables to facilitate the addition/removal of energy to/from the process, power process equipment and inhibit unwanted reaction. Optimisation activities usually focus on the process itself or on the utility consumption though the generation and distribution of these utilities are often overlooked in this regard. Many utilities are prepared or generated far from the process plant and have to be transported or transmitted, giving rise to more losses and potential inefficiencies. To illustrate the potential benefit of utility optimisation, this paper explores the control of a dual circuit cooling water system with focus on energy reduction subject process constraints. This is accomplished through the development of an advanced regulatory control (ARC) and switching strategy which does not require the development of a system model, only rudimentary knowledge of the behaviour of the process and system constraints. The novelty of this manuscript lies in the fact that it demonstrates that significant energy savings can be obtained by applying ARC to a process utility containing both discrete and continuous dynamics. Furthermore, the proposed ARC strategy does not require a plant model, uses only existing plant equipment, and can be implemented on control system hardware commonly used in industry. The simulation results indicate energy saving potential in the region of 30% on the system under investigation.

  16. Characterization of velocity and temperature fields in a 217 pin wire wrapped fuel bundle of sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Naveen Raj, M.; Velusamy, K.

    2016-01-01

    Highlights: • We simulate flow and temperature fields in fuel subassembly of fast reactor. • We perform high fidelity computations for 217 pin bundle of 7 axial pitch lengths. • We investigate transverse and axial flows in different types of subchannels. • Correlations are proposed for transverse flow, which form input for subchannel analysis. • Periodic variations of large magnitude are observed in subchannel flow rates. - Abstract: RANS based computational fluid dynamic (CFD) simulation of flow and temperature fields in a fast reactor fuel subassembly has been carried out. The sodium cooled prototype subassembly consists of 217 pins with helical wire spacers. An axial length of seven helical wire pitches has been considered for the study adopting a structured mesh having 36 million points and 84 processors in parallel. The computational model has been validated against in-house and published experimental data for friction factor and Nusselt number. Also, the transverse flow in the central subchannel and swirl flow in the peripheral subchannel are compared against reported experimental data and those computed by subchannel models. The focus of the study is investigation of transverse and axial flows in different types of subchannels. Based on the 3-dimensional CFD study, correlations have been proposed for calculation of transverse flow, which forms an important input for development of subchannel analysis codes. Periodic variations have been observed in the subchannel axial flow rates. For the subchannels located in the central region, the peak to peak variation in the axial flow rate is ∼21% and it is found to be contributed by the changes in the flow area and hydraulic resistance due to frequent passage of helical wires through the subchannel. For the subchannels located in the periphery, this variation is as high as 50%. The transverse flow in the central subchannels follows a cosine profile, for all the faces. However, there is a phase lag of 120

  17. Second meeting of the International Working Group on Advanced Technologies for Water Cooled Reactors, Helsinki, 6-9 June 1988

    International Nuclear Information System (INIS)

    1989-05-01

    The Second Meeting of the IAEA International Working Group on Advanced Technologies for Water Cooled Reactors (IWGATWR) was held in Helsinki, Finland, from 6-9 June 1988. The Summary Report (Part II) contains the papers which review the national programmes since the first meeting of IWGATWR in May 1987 in the field of Advanced Technologies for Water Cooled Reactors and other presentations at the Meeting. A separate abstract was prepared for each of these 12 papers presented at the meeting. Figs and tabs

  18. On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

    Directory of Open Access Journals (Sweden)

    Jong-Bum Kim

    2016-10-01

    Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  19. 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

  20. Recent advances on Fe- and Mn-based cathode materials for lithium and sodium ion batteries

    Science.gov (United States)

    Zhu, Xiaobo; Lin, Tongen; Manning, Eric; Zhang, Yuancheng; Yu, Mengmeng; Zuo, Bin; Wang, Lianzhou

    2018-06-01

    The ever-growing market of electrochemical energy storage impels the advances on cost-effective and environmentally friendly battery chemistries. Lithium-ion batteries (LIBs) are currently the most critical energy storage devices for a variety of applications, while sodium-ion batteries (SIBs) are expected to complement LIBs in large-scale applications. In respect to their constituent components, the cathode part is the most significant sector regarding weight fraction and cost. Therefore, the development of cathode materials based on Earth's abundant elements (Fe and Mn) largely determines the prospects of the batteries. Herein, we offer a comprehensive review of the up-to-date advances on Fe- and Mn-based cathode materials for LIBs and SIBs, highlighting some promising candidates, such as Li- and Mn-rich layered oxides, LiNi0.5Mn1.5O4, LiFe1-xMnxPO4, NaxFeyMn1-yO2, Na4MnFe2(PO4)(P2O7), and Prussian blue analogs. Also, challenges and prospects are discussed to direct the possible development of cost-effective and high-performance cathode materials for future rechargeable batteries.

  1. Development of failed fuel detection and location system in sodium-cooled large reactor. Sampling method of failed fuels under the slit

    International Nuclear Information System (INIS)

    Aizawa, Kousuke; Fujita, Kaoru; Kamide, Hideki; Kasahara, Naoto

    2010-01-01

    A conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR) is in progress as an issue of the 'Fast Reactor Cycle Technology Development (FaCT)' project in Japan. JSFR adopts a Selector-Valve mechanism for the failed fuel detection and location (FFDL) system. The Selector-Valve FFDL system identifies failed fuel subassemblies by sampling sodium from each fuel subassembly outlet and detecting fission product. One of the JSFR design features is employing an upper internal structure (UIS) with a radial slit, in which an arm of fuel handling machine can move and access the fuel assemblies under the UIS. Thus, JSFR cannot place sampling nozzles right above the fuel subassemblies located under the slit. In this study, the sampling method for indentifying under-slit failed fuel subassemblies has been demonstrated by water experiments. (author)

  2. Evaluation of Microstructural and Mechanical Property of Medium-sized HT9 Cladding Forged Material for Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Jun Hwan; Lee, Kang Soo; Kim, Sung Ho; Lee, Chan Bock

    2012-01-01

    Microstructural and mechanical property were evaluated at the medium-sized HT9 (12Cr-1MoWV) forged steel which was considered as primary candidate for the fuel cladding in sodium-cooled fast reactor (SFR). Material was forged at 1170 degrees C after the induction melting to make round bar as 160 mm diameter, 7000 mm length then the radial distribution of microstructure as well as microhardness was evaluated. The results showed that overall microstructure exhibited as ferrite-martensite structure, where small amount (2-3%) of delta ferrite was formed throughout the specimen and maximum 15% of transformed ferrite was formed at the center, where it gradually decreased toward the radial direction. Sensitivity analysis of the cooling curve and Time-Temperature-Transformation (TTT) diagram revealed that formation of transformed ferrite could be avoided when the diameter was decreased down to 120 mm.

  3. Structural instabilities of high temperature alloys and their use in advanced high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Schuster, H.; Ennis, P.J.; Nickel, H.; Czyrska-Filemonowicz, A.

    1989-01-01

    High-temperature, iron-nickel and nickel based alloys are the candidate heat exchanger materials for advanced high temperature gas-cooled reactors supplying process heat for coal gasification, where operation temperatures can reach 850-950 deg. C and service lives of more than 100,000 h are necessary. In the present paper, typical examples of structural changes which occur in two representative alloys (Alloy 800 H, Fe-32Ni-20Cr and Alloy 617, Ni-22Cr-12Co-9Mo-1Al) during high temperature exposure will be given and the effects on the creep rupture properties discussed. At service temperatures, precipitation of carbides occurs which has a significant effect on the creep behaviour, especially in the early stages of creep when the precipitate particles are very fine. During coarsening of the carbides, carbides at grain boundaries restrict grain boundary sliding which retards the development of creep damage. In the service environments, enhanced carbide precipitation may occur due to the ingress of carbon from the environment (carburization). Although the creep rate is not adversely affected, the ductility of the carburized material at low and intermediate temperatures is very low. During simulated service exposures, the formation of surface corrosion scales, the precipitation of carbides and the formation of internal oxides below the surface leads to depletion of the matrix in the alloying elements involved in the corrosion processes. In thin-walled tubes the depletion of Cr due to Cr 2 O 3 formation on the surface can lead to a loss of creep strength. An additional depletion effect resulting from environmental-metal reactions is the loss of carbon (decarburization) which may occur in specific environments. The compositions of the cooling gases which decarburize the material have been determined; they are to be avoided during reactor operation

  4. Experience of the remote dismantling of the Windscale advanced gas-cooled reactor and Windscale pile chimneys

    International Nuclear Information System (INIS)

    Wright, E.M.

    1993-01-01

    This paper gives brief descriptions of some of the remote dismantling work and equipment used on two large decommissioning projects: the BNFL Windscale Pile Chimneys Project (remote handling machine, waste packaging machine, remotely controlled excavator, remotely controlled demolition machine) and the AEA Windscale Advanced Gas-cooled Reactor Project (remote dismantling machine, operational waste, bulk removal techniques, semi-remote cutting operations)

  5. Proceedings of the GCNEP-IAEA course on natural circulation phenomena and passive safety systems in advanced water cooled reactors. V.1

    International Nuclear Information System (INIS)

    2014-01-01

    The current status and prospect, economics, advanced designs and applications of reactors in operation and construction, safety of advanced water cooled reactors is discussed. Papers relevant to INIS are indexed separately

  6. Proceedings of the GCNEP-IAEA course on natural circulation phenomena and passive safety systems in advanced water cooled reactors. V.2

    International Nuclear Information System (INIS)

    2014-01-01

    The current status and prospect, economics, advanced designs and applications of reactors in operation and construction, safety of advanced water cooled reactors is discussed. Papers relevant to INIS are indexed separately

  7. Computation, measurement and analysis of the reactivity-to-power-transfer-function for the sodium cooled nuclear power plant KNK I

    International Nuclear Information System (INIS)

    Hoppe, P.; Mitzel, F.

    1977-02-01

    The Reactivity-to-Power-Transfer-Function for the sodium cooled nuclear power plant KNK I (Kompakte Natriumgekuehlte Kernenergieanlage) has been measured and compared with theoretical results. The measurements have been performed with the help of pseudostochastic reactivity perturbations. The transfer function has been determined by computing the auto- and cross-power-spectral-densities for the reactivity- and neutron flux signals. The agreement between the experimental and theoretical transfer function could be improved by adjusting the reactivity coefficients. The applications of these measurements with respect to reactor diagnosis and malfunction detection are discussed. For this purpose the accuracy of the measured transfer function is of great importance. Therefore an extensive error analysis has been performed. It turned out, that the inherent instability of the reactor without control system and the feedback by the primary coolant system were the reasons for comparatively big systematical errors. The conditions have been derived under which these types of errors can be considerably reduced. The conclusions can also be applied to analogical measurements at fast sodium cooled reactors. Because of their inherent stability the systematical errors will be reduced. (orig.) [de

  8. Metrological certification of systems to monitor the seal integrity of fuel-element cladding based on exposed fuel in sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Eliseev, A.V.; Filonov, V.S.; Ushakov, V.M.; Belov, S.P.; Pedyash, B.V.; Zemtsev, B.V.; Skorikov, N.V.

    1992-01-01

    In sodium-cooled fast reactors, the clad monitoring system for seal integrity of the fuel element cladding is practically the only source of operator information on the serviceability of fuel elements in the core. The monitoring system can be used as the basis for critical decisions whether the reactor must be shut down of whether operation can continue, but only if the meterologically provided measurements are reliable. This article describes a method developed for certifying working rods on the basis of the domestic standard. The method includes a combined irradiation of the sample and the rod to be certified in an arbitrary field of a plutonium-beryllium neutron source with an output rate greater than 10 8 sec -1 , which is mounted in a paraffin moderator. The positive results of the metrological certification of the system to monitor cladding seal integrity leads the authors to recommend this method for other current and planned sodium-cooled fast reactors. 6 refs., 2 tabs

  9. JSFR design progress related to development of safety design criteria for generation IV sodium-cooled fast reactors. (3) Progress of component design

    International Nuclear Information System (INIS)

    Enuma, Yasuhiro; Kawasaki, Nobuchika; Orita, Junichi; Eto, Masao; Miyagawa, Takayuki

    2015-01-01

    In the frame work of generation IV international forum (GIF), safety design criteria (SDC) and safety design guideline (SDG) for the generation IV sodium-cooled fast reactors have been developing in the circumstance of worldwide deployment of SFRs. JAEA, JAPC, MFBR have been investigating design study for JSFR to satisfy SDC in the feasibility study of SDG for Sodium-cooled Fast Reactor (SFR). In addition to the safety measures, maintainability, reparability and manufacturability are taken into account in the JSFR design study. This paper describes the design of main components. Enlargement of the access route for the inspection devices and addition of the access routes were carried out for the reactor structure. The pump-integrated IHX (pump/IHX) was modified for the primary heat exchanger (PHX), which was installed for the decay heat removal in the IHX at the upper plenum, to be removable for improved repair and maintenance. For the steam generator (SG), protective wall tube type design is under investigation as an option with less R and D risks. (author)

  10. Advanced treatment of sodium dithionite wastewater using the combination of coagulation, catalytic ozonation, and SBR.

    Science.gov (United States)

    Zou, Xiao-Ling

    2017-10-01

    A combined process of coagulation-catalytic ozonation-anaerobic sequencing batch reactor (ASBR)-SBR was developed at lab scale for treating a real sodium dithionite wastewater with an initial chemical oxygen demand (COD) of 21,760-22,450 mg/L. Catalytic ozonation with the prepared cerium oxide (CeO 2 )/granular activated carbon catalyst significantly enhances wastewater biodegradability and reduces wastewater microtoxicity. The results show that, under the optimum conditions, the removal efficiencies of COD and suspended solids are averagely 99.3% and 95.6%, respectively, and the quality of final effluent can meet the national discharge standard of China. The coagulation and ASBR processes remove a considerable proportion of organic matter, while the SBR plays an important role in post-polish of final effluent. The ecotoxicity of the wastewater is greatly reduced after undergoing the hybrid treatment. This work demonstrates that the hybrid system has the potential to be applied for the advanced treatment of high-strength industrial wastewater.

  11. Axisymmetric whole pin life modelling of advanced gas-cooled reactor nuclear fuel

    International Nuclear Information System (INIS)

    Mella, R.; Wenman, M.R.

    2013-01-01

    Thermo-mechanical contributions to pellet–clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS’s well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used

  12. Remote handling equipment for the decommissioning of the Windscale Advanced Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Barker, A.; Birss, I.R.; Fish, G.

    1984-01-01

    A decision to decommission the Windscale Advanced Gas Cooled Reactor was taken shortly after reactor shutdown in 1981. The fuel has now been discharged and the decommissioning programme will last about 10-12 years. The paper describes the programme and objectives and deals with methods of handling and disposing of the radioactive waste material. The main new facility required is a Waste Packaging Building adjacent to the existing reactor in which the waste boxes will be filled, active waste encapsulated in concrete and the boxes cleaned, swabbed and monitored to comply with IAEA transport regulations. The handling machine concept and features are described. The assaying and packaging of the waste material, the control of box movement and the process of concrete encapsulation is described. The paper concludes with a description of the development programme to support the Project. The tasks include a study of cutting techniques, production and control of dust and smoke, viewing and lighting methods, filtration, decontamination and fixing of contamination

  13. Sensitivity analysis of an Advanced Gas-cooled Reactor control rod model

    International Nuclear Information System (INIS)

    Scott, M.; Green, P.L.; O’Driscoll, D.; Worden, K.; Sims, N.D.

    2016-01-01

    Highlights: • A model was made of the AGR control rod mechanism. • The aim was to better understand the performance when shutting down the reactor. • The model showed good agreement with test data. • Sensitivity analysis was carried out. • The results demonstrated the robustness of the system. - Abstract: A model has been made of the primary shutdown system of an Advanced Gas-cooled Reactor nuclear power station. The aim of this paper is to explore the use of sensitivity analysis techniques on this model. The two motivations for performing sensitivity analysis are to quantify how much individual uncertain parameters are responsible for the model output uncertainty, and to make predictions about what could happen if one or several parameters were to change. Global sensitivity analysis techniques were used based on Gaussian process emulation; the software package GEM-SA was used to calculate the main effects, the main effect index and the total sensitivity index for each parameter and these were compared to local sensitivity analysis results. The results suggest that the system performance is resistant to adverse changes in several parameters at once.

  14. Multicycle Optimization of Advanced Gas-Cooled Reactor Loading Patterns Using Genetic Algorithms

    International Nuclear Information System (INIS)

    Ziver, A. Kemal; Carter, Jonathan N.; Pain, Christopher C.; Oliveira, Cassiano R.E. de; Goddard, Antony J. H.; Overton, Richard S.

    2003-01-01

    A genetic algorithm (GA)-based optimizer (GAOPT) has been developed for in-core fuel management of advanced gas-cooled reactors (AGRs) at HINKLEY B and HARTLEPOOL, which employ on-load and off-load refueling, respectively. The optimizer has been linked to the reactor analysis code PANTHER for the automated evaluation of loading patterns in a two-dimensional geometry, which is collapsed from the three-dimensional reactor model. GAOPT uses a directed stochastic (Monte Carlo) algorithm to generate initial population members, within predetermined constraints, for use in GAs, which apply the standard genetic operators: selection by tournament, crossover, and mutation. The GAOPT is able to generate and optimize loading patterns for successive reactor cycles (multicycle) within acceptable CPU times even on single-processor systems. The algorithm allows radial shuffling of fuel assemblies in a multicycle refueling optimization, which is constructed to aid long-term core management planning decisions. This paper presents the application of the GA-based optimization to two AGR stations, which apply different in-core management operational rules. Results obtained from the testing of GAOPT are discussed

  15. Genetic algorithms and artificial neural networks for loading pattern optimisation of advanced gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ziver, A.K. E-mail: a.k.ziver@imperial.ac.uk; Pain, C.C; Carter, J.N.; Oliveira, C.R.E. de; Goddard, A.J.H.; Overton, R.S

    2004-03-01

    A non-generational genetic algorithm (GA) has been developed for fuel management optimisation of Advanced Gas-Cooled Reactors, which are operated by British Energy and produce around 20% of the UK's electricity requirements. An evolutionary search is coded using the genetic operators; namely selection by tournament, two-point crossover, mutation and random assessment of population for multi-cycle loading pattern (LP) optimisation. A detailed description of the chromosomes in the genetic algorithm coded is presented. Artificial Neural Networks (ANNs) have been constructed and trained to accelerate the GA-based search during the optimisation process. The whole package, called GAOPT, is linked to the reactor analysis code PANTHER, which performs fresh fuel loading, burn-up and power shaping calculations for each reactor cycle by imposing station-specific safety and operational constraints. GAOPT has been verified by performing a number of tests, which are applied to the Hinkley Point B and Hartlepool reactors. The test results giving loading pattern (LP) scenarios obtained from single and multi-cycle optimisation calculations applied to realistic reactor states of the Hartlepool and Hinkley Point B reactors are discussed. The results have shown that the GA/ANN algorithms developed can help the fuel engineer to optimise loading patterns in an efficient and more profitable way than currently available for multi-cycle refuelling of AGRs. Research leading to parallel GAs applied to LP optimisation are outlined, which can be adapted to present day LWR fuel management problems.

  16. Data Acquisition System Design for Advanced Core-Cooling Mechanism Experiment

    International Nuclear Information System (INIS)

    Zhang, Ziyang; Tian, Fang; Zhang, Tao; Wang, Shen

    2011-01-01

    Data Acquisition System (DAS) design for Advanced Core-Cooling Mechanism Experiment(ACME) is studied in the paper. DAS is an important connection between test facility and result analysis. Firstly, it introduces DAS and its design requirement for ACME. Nearly one thousand data resources need record in ACME. They have different types and acquisition frequencies. In order to record these data, a large scale and high speed layered data acquisition system is developed. Secondly, it discusses the DAS design for ACME, including the analog signal adjusting circuits, clock circuit design, sampling frequencies, data storage and transmission by large database system, anti-interference and etc. Analog signal adjusting circuits are necessary to deal with different kinds of input data to gain standard data resources. Some data change slowly and others change in several seconds according to the test performed on ACME. So it is difficult to use uniform sampling frequencies, and a layered data acquisition system is introduced. A large database is built to store data for ACME test, which keeps data safer and makes subsequent data handling more convenient. A database hot backup is also applied to ensure data safety. The software of DAS is built by Labview, which can provide intuitionist result and friendly interface. Another important function of DAS is the ACME safety protection. Finally, the characteristics and improvement of DAS for ACME is analyzed compared to other test facility. Besides friendly user interface, DAS of ACME can also assure higher data precision and sampling frequency

  17. Mechanical Properties of Advanced Gas-Cooled Reactor Stainless Steel Cladding After Irradiation

    Science.gov (United States)

    Degueldre, Claude; Fahy, James; Kolosov, Oleg; Wilbraham, Richard J.; Döbeli, Max; Renevier, Nathalie; Ball, Jonathan; Ritter, Stefan

    2018-05-01

    The production of helium bubbles in advanced gas-cooled reactor (AGR) cladding could represent a significant hazard for both the mechanical stability and long-term storage of such materials. However, the high radioactivity of AGR cladding after operation presents a significant barrier to the scientific study of the mechanical properties of helium incorporation, said cladding typically being analyzed in industrial hot cells. An alternative non-active approach is to implant He2+ into unused AGR cladding material via an accelerator. Here, a feasibility study of such a process, using sequential implantations of helium in AGR cladding steel with decreasing energy is carried out to mimic the buildup of He (e.g., 50 appm) that would occur for in-reactor AGR clad in layers of the order of 10 µm in depth, is described. The implanted sample is subsequently analyzed by scanning electron microscopy, nanoindentation, atomic force and ultrasonic force microscopies. As expected, the irradiated zones were affected by implantation damage (steel cladding is retained despite He2+ implantation.

  18. Study on in-service inspection and repair program and related plant design for Japan Sodium-Cooled Fast Reactor (JSFR)

    International Nuclear Information System (INIS)

    Kubo, Shigenobu; Suzuki, Shinichi; Kotake, Shoji; Nishiyama, Noboru; Uzawa, Masayuki

    2011-01-01

    Maintenance and repair program and conformity with them were investigated as a part of the conceptual design study of Japan Sodium-cooled Fast Reactor (JSFR). The maintenance program was set by taking the feature of sodium-cooled reactors and domestic practice of LWRs into account. Both of regulatory required inspection and voluntary inspection, which are conducted in the domestic LWRs, were counted. The regulatory required ISI program was based on that of the previous Japanese SFRs, LWRs (JSME S NA1) and liquid metal cooled reactors (ASME section XI division 3). Parts to be inspected, methods of inspection were identified for major structures and components. Concerning the repair program, we set three levels of repair requirements based on estimated frequency of defect and failure during the plant life time. For level 1, which might be occur several times during the plant life time, it is required to be easily repaired in a short period. Access routes and working space are considered in the component design and its arrangement. For level 2, which might be unlikely to occur during the plant life time, it is required to check that the repair work is feasible in a practical time range. For level 3, which frequency is negligible small, repair is not taken into account but the feasibility was investigated. The plant design shall be done so that all of above mentioned inspection and repair can be conducted. It is desired to ensure accessibility for all of the coolant and cover gas boundaries and the internal structures in order to cope with unforeseen troubles. Access routes for the reactor vessel and its internal structures, piping, pumps and intermediate heat exchangers and steam generators were investigated. As the results of that, possible ways for implementation of the maintenance and repair were identified. (author)

  19. A study of sodium-cooled fast breeder reactor with thorium blanket for supply of U-233 to high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Yoshida, H.; Nishimura, H.; Osugi, T.

    1978-08-01

    Symbiotic energy system between fast breeder reactor and thermal reactor would have a potential merit for nuclear proliferation problem. And when using HTGR as the thermal reactor in the system, the energy system appears to be promising as an energy system self-sufficient in fuels, which can generate both electricity and high temperature process heat. In the system the fast breeder reactor has to supply sufficient amount of fissile plutonium to keep the reactor going, and also produce U-233 necessary to the associated U-233 fuelled process heat production HTGR. Three types of LMFBR concepts with thorium blanket, conventional homogeneous core LMFBR, and axial and radial parfait heterogeneous core LMFBRs, have been investigated to find out suitable configurations of LMFBR for supply of U-233 to the HTGR with relatively high conversion ratio of 0.85, in the symbiotic energy system between LMFBR and HTGR. The investigation on LMFBR has been made on fuel sufficiency of the system, inherent safety such as sodium-void and Doppler coefficients, and fuel cycle cost. The followings were revealed; (1) Conventional homogeneous core LMFBR with thorium radial blanket well satisfies the condition of fuel sufficiency, if adequate radial blanket thickness is chosen. However, the sodium-void coefficient and fuel cycle cost are inferior to the other concepts. (2) Axial parfait heterogeneous core LMFBR can be regarded as one of the best LMFBR concepts installed in the symbiotic energy system, from the viewpoints of fuel sufficiency, inherent safety and fuel cycle cost. However, further investigations should be needed on reliability and operationability of the concept. (3) Radial parfait heterogeneous core LMFBR seems inadequate as the LMFBR in the system, because the configurations based on this concept does not satisfy plutonium and U-233 breedings, simultaneously. This LMFBR concept, however, has excellent breeding performance in the internal radial blanket. So further

  20. SnSe2 Two Dimensional Anodes for Advanced Sodium Ion Batteries

    KAUST Repository

    Zhang, Fan

    2017-01-01

    Sodium-ion batteries (SIBs) are considered as a promising alternative to lithium-ion batteries (LIBs) for large-scale renewable energy storage units due to the abundance of sodium resource and its low cost. However, the development of anode

  1. Spacesuit Water Membrane Evaporator; An Enhanced Evaporative Cooling Systems for the Advanced Extravehicular Mobility Unit Portable Life Support System

    Science.gov (United States)

    Bue, Grant C.; Makinen, Janice V.; Miller, Sean.; Campbell, Colin; Lynch, Bill; Vogel, Matt; Craft, Jesse; Petty, Brian

    2014-01-01

    Spacesuit Water Membrane Evaporator - Baseline heat rejection technology for the Portable Life Support System of the Advanced EMU center dot Replaces sublimator in the current EMU center dot Contamination insensitive center dot Can work with Lithium Chloride Absorber Radiator in Spacesuit Evaporator Absorber Radiator (SEAR) to reject heat and reuse evaporated water The Spacesuit Water Membrane Evaporator (SWME) is being developed to replace the sublimator for future generation spacesuits. Water in LCVG absorbs body heat while circulating center dot Warm water pumped through SWME center dot SWME evaporates water vapor, while maintaining liquid water - Cools water center dot Cooled water is then recirculated through LCVG. center dot LCVG water lost due to evaporation (cooling) is replaced from feedwater The Independent TCV Manifold reduces design complexity and manufacturing difficulty of the SWME End Cap. center dot The offset motor for the new BPV reduces the volume profile of the SWME by laying the motor flat on the End Cap alongside the TCV.

  2. Effects of system size and cooling rate on the structure and properties of sodium borosilicate glasses from molecular dynamics simulations.

    Science.gov (United States)

    Deng, Lu; Du, Jincheng

    2018-01-14

    Borosilicate glasses form an important glass forming system in both glass science and technologies. The structure and property changes of borosilicate glasses as a function of thermal history in terms of cooling rate during glass formation and simulation system sizes used in classical molecular dynamics (MD) simulation were investigated with recently developed composition dependent partial charge potentials. Short and medium range structural features such as boron coordination, Si and B Q n distributions, and ring size distributions were analyzed to elucidate the effects of cooling rate and simulation system size on these structure features and selected glass properties such as glass transition temperature, vibration density of states, and mechanical properties. Neutron structure factors, neutron broadened pair distribution functions, and vibrational density of states were calculated and compared with results from experiments as well as ab initio calculations to validate the structure models. The results clearly indicate that both cooling rate and system size play an important role on the structures of these glasses, mainly by affecting the 3 B and 4 B distributions and consequently properties of the glasses. It was also found that different structure features and properties converge at different sizes or cooling rates; thus convergence tests are needed in simulations of the borosilicate glasses depending on the targeted properties. The results also shed light on the complex thermal history dependence on structure and properties in borosilicate glasses and the protocols in MD simulations of these and other glass materials.

  3. Effects of system size and cooling rate on the structure and properties of sodium borosilicate glasses from molecular dynamics simulations

    Science.gov (United States)

    Deng, Lu; Du, Jincheng

    2018-01-01

    Borosilicate glasses form an important glass forming system in both glass science and technologies. The structure and property changes of borosilicate glasses as a function of thermal history in terms of cooling rate during glass formation and simulation system sizes used in classical molecular dynamics (MD) simulation were investigated with recently developed composition dependent partial charge potentials. Short and medium range structural features such as boron coordination, Si and B Qn distributions, and ring size distributions were analyzed to elucidate the effects of cooling rate and simulation system size on these structure features and selected glass properties such as glass transition temperature, vibration density of states, and mechanical properties. Neutron structure factors, neutron broadened pair distribution functions, and vibrational density of states were calculated and compared with results from experiments as well as ab initio calculations to validate the structure models. The results clearly indicate that both cooling rate and system size play an important role on the structures of these glasses, mainly by affecting the 3B and 4B distributions and consequently properties of the glasses. It was also found that different structure features and properties converge at different sizes or cooling rates; thus convergence tests are needed in simulations of the borosilicate glasses depending on the targeted properties. The results also shed light on the complex thermal history dependence on structure and properties in borosilicate glasses and the protocols in MD simulations of these and other glass materials.

  4. Thermodynamic analysis and preliminary design of closed Brayton cycle using nitrogen as working fluid and coupled to small modular Sodium-cooled fast reactor (SM-SFR)

    International Nuclear Information System (INIS)

    Olumayegun, Olumide; Wang, Meihong; Kelsall, Greg

    2017-01-01

    Highlights: • Nitrogen closed Brayton cycle for small modular sodium-cooled fast reactor studied. • Thermodynamic modelling and analysis of closed Brayton cycle performed. • Two-shaft configuration proposed and performance compared to single shaft. • Preliminary design of heat exchangers and turbomachinery carried out. - Abstract: Sodium-cooled fast reactor (SFR) is considered the most promising of the Generation IV reactors for their near-term demonstration of power generation. Small modular SFRs (SM-SFRs) have less investment risk, can be deployed more quickly, are easier to operate and are more flexible in comparison to large nuclear reactor. Currently, SFRs use the proven Rankine steam cycle as the power conversion system. However, a key challenge is to prevent dangerous sodium-water reaction that could happen in SFR coupled to steam cycle. Nitrogen gas is inert and does not react with sodium. Hence, intercooled closed Brayton cycle (CBC) using nitrogen as working fluid and with a single shaft configuration has been one common power conversion system option for possible near-term demonstration of SFR. In this work, a new two shaft nitrogen CBC with parallel turbines was proposed to further simplify the design of the turbomachinery and reduce turbomachinery size without compromising the cycle efficiency. Furthermore, thermodynamic performance analysis and preliminary design of components were carried out in comparison with a reference single shaft nitrogen cycle. Mathematical models in Matlab were developed for steady state thermodynamic analysis of the cycles and for preliminary design of the heat exchangers, turbines and compressors. Studies were performed to investigate the impact of the recuperator minimum terminal temperature difference (TTD) on the overall cycle efficiency and recuperator size. The effect of turbomachinery efficiencies on the overall cycle efficiency was examined. The results showed that the cycle efficiency of the proposed

  5. Subtask 5.10 - Testing of an Advanced Dry Cooling Technology for Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Christopher L. [Univ. of Oklahoma, Norman, OK (United States); Pavlish, John H. [Univ. of Oklahoma, Norman, OK (United States)

    2013-09-30

    The University of North Dakota’s Energy & Environmental Research Center (EERC) is developing a market-focused dry cooling technology that is intended to address the key shortcomings of conventional dry cooling technologies: high capital cost and degraded cooling performance during daytime temperature peaks. The unique aspect of desiccant dry cooling (DDC) is the use of a hygroscopic working fluid—a liquid desiccant—as a heat-transfer medium between a power plant’s steam condenser and the atmosphere. This configuration enables a number of beneficial features for large-scale heat dissipation to the atmosphere, without the consumptive use of cooling water. The overall goal of this project was to accurately define the performance and cost characteristics of DDC to determine if further development of the concept is warranted. A balanced approach of modeling grounded in applied experimentation was pursued to substantiate DDC-modeling efforts and outline the potential for this technology to cool full-scale power plants. The resulting analysis shows that DDC can be a lower-cost dry cooling alternative to an air-cooled condenser (ACC) and can even be competitive with conventional wet recirculating cooling under certain circumstances. This project has also highlighted the key technological steps that must be taken in order to transfer DDC into the marketplace. To address these issues and to offer an extended demonstration of DDC technology, a next-stage project should include the opportunity for outdoor ambient testing of a small DDC cooling cell. This subtask was funded through the EERC–U.S. Department of Energy (DOE) Joint Program on Research and Development for Fossil Energy-Related Resources Cooperative Agreement No. DE-FC26-08NT43291. Nonfederal funding was provided by the Wyoming State Legislature under an award made through the Wyoming Clean Coal Technologies Research Program.

  6. 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

  7. High energy resolution and high count rate gamma spectrometry measurement of primary coolant of generation 4 sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Coulon, R.

    2010-01-01

    Sodium-cooled Fast Reactors are under development for the fourth generation of nuclear reactor. Breeders reactors could gives solutions for the need of energy and the preservation of uranium resources. An other purpose is the radioactive wastes production reduction by transmutation and the control of non-proliferation using a closed-cycle. These thesis shows safety and profit advantages that could be obtained by a new generation of gamma spectrometry system for SFR. Now, the high count rate abilities, allow us to study new methods of accurate power measurement and fast clad failure detection. Simulations have been done and an experimental test has been performed at the French Phenix SFR of the CEA Marcoule showing promising results for these new measurements. (author) [fr

  8. Flow distribution and pressure loss in subchannels of a wire-wrapped 37-pin rod bundle for sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Seok Kyu; Euh, Dong Jin; Choi, Hae Seob; Kim, Hyung Mo; Choi, Sun Rock; Lee, Hyeong Yeon [Thermal-Hydraulic Safety Research Department, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    A hexagonally arrayed 37-pin wire-wrapped rod bundle has been chosen to provide the experimental data of the pressure loss and flow rate in subchannels for validating subchannel analysis codes for the sodium-cooled fast reactor core thermal/hydraulic design. The iso-kinetic sampling method has been adopted to measure the flow rate at subchannels, and newly designed sampling probes which preserve the flow area of subchannels have been devised. Experimental tests have been performed at 20-115% of the nominal flow rate and 60 degrees C (equivalent to Re ∼ 37,100) at the inlet of the test rig. The pressure loss data in three measured subchannels were almost identical regardless of the subchannel locations. The flow rate at each type of subchannel was identified and the flow split factors were evaluated from the measured data. The predicted correlations and the computational fluid dynamics results agreed reasonably with the experimental data.

  9. Inelastic Cyclic Deformation Behaviors of Type 316H Stainless Steel for Reactor Pressure Vessel of Sodium-Cooled Fast Reactor at Elevated Temperatures

    International Nuclear Information System (INIS)

    Yoon, Ji-Hyun; Hong, Seokmin; Koo, Gyeong-Hoi; Lee, Bong-Sang; Kim, Young-Chun

    2015-01-01

    Type 316H stainless steel is a primary candidate material for a reactor pressure vessel of a sodium-cooled fast (SFR) reactor which is under development in Korea. The reactor pressure vessel for a SFR is subjected to inelastic deformation induced by cyclic thermal stress. Fully reversed cyclic testing and ratcheting testing at elevated temperatures were performed to characterize the inelastic cyclic deformation behaviors of Type 316H stainless steel at the SFR operating temperature. It was found that cyclic hardening of Type 316H stainless steel was enhanced, and the accumulation of ratcheting deformation of Type 316H stainless steel was retarded at around the SFR operating temperature. The results of the tensile testing and the microstructural investigation for dislocated structures after the inelastic deformation testing showed that dynamic strain aging affected the inelastic cyclic deformation behavior of Type 316 stainless steel at around the SFR operating temperature.

  10. Third Joint GIF–IAEA Workshop on Safety Design Criteria for Sodium-Cooled Fast Reactors, 26-27 February 2013, Vienna, Austria. Summary Report

    International Nuclear Information System (INIS)

    2013-01-01

    The main objectives of the meeting were to: • Present and share information on the work carried out by GIF, the IAEA and the Member States on the definition of safety design criteria for SFR, including safety approach and requirements on general plant design; • Present the document prepared by the GIF-SFR Task Force on Safety Design Criteria; • Present and discuss safety design concepts of SFRs under development in Member States, with particular emphasis on design measures against Design Basis Accidents and Design Extended Conditions, as well as the associated safety evaluations and supporting R&D; • Draft a room document which should be the basis of the discussion for the Panel on Safety Design Criteria of the FR13 Conference in Paris. • Discuss the results and agree on the future actions of the 3rd Joint GIF-IAEA Workshop on Safety of Sodium-Cooled Fast Reactors

  11. Conceptual design for Japan Sodium-Cooled Fast Reactor. (4) Developmental study of steel plate reinforced concrete containment vessel for JSFR

    International Nuclear Information System (INIS)

    Hosoya, Takusaburo; Negishi, Kazuo; Satoh, Kenichiro; Somaki, Takahiro; Matsuo, Ippei; Shimizu, Katsusuke

    2009-01-01

    An innovative containment vessel, namely Steel plate reinforced Concrete Containment Vessel (SCCV) is developed for Japan Sodium-Cooled Fast Reactor (JSFR). Reducing plant construction cost is one of the most important issues for commercialization of fast reactors. This study investigated construction issues including the building structure and the construction method as well as design issues in terms of the applicability of SCCV to fast reactors. An experimental study including loading and/or heating tests has been carried out to investigate the fundamental structural features, which would be provided to develop methodology to evaluate the feasibility of SCCV under the severe conditions. In this paper, the test plan is described as well as the first test results. (author)

  12. 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)

  13. Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1996-04-01

    The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs

  14. Progress in design, research and development and testing of safety systems for advanced water cooled reactors. Proceedings of a technical committee meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-04-01

    The meeting covered the following topics: Developments in design of safety-related heat removal components and systems for advanced water cooled reactors; status of test programmes on heat removal components and systems of new designs; range of validity and extrapolation of test results for the qualification of design/licensing computer models and codes for advanced water cooled reactors; future needs and trends in testing of safety systems for advanced water cooled reactors. Tests of heat removal safety systems have been conducted by various groups supporting the design, testing and certification of advanced water cooled reactors. The Technical Committee concluded that the reported test results generally confirm the predicted performance features of the advanced designs. Refs, figs, tabs.

  15. Progress of thermal hydraulic evaluation methods and experimental studies on a sodium-cooled fast reactor and its safety in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kamide, Hideki, E-mail: kamide.hideki@jaea.go.jp; Ohshima, Hiroyuki, E-mail: ohshima.hiroyuki@jaea.go.jp; Sakai, Takaaki, E-mail: sakai.takaaki@jaea.go.jp; Tanaka, Masaaki, E-mail: tanaka.masaaki@jaea.go.jp

    2017-02-15

    Highlights: • Thermal hydraulic issues for safety design criteria of sodium cooled fast reactors. • Measurement of velocity data in a subchannel surrounded by wire wrapped fuel-pins. • Statistical evaluation of core hot spot temperature during natural circulation. • Simulation of dynamics of molten fuel pool in a core disruptive accident. • V&V procedure of a multi-dimensional thermal hydraulic code on thermal striping. - Abstract: In the framework of the Generation-IV International Forum, the safety design criteria (SDC) incorporating safety-related R&D results on innovative technologies and lessons learned from Fukushima Dai-ichi nuclear power plants accident has been established to provide the set of general criteria for the safety designs of structures, systems and components of Generation-IV Sodium-cooled Fast Reactors (Gen-IV SFRs). A number of thermal-hydraulic evaluations are necessary to meet the concept of the criteria in the design studies of Gen-IV SFRs. This paper focuses on four kinds of thermal-hydraulic issues associated with the SDC, i.e., fuel subassembly thermal-hydraulics, natural circulation decay heat removal, core disruptive accidents, and thermal striping. Progress of evaluation methods on these issues is shown with activities on verification and validation (V&V) and experimental studies towards commercialization of SFR in Japan. These evaluation methods are planned to be eventually integrated into a comprehensive numerical simulation system that can be applied to all possible phenomena in SFR systems and that can be expected to become an effective tool for the development of human resource and the handing our knowledge and technologies down.

  16. Possibilities of achieving non-positive void reactivity effect in fast sodium-cooled reactors with increased self-protection

    International Nuclear Information System (INIS)

    Alekseev, P.N.; Zverkov, Yu.A.; Morozov, A.G.; Orlov, V.V.; Slesarev, I.S.; Subbotin, S.A.

    1989-01-01

    The problems of self-protection inhancement for the liquid-metal cooled fast reactors with intra-assembly heterogeneity of the core are studied. Possible approaches to arrangement of such reactors with various powers characterized by high levels of coolant natural circulation, minimum reactivity changes during fuel burn-up and non-positive void effect of reactivity are found. 10 refs.; 11 figs

  17. An ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Anish, E-mail: anish@igcar.gov.in; Rajkumar, K.V.; Sharma, Govind K.; Dhayalan, R.; Jayakumar, T.

    2015-02-15

    Highlights: • We demonstrate a novel ultrasonic methodology for in-service inspection of shell weld of core support structure in a sodium cooled fast breeder reactor. • The methodology comprises of the inspection of shell weld immersed in sodium from the outside surface of the main vessel using ultrasonic guided wave. • The formation and propagation of guided wave modes are validated by finite element simulation of the inspection methodology. • A defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably using the developed methodology. - Abstract: The paper presents a novel ultrasonic methodology developed for in-service inspection (ISI) of shell weld of core support structure of main vessel of 500 MWe prototype fast breeder reactor (PFBR). The methodology comprises of the inspection of shell weld immersed in sodium from the outsider surface of the main vessel using a normal beam longitudinal wave ultrasonic transducer. Because of the presence of curvature in the knuckle region of the main vessel, the normal beam longitudinal wave enters the support shell plate at an angle and forms the guided waves by mode conversion and multiple reflections from the boundaries of the shell plate. Hence, this methodology can be used to detect defects in the shell weld of the core support structure. The successful demonstration of the methodology on a mock-up sector made of stainless steel indicated that an artificial defect down to 20% of 30 mm thick wall (∼6 mm) in the shell weld can be detected reliably.

  18. Spacesuit Water Membrane Evaporator; An Enhanced Evaporative Cooling System for the Advanced Extravehicular Mobility Unit Portable Life Support System

    Science.gov (United States)

    Bue, Grant C.; Makinen, Janice V.; Miller, Sean; Campbell, Colin; Lynch, Bill; Vogel, Matt; Craft, Jesse; Wilkes, Robert; Kuehnel, Eric

    2014-01-01

    Development of the Advanced Extravehicular Mobility Unit (AEMU) portable life support subsystem (PLSS) is currently under way at NASA Johnson Space Center. The AEMU PLSS features a new evaporative cooling system, the Generation 4 Spacesuit Water Membrane Evaporator (Gen4 SWME). The SWME offers several advantages when compared with prior crewmember cooling technologies, including the ability to reject heat at increased atmospheric pressures, reduced loop infrastructure, and higher tolerance to fouling. Like its predecessors, Gen4 SWME provides nominal crew member and electronics cooling by flowing water through porous hollow fibers. Water vapor escapes through the hollow fiber pores, thereby cooling the liquid water that remains inside of the fibers. This cooled water is then recirculated to remove heat from the crew member and PLSS electronics. Test results from the backup cooling system which is based on a similar design and the subject of a companion paper, suggested that further volume reductions could be achieved through fiber density optimization. Testing was performed with four fiber bundle configurations ranging from 35,850 fibers to 41,180 fibers. The optimal configuration reduced the Gen4 SWME envelope volume by 15% from that of Gen3 while dramatically increasing the performance margin of the system. A rectangular block design was chosen over the Gen3 cylindrical design, for packaging configurations within the AEMU PLSS envelope. Several important innovations were made in the redesign of the backpressure valve which is used to control evaporation. A twin-port pivot concept was selected from among three low profile valve designs for superior robustness, control and packaging. The backpressure valve motor, the thermal control valve, delta pressure sensors and temperature sensors were incorporated into the manifold endcaps, also for packaging considerations. Flight-like materials including a titanium housing were used for all components. Performance testing

  19. Mitigating the Risk of Stress Corrosion of Austenitic Stainless Steels in Advanced Gas Cooled Reactor Boilers

    International Nuclear Information System (INIS)

    Bull, A.; Owen, J.; Quirk, G.; G, Lewis; Rudge, A.; Woolsey, I.S.

    2012-09-01

    Advanced Gas-Cooled Reactors (AGRs) operated in the UK by EDF Energy have once-through boilers, which deliver superheated steam at high temperature (∼500 deg. C) and pressure (∼150 bar) to the HP turbine. The boilers have either a serpentine or helical geometry for the tubing of the main heat transfer sections of the boiler and each individual tube is fabricated from mild steel, 9%Cr1%Mo and Type 316 austenitic stainless steel tubing. Type 316 austenitic stainless steel is used for the secondary (final) superheater and steam tailpipe sections of the boiler, which, during normal operation, should operate under dry, superheated steam conditions. This is achieved by maintaining a specified margin of superheat at the upper transition joint (UTJ) between the 9%Cr1%Mo primary superheater and the Type 316 secondary superheater sections of the boiler. Operating in this mode should eliminate the possibility of stress corrosion cracking of the Type 316 tube material on-load. In recent years, however, AGRs have suffered a variety of operational problems with their boilers that have made it difficult to maintain the specified superheat margin at the UTJ. In the case of helical boilers, the combined effects of carbon deposition on the gas side and oxide deposition on the waterside of the tubing have resulted in an increasing number of austenitic tubes operating with less than the specified superheat margin at the UTJ and hence the possibility of wetting the austenitic section of the boiler. Some units with serpentine boilers have suffered creep-fatigue damage of the high temperature sections of the boiler, which currently necessitates capping the steam outlet temperature to prevent further damage. The reduction in steam outlet temperature has meant that there is an increased risk of operation with less than the specified superheat margin at the UTJ and hence stress corrosion cracking of the austenitic sections of the boiler. In order to establish the risk of stress

  20. Advanced intermediate temperature sodium-nickel chloride batteries with ultra-high energy density

    Science.gov (United States)

    Li, Guosheng; Lu, Xiaochuan; Kim, Jin Y.; Meinhardt, Kerry D.; Chang, Hee Jung; Canfield, Nathan L.; Sprenkle, Vincent L.

    2016-02-01

    Sodium-metal halide batteries have been considered as one of the more attractive technologies for stationary electrical energy storage, however, they are not used for broader applications despite their relatively well-known redox system. One of the roadblocks hindering market penetration is the high-operating temperature. Here we demonstrate that planar sodium-nickel chloride batteries can be operated at an intermediate temperature of 190 °C with ultra-high energy density. A specific energy density of 350 Wh kg-1, higher than that of conventional tubular sodium-nickel chloride batteries (280 °C), is obtained for planar sodium-nickel chloride batteries operated at 190 °C over a long-term cell test (1,000 cycles), and it attributed to the slower particle growth of the cathode materials at the lower operating temperature. Results reported here demonstrate that planar sodium-nickel chloride batteries operated at an intermediate temperature could greatly benefit this traditional energy storage technology by improving battery energy density, cycle life and reducing material costs.

  1. First meeting of the International Working Group on Advanced Technologies for Water Cooled Reactors, Vienna, 18-21 May 1987. (Pt. 1)

    International Nuclear Information System (INIS)

    1987-12-01

    The first meeting of the IAEA International Working Group on Advanced Technologies for Water Cooled Reactors was held in Vienna, Austria from 18-21 May 1987. Part I of the Summary Report contains the minutes of the meeting

  2. Level II Probabilistic Safety Analysis Methodology for the Application to GEN-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Park, S. Y.; Kim, T. W.; Han, S. H.; Jeong, H. Y.

    2010-03-01

    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 probabilistic safety assessment (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 GEN-IV sodium fast reactor (SFR). An applicability of the PSA methodology of U. S. NRC and PRISM plant to the domestic GEN-IV SFR has been studied. The study contains a plant damage state analysis, a containment event tree analysis, and a source-term release category binning process

  3. Investigations of flow and temperature field development in bare and wire-wrapped reactor fuel pin bundles cooled by sodium

    International Nuclear Information System (INIS)

    Govindha Rasu, N.; Velusamy, K.; Sundararajan, T.; Chellapandi, P.

    2013-01-01

    Highlights: ► We study sodium flow and temperature development in fuel pin bundles. ► Pin diameter, number of pins, wire wrap and ligament gap are varied as parameters. ► Flow development is achieved within ∼30–40 hydraulic diameters. ► Thermal development is attained only for small pin diameter and less number of pins. ► Wire wrap and ligament gap strongly influence Nusselt number. - Abstract: Simultaneous development of liquid sodium flow and temperature fields in the heat generating pin bundles of reactor has been investigated. Development characteristics are seen to be strongly influenced by pin diameter, number of pins, helical wire-wrap, ligament gap between the last row of pins and hexcan wall and Reynolds number. Flow development is achieved within an axial length of ∼125 hydraulic diameters, for all the pin bundle configurations considered. But temperature development is attained only if the pin diameter is small or the number of pins is less. In the case of large pin diameter with more pins, temperature development could not be achieved even after a length of ∼1000 hydraulic diameters. The reason for this behavior is traced to be the weak communication among sub-channels in tightly packed bundles. It is seen that the pin Nusselt number decreases from center to periphery in a bundle. Also, if the ligament gap is narrow, the Nusselt number is large and more uniform. Flow development length is short if the Reynolds number is large and the converse is true for thermal development length. Helical wire-wrap shortens the thermal entry length and significantly enhances the global Nusselt number. But, its influence on hydrodynamic entry length is not significant

  4. First meeting of the International Working Group on Advanced Technologies for Water Cooled Reactors, Vienna, 18-21 May 1987. (Pt. 2)

    International Nuclear Information System (INIS)

    1987-12-01

    The First Meeting of the IAEA International Working Group on Advanced Technologies for Water Cooled Reactors was held in Vienna, Austria from 18-21 May 1987. The Summary Report (Pt. 2) contains the papers which review the national programmes in the field of Advanced Technologies for Water Cooled Reactors and other presentations at the Meeting. A separate abstract was prepared for each of the 10 papers presented at this meeting. Refs, figs

  5. Analysis of clad motion during a loss of flow (LOF) accident in a fast sodium cooled reactor

    International Nuclear Information System (INIS)

    Henkel, P.

    1985-10-01

    A new model describing clad motion during a Loss of Flow (LOF) accident in a Liquid Metal Cooled Fast (Breeder) Reactor (LMFBR) is presented. Its special features are Clad motion is treated within a fuel pin bundle. The bundle geometry is represented by an equivalent annular geometry which serves as the descriptional basis for the clad motion analysis; Several flow regimes are considered. These include a wave or film flow along the fuel pin surfaces as well as a drop flow within the coolant channels. A new entrainment criterion is successfully applied to describe the entrainment of molten cladding and the coolant flow is modelled as a two-dimensional, monstationary flow. Therefore, radial cross flows in a pin bundle can be calculated. Especially, thermal incoherency effects can be treated consistently. The analysis of clad motion in the two experiments STAR1 and STAR2 using the subsequently presented SANDCMOT model gives good agreement with the experimental data. (orig.) [de

  6. Advanced water-cooled reactor technologies. Rationale, state of progress and outlook

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Eighty per cent of the world's power reactors are water cooled and moderated. Many improvements in their design and operation have been implemented since the first such reactor started commercial operation in 1957. This report addresses the safety, environmental and economic rationales for further improvements, as well as their relevance to currently operating water reactors

  7. Nondestructive testing of welds in steam generators for advanced gas cooled reactors at Heyshamm II and Torness

    International Nuclear Information System (INIS)

    Parkin, K.; Bainbridge, A.; Carver, K.; Hammell, R.; Lack, B.J.

    1985-01-01

    The paper concerns non-destructive testing (NDT) of welds in advanced gas cooled steam generators for Heysham II and Torness nuclear power stations. A description is given of the steam generator. The selection of NDT techniques is also outlined, including the factors considered to ascertain the viability of a technique. Examples are given of applied NDT methods which match particular fabrication processes; these include: microfocus radiography, ultrasonic testing of austenitic tube butt welds, gamma-ray isotope projection system, surface crack detection, and automated radiography. Finally, future trends in this field of NDT are highlighted. (UK)

  8. Development of an Alternative Corrosion Inhibitor for the Storage of Advanced Gas-Cooled Reactor Fuel

    International Nuclear Information System (INIS)

    Standring, P.N.; Hands, B.J.; Morgan, S.; Brooks, A.

    2015-01-01

    Sellafield Lt. currently stores AGR fuel in sodium hyrodxide dosed pool water to pH 11.5 to prevent susceptible AGR fuel from failing due to inter-granular attack. The exception to the above storage practice is Thorp Receipt and Storage (TR&S) where an AGR reprocessing buffer is stored in demineralised water as the expected storage durations were short term (up to 5 years). With the extended shut-down of Thorp, storage durations have increased and this has prompted a re-evaluation of the AGR storage regime in TR&S. The use of sodium hydroxide is not feasible due to a compatibility issue with aluminum components used in LWR storage furniture. The implementation process adopted by Sellafield Ltd in developing an alternative corrosion inhibitor for spent AGR fuel is outlined. The two stranded approach evaluates the impact of candidate corrosion inhibitors on fuel integrity and on plant and processes. The development studies in support of the fuel integrity strand are reported. Candidate inhibitors were first evaluated inactively in terms of their ability to arrest propagating corrosion, radiation stability, compatibility with aluminium and environmental impact. Sodium Nitrate was concluded to be the most promising inhibitor. Sodium nitrate was subsequently tested with active AGR brace material. These studies involved the use of bespoke test equipment and techniques. The studies demonstrated that propagating corrosion could be arrested using 10 ppm nitrate and showed that the resultant nitrate film required relatively high chloride concentrations to break it down over the study duration of 60 days. The development studies to date have provided the confidence that sodium nitrate has the potential to be an effective inhibitor for AGR fuel. The final phase of the fuel integrity strand involves a Lead Container Study using whole AGR pins. A staged approach is being adopted in the study programme where proceeding to a more onerous study is not progressed until positive

  9. Whole Core Thermal-Hydraulic Design of a Sodium Cooled Fast Reactor Considering the Gamma Energy Transport

    International Nuclear Information System (INIS)

    Choi, Sun Rock; Back, Min Ho; Park, Won Seok; Kim, Sang Ji

    2012-01-01

    Since a fuel cladding failure is the most important parameter in a core thermal-hydraulic design, the conceptual design stage only involves fuel assemblies. However, although non-fuel assemblies such as control rod, reflector, and B4C generate a relatively smaller thermal power compared to fuel assemblies, they also require independent flow allocation to properly cool down each assembly. The thermal power in non-fuel assemblies is produced from both neutron and gamma energy, and thus the core thermal-hydraulic design including non-fuel assemblies should consider an energy redistribution by the gamma energy transport. To design non-fuel assemblies, the design-limiting parameters should be determined considering the thermal failure modes. While fuel assemblies set a limiting factor with cladding creep temperature to prevent a fission product ejection from the fuel rods, non-fuel assemblies restrict their outlet temperature to minimize thermally induced stress on the upper internal structure (UIS). This work employs a heat generation distribution reflecting both neutron and gamma transport. The whole core thermal-hydraulic design including fuel and non-fuel assemblies is then conducted using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code. The other procedures follow from the previous conceptual design

  10. Studies on the behaviour of a passive containment cooling system for the Indian advanced heavy water reactor

    International Nuclear Information System (INIS)

    Maheshwari, N.K.; Saha, D.; Chandraker, D.K.; Kakodkar, A.; Venkat Raj, V.

    2001-01-01

    A passive containment cooling system has been proposed for the advanced heavy water reactor being designed in India. This is to provide long term cooling for the reactor containment following a loss of coolant accident. The system removes energy released into the containment through immersed condensers kept in a pool of water. An important aspect of immersed condenser's working is the potential degradation of immersed condenser's performance due to the presence of noncondensable gases. An experimental programme to investigate the passive containment cooling system behaviour and performance has been undertaken in a phased manner. In the first phase, system response tests were conducted on a small scale model to understand the phenomena involved. Tests were conducted with constant energy input rate and with varying energy input rate simulating decay heat. With constant energy input rate, pressures in volume V 1 and V 2 reached almost steady value. With varying energy input rate V 1 pressure dropped below the pressure in V 2 . The system could efficiently purge air from V 1 to V 2 . The paper deals with the details of the tests conducted and the results obtained. (orig.) [de

  11. Electrically-cooled HPGe detector for advanced x-ray spectroscopy and imaging

    Energy Technology Data Exchange (ETDEWEB)

    Marian, V.; Clauss, J.; Pirard, B.; Quirin, P.; Flamanc, J.; Lampert, M.O. [CANBERRA France, Parc des Tanneries, 1, chemin de la roseraie, 67380 Lingolsheim (France)

    2015-07-01

    High Purity Germanium (HPGe) detectors are used for high-resolution x- and gamma-ray spectroscopy. For their operation, the necessary cryogenic cooling is performed with liquid nitrogen or with electromechanical coolers. Although mature and industrialized solutions, most of HPGe detectors integrating electrical coolers present a limited spectroscopic performance due to the generated mechanical vibration and electromagnetic interference. This paper describes a novel HPGe detector, specifically designed to address the challenges of ultimate x-ray spectroscopy and imaging applications. Due to the stringent demands associated with nano-scale imaging in synchrotron applications, a custom-designed cryostat was built around a Canberra CP5-Plus electrical cooler featuring extremely low vibration levels and high cooling power. The heat generated by the cryo-cooler itself, as well as the electronics, is evacuated via an original liquid cooling circuit. This architecture can also be used to address high ambient temperature, which does not allow conventional cryo-coolers to work properly. The multichannel detector head can consist of a segmented monolithic HPGe sensor, or several closely packed sensors. Each sensor channel is read out by state-of-the-art pulse-reset preamplifiers in order to achieve excellent energy resolution for count rates in excess of 1 Mcps. The sensitive electronics are located in EMI-proof housings to avoid any interference from other devices on a beam-line. The front-end of the detector is built using selected high-purity materials and alloys to avoid any fluorescence effects. We present a detailed description of the detector design and we report on its performance. A discussion is also given on the use of electrically cooled HPGe detectors for applications requiring ultimate energy resolution, such as synchrotron, medicine or nuclear industry. (authors)

  12. Estimating the occurrence of foreign material in Advanced Gas-cooled Reactors: A Bayesian Monte Carlo approach

    International Nuclear Information System (INIS)

    Mason, Paolo

    2014-01-01

    Highlights: • The amount of a specific type of foreign material found in UK AGRs has been estimated. • The estimate is based on very few instances of detection in numerous inspections. • A Bayesian Monte Carlo approach was used. • The study supports safety case claims on coolant flow impairment. • The methodology is applicable to any inspection campaign on any plant system. - Abstract: The current occurrence of a particular sort of foreign material in eight UK Advanced Gas-cooled Reactors has been estimated by means of a parametric approach. The study includes both variability, treated in analytic fashion via the combination of standard probability distributions, and the uncertainty in the parameters of the model of choice, whose posterior distribution was inferred in Bayesian fashion by means of a Monte Carlo route consisting in the conditional acceptance of sets of model parameters drawn from a prior distribution based on engineering judgement. The model underlying the present study specifically refers to the re-loading and inspection routines of UK Advanced Gas-cooled Reactors. The approach to inference here presented, however, is of general validity and can be applied to the outcome of any inspection campaign on any plant system, and indeed to any situation in which the outcome of a stochastic process is more easily simulated than described by a probability density or mass function

  13. Design measures for prevention and mitigation of severe accidents at advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1998-06-01

    Over 8500 reactor-years of operating experience have been accumulated with the current nuclear energy systems. New generations of nuclear power plants are being developed, building upon this background of experience. During the last decade, requirements for equipment specifically intended to minimize releases of radioactive material to the environment in the event of a core melt accident have been introduced, and designs for new plants include measures for preventing and mitigating a range of severe accident scenarios. The IAEA Technical Committee Meeting on Impact of Severe Accidents on Plant Design and Layout of Advanced Water Cooled Reactors was jointly organized by the Department of Nuclear Energy and the Department of Nuclear Safety to review measures which are being incorporated into advanced water cooled reactor designs for preventing and mitigating severe accidents, the status of experimental and analytical investigations of severe accident phenomena and challenges which support design decisions and accident management procedures, and to understand the impact of explicitly addressing severe accidents on the cost of nuclear power plants. This publication is intended to provide an objective source of information on this topic. It includes 14 papers presented at the Technical Committee meeting held in Vienna between 21-25 October 1996. It also includes a Summary and Findings of the Working Groups. The papers were grouped in three sections. A separate abstract was prepared for each paper

  14. Advanced chip designs and novel cooling techniques for brightness scaling of industrial, high power diode laser bars

    Science.gov (United States)

    Heinemann, S.; McDougall, S. D.; Ryu, G.; Zhao, L.; Liu, X.; Holy, C.; Jiang, C.-L.; Modak, P.; Xiong, Y.; Vethake, T.; Strohmaier, S. G.; Schmidt, B.; Zimer, H.

    2018-02-01

    The advance of high power semiconductor diode laser technology is driven by the rapidly growing industrial laser market, with such high power solid state laser systems requiring ever more reliable diode sources with higher brightness and efficiency at lower cost. In this paper we report simulation and experimental data demonstrating most recent progress in high brightness semiconductor laser bars for industrial applications. The advancements are in three principle areas: vertical laser chip epitaxy design, lateral laser chip current injection control, and chip cooling technology. With such improvements, we demonstrate disk laser pump laser bars with output power over 250W with 60% efficiency at the operating current. Ion implantation was investigated for improved current confinement. Initial lifetime tests show excellent reliability. For direct diode applications 96% polarization are additional requirements. Double sided cooling deploying hard solder and optimized laser design enable single emitter performance also for high fill factor bars and allow further power scaling to more than 350W with 65% peak efficiency with less than 8 degrees slow axis divergence and high polarization.

  15. Studies of S-CO{sub 2} Power Plant Pipe Design for Small Modular Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Seok; Ahn, Yoon Han; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-10-15

    If SFR can be developed into the economical small modular reactor (SMR) for an export from Korea, the expected value can be greater. However, current SFR design may face difficulty in public acceptance due to the potential hazard from sodium-water reaction (SWR) when the current conventional steam Rankine cycle is utilized as a power conversion system for a SFR. In order to eliminate SWR, the Supercritical CO{sub 2} (S-CO{sub 2}) cycle has been proposed. Although there are many researches on S-CO{sub 2} cycle concept and turbomachinery, very few research works considered pipe selection criteria for the S-CO{sub 2} cycle. As one of the most important parts of the plant, this paper will discuss how to select a suitable pipe considering thermal expansion for the S-CO{sub 2} power plant and perform a conceptual design of SFR type SMR. The S-CO{sub 2} cycle can improve the safety of SFR as preventing the SWR by changing the working fluid. Additionally, not only the relatively high efficiency with 450-750 .deg. C turbine inlet temperature, but also the physically compact footprint are advantages of the S-CO{sub 2} cycle. However the pipe design is more complicated than existing power plant because it has high pressure and temperature conditions and needs high mass flow rate. By designing the piping system for a small modular -SFR, the compactness and simplicity of the S-CO{sub 2} cycle are re-confirmed. Moreover, in this paper, realistic and safe pipe design was conducted by considering thermal expansion in the high pressure and temperature conditions. Although total pipe pressure drop is somewhat high, the cycle thermal efficiency is still higher than the existing steam Rankine cycle. Additional study for a larger system such as 300MW class system in MIT report will be conducted in the future study. From the preliminary estimation when the S-CO{sub 2} system becomes large, the pipe diameter may exceed the current ASME standard. This means that more innovative approach

  16. Progress in the development of tooling and dismantling methodologies for the Windscale advanced gas cooled reactor (WAGR)

    International Nuclear Information System (INIS)

    Cross, M.T.; Wareing, M.I.; Dixon, C.

    1998-01-01

    Decommissioning of the Windscale Advanced Gas-Cooled Reactor (WAGR) is a major UK reactor decommissioning project co-funded by the UK Government, the European Commission and Magnox Electric. WAGR was a CO 2 cooled, graphite moderated reactor which served as a test bed for the development of Advanced Gas-Cooled Reactor technology in the UK. It operated from 1963 until shutdown in 1981. AEA Technology plc are currently the Managing Agents on behalf of UKAEA for the WAGR decommissioning project and are responsible for the co-ordination of the project up to the point when the contents of the reactor core and associated radioactive materials are removed and either disposed of or packaged for disposal at some time in the future. Decommissioning has progressed to the point where the reactor has been dismantled down to the level of the hot gas collection manifold with the removal of the top biological shield, the refuelling standpipes and the top section of the reactor pressure vessel. The 4 heat exchangers have also been removed and committed to shallow land burial. This paper describes the work carried out by AEA Technology under separate contracts of UKAEA in developing some of the equipment and deployment methods for the next phase of active operations required in preparation for the dismantling of the core structure. Most recent work has concentrated on the development of specialist tooling for removal of items of operational waste stored within the reactor core, equipment for cutting and removal of the highly radioactive stainless steel 'loop' pressure tubes, diamond wire cutting equipment for sectioning large diameter pipework, and equipment for dismantling the reactor neutron shield. The paper emphasises the process of adaptation and extension of existing technologies for cost-effective application in the decommissioning environment, the need for adequate forward planning of decommissioning methodologies together with large-scale 'mock-up' testing of equipment to

  17. Comparison and validation of the results of the AZNHEX v.1.0 code with the MCNP code simulating the core of a fast reactor cooled with sodium

    International Nuclear Information System (INIS)

    Galicia A, J.; Francois L, J. L.; Bastida O, G. E.; Esquivel E, J.

    2016-09-01

    The development of the AZTLAN platform for the analysis and design of nuclear reactors is led by Instituto Nacional de Investigaciones Nucleares (ININ) and divided into four working groups, which have well-defined activities to achieve significant progress in this project individually and jointly. Within these working groups is the users group, whose main task is to use the codes that make up the AZTLAN platform to provide feedback to the developers, and in this way to make the final versions of the codes are efficient and at the same time reliable and easy to understand. In this paper we present the results provided by the AZNHEX v.1.0 code when simulating the core of a fast reactor cooled with sodium at steady state. The validation of these results is a fundamental part of the platform development and responsibility of the users group, so in this research the results obtained with AZNHEX are compared and analyzed with those provided by the Monte Carlo code MCNP-5, software worldwide used and recognized. A description of the methodology used with MCNP-5 is also presented for the calculation of the interest variables and the difference that is obtained with respect to the calculated with AZNHEX. (Author)

  18. First-principles investigation of neutron-irradiation-induced point defects in B4C, a neutron absorber for sodium-cooled fast nuclear reactors

    Science.gov (United States)

    You, Yan; Yoshida, Katsumi; Yano, Toyohiko

    2018-05-01

    Boron carbide (B4C) is a leading candidate neutron absorber material for sodium-cooled fast nuclear reactors owing to its excellent neutron-capture capability. The formation and migration energies of the neutron-irradiation-induced defects, including vacancies, neutron-capture reaction products, and knocked-out atoms were studied by density functional theory calculations. The vacancy-type defects tend to migrate to the C–B–C chains of B4C, which indicates that the icosahedral cage structures of B4C have strong resistance to neutron irradiation. We found that lithium and helium atoms had significantly lower migration barriers along the rhombohedral (111) plane of B4C than perpendicular to this plane. This implies that the helium and lithium interstitials tended to follow a two-dimensional diffusion regime in B4C at low temperatures which explains the formation of flat disk like helium bubbles experimentally observed in B4C pellets after neutron irradiation. The knocked-out atoms are considered to be annihilated by the recombination of the close pairs of self-interstitials and vacancies.

  19. Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

    Science.gov (United States)

    Guidez, Joel; Saturnin, Anne

    2017-11-01

    During the operation of a nuclear reactor, the external individual doses received by the personnel are measured and recorded, in conformity with the regulations in force. The sum of these measurements enables an evaluation of the annual collective dose expressed in man·Sv/year. This information is a useful tool when comparing the different design types and reactors. This article discusses the evolution of the collective dose for several types of reactors, mainly based on publications from the NEA and the IAEA. The spread of good practices (optimization of working conditions and of the organization, sharing of lessons learned, etc.) and ongoing improvements in reactor design have meant that over time, the doses of various origins received by the personnel have decreased. In the case of sodium-cooled fast reactors (SFRs), the compilation and summarizing of various documentary resources has enabled them to be situated and compared to other types of reactors of the second and third generations (respectively pressurized water reactors in operation and EPR under construction). From these results, it can be seen that the doses received during the operation of SFR are significantly lower for this type of reactor.

  20. Study on velocity field in a wire wrapped fuel pin bundle of sodium cooled reactor. Detailed velocity distribution in a subchannel

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Kobayashi, Jun; Miyakoshi, Hiroyuki; Kamide, Hideki

    2009-01-01

    A sodium cooled fast reactor is designed to attain a high burn-up core in a feasibility study on commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity via change of flow area in the subassembly and influence the heat removal capability. Therefore, it is of importance to obtain the flow velocity distribution in a wire wrapped pin bundle. A 2.5 times enlarged 7-pin bundle water model was applied to investigate the detailed velocity distribution in an inner subchannel surrounded by 3 pins with wrapping wire. The test section consisted of a hexagonal acrylic duct tube and fluorinated resin pins which had nearly the same refractive index with that of water and a high light transmission rate. The velocity distribution in an inner subchannel with the wrapping wire was measured by PIV (Particle Image Velocimetry) through the front and lateral sides of the duct tube. In the vertical velocity distribution in a narrow space between the pins, the wrapping wire decreased the velocity downstream of the wire and asymmetric flow distribution was formed between the pin and wire. In the horizontal velocity distribution, swirl flow around the wrapping wire was obviously observed. The measured velocity data are useful for code validation of pin bundle thermalhydraulics. (author)

  1. Development of core hot spot evaluation method for decay heat removal by natural circulation under transient conditions in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Ohshima, Hiroyuki; Doda, Norihiro; Kamide, Hideki; Watanabe, Osamu; Ohkubo, Yoshiyuki

    2010-01-01

    Toward the commercialization of fast reactors, a design study of Japan Sodium-cooled Fast Reactor (JSFR) is being performed. In this design study, the adoption of decay heat removal system operated by fully natural circulation is being examined from viewpoints of economic competitiveness and passive safety. This paper describes a new evaluation method of core hot spot under transient conditions from forced to natural circulation operations that is necessary for confirming feasibility of the fully natural circulation decay heat removal system. The new method consists of three analysis steps in order to include effects of thermal hydraulic phenomena particular to the natural circulation decay heat removal, e.g., flow redistribution in fuel assemblies caused by buoyancy force, and therefore it enables more rational hot spot evaluation rather than conventional ones. This method was applied to a hot spot evaluation of loss-of-external-power event and the result was compared with those by conventional 1D and detailed 3D simulations. It was confirmed that the proposed method can estimate the hot spot with reasonable degree of conservativeness. (author)

  2. Experimental determination of the local temperature distribution in the cladding tubes of a sodium-cooled pin bundle caused by grid spacers

    International Nuclear Information System (INIS)

    Moeller, R.; Tschoeke, H.

    1980-01-01

    The cladding tubes of reactor core elements are highly stressed structural elements. Their careful design includes the following: (a) the mathematical determination of the maximum cladding tube temperatures; (b) the determination of the maximum permissible fatigue strengths and creep strains of the materials; and (c) the safety distance between the nominal cladding tube hot spots and the permissible extreme cladding tube temperature. The maximum cladding tube temperatures occur on the top edge of the core and, due to radial power gradients, in the wrapper-wall region of a pin bundle. If grid spacers are now used for fixing the pins as in the SNR fuel elements, a careful check must be made of whether and to what degree temperature peaks in the region of the supports have an influence on the cladding tube design. Initial experimental investigations on a sodium-cooled pin bundle model of the SNR-300 fuel element were carried out to throw light on these special problems. This is reported in the following together with the results so far obtained. (U.K.)

  3. Updating of adventitious fuel pin failure frequency in sodium-cooled fast reactors and probabilistic risk assessment on consequent severe accident in Monju

    International Nuclear Information System (INIS)

    Fukano, Yoshitaka; Kurisaka, Kenichi; Nishimura, Masahiro; Naruto, Kenichi

    2015-01-01

    Experimental studies, deterministic approaches and probabilistic risk assessments (PRAs) on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious-fuel-pin-failures (AFPFs) have been considered to be the most dominant initiators of LFs in these PRAs because of their high frequency of occurrence during reactor operation and possibility of fuel-element-failure-propagation (FEFP). A PRA on FEFP from AFPF (FEFPA) in the Japanese prototype SFR (Monju) was performed in this study based on the state-of-the-art knowledge, reflecting the most recent operation procedures under off-normal conditions. Frequency of occurrence of AFPF in SFRs which was the initiating event of the event tree in this PRA was updated using a variety of methods based on the above-mentioned latest review on experiences of this phenomenon. As a result, the frequency of occurrence of, and the core damage frequency (CDF) from, AFPF in Monju was significantly reduced to a negligible magnitude compared with those in the existing PRAs. It was, therefore concluded that the CDF of FEFPA in Monju could be comprised in that of anticipated transient without scram or protected loss of heat sink events from both the viewpoint of occurrence probability and consequences. (author)

  4. SnSe2 Two Dimensional Anodes for Advanced Sodium Ion Batteries

    KAUST Repository

    Zhang, Fan

    2017-05-30

    Sodium-ion batteries (SIBs) are considered as a promising alternative to lithium-ion batteries (LIBs) for large-scale renewable energy storage units due to the abundance of sodium resource and its low cost. However, the development of anode materials for SIBs to date has been mainly limited to some traditional anodes for LIBs, such as carbonaceous materials. SnSe2 is a member of two dimensional layered transition metal dichalcogenide (TMD) family, which has been predicted to have high theoretical capacity as anode material for sodium ion batteries (756 mAh g-1), thanks to its layered crystal structure. Yet, there have been no studies on using SnSe2 as Na ion battery anode. In this thesis, we developed a simple synthesis method to prepare pure SnSe2 nanosheets, employing N2 saturated NaHSe solution as a new selenium source. The SnSe2 2D sheets achieve theoretical capacity during the first cycle, and a stable and reversible specific capacity of 515 mAh g-1 at 0.1 A g-1 after 100 cycles, with excellent rate performance. Among all of the reported transition metal selenides, our SnSe2 sample has the highest reversible capacity and the best rate performances. A combination of ex-situ high resolution transmission electron microscopy (HRTEM) and X-ray diffraction was used to study the mechanism of sodiation and desodiation process in this SnSe2, and to understand the reason for the excellent results that we have obtained. The analysis indicate that a combination of conversion and alloying reactions take place with SnSe2 anodes during battery operation, which helps to explain the high capacity of SnSe2 anodes for SIBs compared to other binary selenides. Density functional theory was used to elucidate the volume changes taking place in this important 2D material.

  5. Hypothetical air ingress scenarios in advanced modular high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    Kroeger, P.G.

    1988-01-01

    Considering an extremely hypothetical scenario of complete cross duct failure and unlimited air supply into the reactor vessel of a modular high temperature gas cooled ractor, it is found that the potential air inflow remains limited due to the high friction pressure drop through the active core. All incoming air will be oxidized to CO and some local external burning would be temporarily possible in such a scenario. The accident would have to continue with unlimited air supply for hundreds of hours before the core structural integrity would be jeopardized

  6. Characterization of natural circulation looping of emergency cooling systems in naval and advanced reactors

    International Nuclear Information System (INIS)

    Macedo, Luiz Alberto; Baptista Filho, Benedito Dias

    2000-01-01

    This paper describes the natural circuit looping, resumes the main project characteristics, presents results of the hydraulic characterization, consisting of pressure loss measurements, and presents results from calibration tests of the power and flow measurements and the first experiments in natural circulation. Those experiments comprised transients in natural circulation with application of application of power steps. The results shown a non linear behaviour of the magnetic flow meter and a dependence on the fluid temperature as well. The assembly circuit/instrumentation/data acquisition system is suitable for the research on emergency cooling passive systems

  7. Advanced phase change materials and systems for solar passive heating and cooling of residential buildings

    Energy Technology Data Exchange (ETDEWEB)

    Salyer, I.O.; Sircar, A.K.; Dantiki, S.

    1988-01-01

    During the last three years under the sponsorship of the DOE Solar Passive Division, the University of Dayton Research Institute (UDRI) has investigated four phase change material (PCM) systems for utility in thermal energy storage for solar passive heating and cooling applications. From this research on the basis of cost, performance, containment, and environmental acceptability, we have selected as our current and most promising series of candidate phase change materials, C-15 to C-24 linear crystalline alkyl hydrocarbons. The major part of the research during this contract period was directed toward the following three objectives. Find, test, and develop low-cost effective phase change materials (PCM) that melt and freeze sharply in the comfort temperature range of 73--77{degree}F for use in solar passive heating and cooling of buildings. Define practical materials and processes for fire retarding plasterboard/PCM building products. Develop cost-effective methods for incorporating PCM into building construction materials (concrete, plasterboard, etc.) which will lead to the commercial manufacture and sale of PCM-containing products resulting in significant energy conservation.

  8. Reduced Volume Prototype Spacesuit Water Membrane Evaporator; A Next-Generation Evaporative Cooling System for the Advanced Extravehicular Mobility Unit Portable Life Support System

    Science.gov (United States)

    Makinen, Janice V.; Anchondo, Ian; Bue, Grant C.; Campbell, Colin; Colunga, Aaron

    2013-01-01

    Development of the Advanced Extravehicular Mobility Unit (AEMU) portable life support subsystem (PLSS) is currently under way at NASA Johnson Space Center. The AEMU PLSS features a new evaporative cooling system, the reduced volume prototype (RVP) spacesuit water membrane evaporator (SWME). The RVP SWME is the third generation of hollow fiber SWME hardware. Like its predecessors, RVP SWME provides nominal crew member and electronics cooling by flowing water through porous hollow fibers. Water vapor escapes through the hollow fiber pores, thereby cooling the liquid water that remains inside of the fibers. This cooled water is then recirculated to remove heat from the crew member and PLSS electronics. Major design improvements, including a 36% reduction in volume, reduced weight, and a more flight-like backpressure valve, facilitate the packaging of RVP SWME in the AEMU PLSS envelope. The development of these evaporative cooling systems will contribute to a more robust and comprehensive AEMU PLSS.

  9. Sodium fast reactors with closed fuel cycle

    CERN Document Server

    Raj, Baldev; Vasudeva Rao, PR 0

    2015-01-01

    Sodium Fast Reactors with Closed Fuel Cycle delivers a detailed discussion of an important technology that is being harnessed for commercial energy production in many parts of the world. Presenting the state of the art of sodium-cooled fast reactors with closed fuel cycles, this book:Offers in-depth coverage of reactor physics, materials, design, safety analysis, validations, engineering, construction, and commissioning aspectsFeatures a special chapter on allied sciences to highlight advanced reactor core materials, specialized manufacturing technologies, chemical sensors, in-service inspecti

  10. Recent IAEA activities to support advanced water cooled reactor technology development

    International Nuclear Information System (INIS)

    Choi, J.-H.; Bilbao y Leon, S.; Rao, A.S.

    2009-01-01

    The International Atomic Energy Agency (IAEA) is the world's center of cooperation in the nuclear field. The IAEA works with its Member States and multiple partners worldwide to promote safe, secure and peaceful nuclear technologies. To catalyse innovation in nuclear power technology in Member States, the IAEA coordinates cooperative research, promotes information exchange, and analyses technical data and results, with a focus on reducing capital costs and construction periods while further improving performance, safety and proliferation resistance. This paper summarizes the recent major IAEA activities to support technology development for water cooled reactors, which is the most common type of reactor design at present and will probably still be in the near future. (author)

  11. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 1. Numerical investigation for the rationalization of hydrodynamics in the upper plenum

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Yamaguchi, Akira

    2002-02-01

    A large-scale sodium-cooled fast breeder reactor in feasibility studies on commercialized fast reactors has a tendency of consideration of thorough simplified and compacted system designs to realize drastic economical improvements. Therefore, special attention should be paid to thermohydraulic designs for a gas entrainment behavior from free surfaces, a flow-induced vibration of in-vessel components, a thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermohydraulic characteristics in the upper plenum, and to investigate suitable in-vessel structure for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) It is difficult to rationale in-vessel flow patterns through adjustments of porous ratio and pressure loss for a hold down plate and baffle plates installed in an upper core structure. (2) Dummy plug insertion to a slit of the upper core structure is one of effective measures to stabilize in-vessel flow patterns. (3) Flow guide devices such as a baffle ring and a partial inner barrel are also effective measures to eliminate impinging jet to a dipped plate (D/P) and to reduce horizontal flow velocity components at free surface. (4) Installations of labyrinth structures to a R/V - D/P gap is successful for decreasing of free surface horizontal flows. (5) Gap closing of an in-vessel fuel pot and two cold trap components has the effects of reductions for free surface horizontal flows and for the difference of free surface levels. Following future investigations are important preventive measures against the gas entrainment from the free surface. (1) Flattening of spatial axial velocity distributions at the R/V - D/P gap. (2) Alleviation measures of vortex concentration at free surface. (3) Separation measures of 3-dimensional vortex

  12. Safety analyses for sodium-cooled fast reactors with pelletized and sphere-pac oxide fuels within the FP-7 European project PELGRIMM - 15386

    International Nuclear Information System (INIS)

    Maschek, W.; Andriolo, L.; Matzerath-Boccaccini, C.; Delage, F.; Parisi, C.; Del Nevo, A.; Abbate, G.; Schmitt, D.

    2015-01-01

    The European FP-7 project PELGRIMM addresses the development of Minor-Actinide (MA) bearing oxide fuel for Sodium-cooled Fast Reactors. Optionally, both MA homogeneous recycling and heterogeneous recycling is investigated with pellet and sphere-pac fuel. A first safety assessment of sphere-pac fuelled cores should be given in the Work Package 4 of the project. This assessment is in continuity with the former FP-7 CP-ESFR project. Within the CP-ESFR project the CONF2 core design has been developed characterized by a core with a large upper sodium plenum to reduce the coolant void worth. This optimized core has been chosen for the safety analyses in PELGRIMM. The task within the PELGRIMM project is thus a safety assessment of the CONF2 core loaded either with pellets or with sphere-pac fuel. The investigations started with the design of the CONF2 core with sphere-pac fuel and the determination of core safety parameters and burn-up behavior. The neutronic analyses have been performed with the MCNPX code. Variants of the CONF2 core contain up to 4% Am in the fuel. The results revealed an extended void worth (core + upper plenum) for an Am free core of 1 up to 3 dollars for the 4% Am core. Thermal-hydraulic design analyses have been performed by RELAP5-3D. The accident simulations should be performed by different codes, some of which focus on the initiation phase of the accident, as SAS4A, BELLA and the MAT5DYN code, whereas the SIMMER-III code will also deal with the later accident phases and a potential whole core melting. The codes had to be adapted to the specifics of the sphere-pac fuel, in particular to the thermal conductivity and gap conditions. Analyses showed that the safety assessment has to take into account two main phases. Starting up the core, the green fuel shows a reduced fuel thermal conductivity. After restructuring within a couple of hours, the thermal conductivity recovers and the fuel temperature decreases. The main objective of the safety analyses

  13. Regional cooling caused recent New Zealand glacier advances in a period of global warming.

    Science.gov (United States)

    Mackintosh, Andrew N; Anderson, Brian M; Lorrey, Andrew M; Renwick, James A; Frei, Prisco; Dean, Sam M

    2017-02-14

    Glaciers experienced worldwide retreat during the twentieth and early twenty first centuries, and the negative trend in global glacier mass balance since the early 1990s is predominantly a response to anthropogenic climate warming. The exceptional terminus advance of some glaciers during recent global warming is thought to relate to locally specific climate conditions, such as increased precipitation. In New Zealand, at least 58 glaciers advanced between 1983 and 2008, and Franz Josef and Fox glaciers advanced nearly continuously during this time. Here we show that the glacier advance phase resulted predominantly from discrete periods of reduced air temperature, rather than increased precipitation. The lower temperatures were associated with anomalous southerly winds and low sea surface temperature in the Tasman Sea region. These conditions result from variability in the structure of the extratropical atmospheric circulation over the South Pacific. While this sequence of climate variability and its effect on New Zealand glaciers is unusual on a global scale, it remains consistent with a climate system that is being modified by humans.

  14. Regulatory Considerations for the Long Term Cooling Safe Shutdown Requirements of the Passive Residual Heat Removal Systems in Advanced Reactors

    International Nuclear Information System (INIS)

    Sim, S. K.; Bae, S. H.; Kim, Y. S.; Hwang, Min Jeong; Bang, Young Seok; Hwang, Taesuk

    2016-01-01

    USNRC approved safe shutdown at 215.6 .deg. C for a safe and long term cooling state for the redundant passive RHRSs by SECY-94-084. USNRC issued COLA(Combined Construction and Operating License) for the Levy County NP Unit-1/2 for the AP1000 passive RHRSs in 2014. Korea Hydro and Nuclear Power(KHNP) is developing APR+ and adopted Passive Auxiliary Feedwater System(PAFS) as a new passive RHRS design. Korea Institute of Nuclear Safety(KINS) has been developing regulatory guides for the advanced safety design features of the advanced ALWRs which has plan to construct in near future in Korea[5]. Safety and regulatory issues as well as the safe shut down requirements of the passive RHRS are discussed and considerations in developing regulatory guides for the passive RHRS are presented herein. Passive RHRSs have been introduced as new safety design features for the advanced reactors under development in Korea. These passive RHRSs have potential advantages over existing active RHRS, however, their functions are limited due to inherent ability of passive heat removal processes. It is high time to evaluate the performance of the passive PRHRs and develop regulatory guides for the safety as well as the performance analyses of the passive RHRS

  15. Benchmarking of thermalhydraulic loop models for lead-alloy-cooled advanced nuclear energy systems. Phase I: Isothermal forced convection case

    International Nuclear Information System (INIS)

    2012-06-01

    Under the auspices of the NEA Nuclear Science Committee (NSC), the Working Party on Scientific Issues of the Fuel Cycle (WPFC) has been established to co-ordinate scientific activities regarding various existing and advanced nuclear fuel cycles, including advanced reactor systems, associated chemistry and flowsheets, development and performance of fuel and materials and accelerators and spallation targets. The WPFC has different expert groups to cover a wide range of scientific issues in the field of nuclear fuel cycle. The Task Force on Lead-Alloy-Cooled Advanced Nuclear Energy Systems (LACANES) was created in 2006 to study thermal-hydraulic characteristics of heavy liquid metal coolant loop. The objectives of the task force are to (1) validate thermal-hydraulic loop models for application to LACANES design analysis in participating organisations, by benchmarking with a set of well-characterised lead-alloy coolant loop test data, (2) establish guidelines for quantifying thermal-hydraulic modelling parameters related to friction and heat transfer by lead-alloy coolant and (3) identify specific issues, either in modelling and/or in loop testing, which need to be addressed via possible future work. Nine participants from seven different institutes participated in the first phase of the benchmark. This report provides details of the benchmark specifications, method and code characteristics and results of the preliminary study: pressure loss coefficient and Phase-I. A comparison and analysis of the results will be performed together with Phase-II

  16. Stability, rheology and thermal analysis of functionalized alumina- thermal oil-based nanofluids for advanced cooling systems

    International Nuclear Information System (INIS)

    Ilyas, Suhaib Umer; Pendyala, Rajashekhar; Narahari, Marneni; Susin, Lim

    2017-01-01

    Highlights: • Alumina nanoparticles are functionalized with oleic acid. • Functionalization of alumina nanoparticles gives better dispersion in thermal oil. • Thermophysical properties of nanofluids are experimentally measured. • TGA confirms the improvement in life of nanofluids. - Abstract: Thermal oils are widely used as cooling media in heat transfer processes. However, their potential has not been utilised exquisitely in many applications due to low thermal properties. Thermal oil-based nanofluids are prepared by dispersing functionalized alumina with varying concentrations of 0.5–3 wt.% to enhance thermal properties of oil for advanced cooling systems. The oleic acid coated alumina is prepared and then dispersed in the oil to overcome the aggregation of nanoparticles in base fluid. The surface characterizations of functionalized nanoparticles are performed using different analysis such as XRD, EDS, SEM, TEM and FTIR. Dispersion behaviour and agglomeration studies are conducted at natural and functionalized conditions using different analysis to ensure long-term stability of nanofluids. In addition, rheological behaviour of non-Newtonian nanofluids is studied at high shear rates (100–2000 s"−"1). Effective densities and enhancement in thermal conductivities are measured for different nanofluids concentrations. Specific heat capacity is measured using Differential Scanning Calorimetry. The correlations are developed for thermophysical properties of nanofluids. Thermogravimetric analysis is performed with respect to temperature and time to exploit the effect of the addition of nanoparticles on the degradation of nanofluids. Significant improvement in the thermal properties of oil is observed using highly stable functionalized alumina nano-additives.

  17. Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Greene, Sherrell R [ORNL; Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Carbajo, Juan J [ORNL; Ilas, Dan [ORNL; Cisneros, Anselmo T [ORNL; Varma, Venugopal Koikal [ORNL; Corwin, William R [ORNL; Wilson, Dane F [ORNL; Yoder Jr, Graydon L [ORNL; Qualls, A L [ORNL; Peretz, Fred J [ORNL; Flanagan, George F [ORNL; Clayton, Dwight A [ORNL; Bradley, Eric Craig [ORNL; Bell, Gary L [ORNL; Hunn, John D [ORNL; Pappano, Peter J [ORNL; Cetiner, Sacit M [ORNL

    2011-02-01

    This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

  18. Development of advanced fabrication technology for high-temperature gas-cooled reactor fuel. Reduction of coating failure fraction

    International Nuclear Information System (INIS)

    Minato, Kazuo; Kikuchi, Hironobu; Fukuda, Kousaku; Tobita, Tsutomu; Yoshimuta, Sigeharu; Suzuki, Nobuyuki; Tomimoto, Hiroshi; Nishimura, Kazuhisa; Oda, Takafumi

    1998-11-01

    The advanced fabrication technology for high-temperature gas-cooled reactor fuel has been developed to reduce the coating failure fraction of the fuel particles, which leads to an improvement of the reactor safety. The present report reviews the results of the relevant work. The mechanisms of the coating failure of the fuel particles during coating and compaction processes of the fuel fabrication were studied to determine a way to reduce the coating failure fraction of the fuel. The coating process was improved by optimizing the mode of the particle fluidization and by developing the process without unloading and loading of the particles at intermediate coating process. The compaction process was improved by optimizing the combination of the pressing temperature and the pressing speed of the overcoated particles. Through these modifications of the fabrication process, the quality of the fuel was improved outstandingly. (author)

  19. Preparation for Future Defuelling and Decommissioning Works on EDF Energy's UK Fleet of Advanced Gas Cooled Reactors

    International Nuclear Information System (INIS)

    Bryers, John; Ashmead, Simon

    2016-01-01

    EDF Energy/Nuclear Generation is the owner and operator of 14 Advanced Gas cooled Reactors (AGR) and one Pressurised Water Reactor (PWR), on 8 nuclear stations in the UK. EDF Energy/Nuclear Generation is responsible for all the activities associated with the end of life of its nuclear installations: de-fuelling, decommissioning and waste management. As the first AGR is forecast to cease generation within 10 years, EDF Energy has started planning for the decommissioning. This paper covers: - broad outline of the technical strategy and arrangements for future de-fuelling and decommissioning works on the UK AGR fleet, - high level strategic drivers and alignment with wider UK nuclear policy, - overall programme of preparation and initial works, - technical approaches to be adopted during decommissioning. (authors)

  20. Analytical evaluation of local fault in sodium cooled small fast reactor (4S). Preliminary evaluation of partial blockage in coolant channel

    International Nuclear Information System (INIS)

    Nishimura, Satoshi; Ueda, Nobuyuki

    2007-01-01

    Local faults are fuel failures that result from heat removal imbalance within a single subassembly especially in FBRs. Although the occurrence frequency of local faults is quite low, the licensing body required local faults evaluations in previous FBR plants to confirm the potential for the occurrence of severe fuel subassembly failure and its propagation. A conceptual design of 4S (Super-Safe, Small and Simple) is a sodium cooled fast reactor, which aims at an application to dispersed energy source and long core lifetime. It has a dense arrangement of fuel pins to achieve a long lifetime. Therefore, from the viewpoint of thermal hydraulics, the 4S reactor is considered to have more potential for coolant boiling and fuel pin failure caused by formation of local blockage, comparing these potential in the conventional FBRs. The objective of the present study is to evaluate the effect of local blockage on the coolant flow pattern and temperature rise in the 4S-type fuel subassembly under the normal operation condition. A series of three-dimensional thermal-hydraulic analysis in a single subassembly with local blockage was conducted by the commercialized CFD code 'PHOENICS'. Analytical results show that the peak coolant temperature behind the blockage rises with increasing the blockage area, however, the coolant boiling does not occur under the present analytical conditions. On the other hand, it is found that the liquid phase formation caused by eutectic reactions will occur between the metallic fuel and the cladding under the local blockage condition. However, the penetration rate of liquid phase at fuel-cladding interface is quit low. Therefore, it is expected that rapid fuel pin failure and its propagation to surrounding pins due to liquid phase formation will not occur. (author)