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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. Sodium-cooled nuclear reactors

    International Nuclear Information System (INIS)

    Berthoud, Georges; Ducros, Gerard; Feron, Damien; Guerin, Yannick; Latge, Christian; Limoge, Yves; Santarini, Gerard; Seiler, Jean-Marie; Vernaz, Etienne; Guidez, Joel; Andrieux, Catherine; Baque, Francois; Bonin, Bernard; Boullis, Bernard; Cabet, Celine; Carre, Frank; Dufour, Philippe; Gauche, Francois; Grouiller, Jean-Paul; Jeannot, Jean-Philippe; Le Flem, Marion; Le Coz, Pierre; Martin, Laurent; Masson, Michel; Mathonniere, Gilles; Nokhamzon, Jean-Guy; Pelletier, Michel; Rodriguez, Gilles; Saez, Manuel; Seran, Jean-Louis; Varaine, Frederic; Zaetta, Alain; Behar, Christophe; Provitina, Olivier; Lecomte, Michael; Forestier, Alain; Bender, Alexandra; Parisot, Jean-Francois; Finot, Pierre

    2014-01-01

    This book first explains the choice of sodium-cooled reactors by outlining the reasons of the choice of fast neutron reactors (fast neutrons instead of thermal neutrons, recycling opportunity for plutonium, full use of natural uranium, nuclear waste optimization, flexibility of fast neutron reactors in nuclear material management, fast neutron reactors as complements of water-cooled reactors), and by outlining the reasons for the choice of sodium as heat-transfer material. Physical, chemical, and neutron properties of sodium are presented. The second part of the book first presents the main design principles for sodium-cooled fast neutron reactors and their core. The third part proposes an historical overview and an assessment of previously operated sodium-cooled fast neutron reactors (French reactors from Rapsodie to Superphenix, other reactors in the world), and an assessment of the main incidents which occurred in these reactors. It also reports the experience and lessons learned from the dismantling of various sodium-cooled fast breeder reactors in the world. The next chapter addresses safety issues (technical and safety aspects related to the use of sodium) and environmental issues (dosimetry, gaseous and liquid releases, solid wastes, and cooling water). Then, various technological aspects of these reactors are addressed: the energy conversion system, main components, sodium chemistry, sodium-related technology, advances in in-service inspection, materials used in reactors and their behaviour, and fuel system. The next chapter addresses the fuel cycle in these reactors: its integrated specific character, report of the French experience in fast neutron reactor fuel processing, description of the transmutation of minor actinides in these reactors. The last chapter proposes an overview of reactors currently projected or under construction in the world, presents the Astrid project, and gives an assessment of the economy of these reactors. A glossary and an index

  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. 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)

  6. 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.

  7. 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

  8. 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.

  9. 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

  10. 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.

  11. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Home; Journals; Bulletin of Materials Science; Volume 32; Issue 3 ... Nuclear energy; fast breeder reactors; materials science; stainless steels; sodium. ... 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 ...

  12. 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.

  13. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Administrator

    Abstract. 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 reac- tors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with.

  14. 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 ...

  15. 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)

  16. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, 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.

  17. 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)

  18. 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

  19. 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)

  20. 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

  1. 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)

  2. 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

  3. 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

  4. Advances in Solar Heating and Cooling Systems

    Science.gov (United States)

    Ward, Dan S.

    1976-01-01

    Reports on technological advancements in the fields of solar collectors, thermal storage systems, and solar heating and cooling systems. Diagrams aid in the understanding of the thermodynamics of the systems. (CP)

  5. 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)

  6. 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)

  7. 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

  8. 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

  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. 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.

  11. An Innovative Hybrid Loop-Pool Design for Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Haihua Zhao; Hongbin Zhang

    2007-01-01

    The existing sodium cooled fast reactors (SFR) have two types of designs--loop type and pool type. In the loop type design, such as JOYO (Japan) [1] and MONJU (Japan), the primary coolant is circulated through intermediate heat exchangers (IHX) external to the reactor tank. The major advantages of loop design include compactness and easy maintenance. The disadvantage is higher possibility of sodium leakage. In the pool type design such as EBR-II (USA), BN-600M(Russia), Superphenix (France) and European Fast Reactor [2], the reactor core, primary pumps, IHXs and direct reactor auxiliary cooling system (DRACS) heat exchangers (DHX) all are immersed in a pool of sodium coolant within the reactor vessel, making a loss of primary coolant extremely unlikely. However, the pool type design makes primary system large. In the latest ANL's Advanced Burner Test Reactor (ABTR) design [3], the primary system is configured in a pool-type arrangement. The hot sodium at core outlet temperature in hot pool is separated from the cold sodium at core inlet temperature in cold pool by a single integrated structure called Redan. Redan provides the exchange of the hot sodium from hot pool to cold pool through IHXs. The IHXs were chosen as the traditional tube-shell design. This type of IHXs is large in size and hence large reactor vessel is needed

  12. 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

  13. 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.

  14. C-scope under-sodium viewer for sodium-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Miyazawa, Tatsuo; Uesugi, Nobuo; Iguchi, Tatsuo; Taguchi, Junzo; Takagi, Nobuyuki

    1976-01-01

    A C-scope under-sodium viewer has been developed for monitoring the interior of sodium-cooled fast breeder reactors. Consisting of a transducer that emits and receives ultrasonic waves under liquid sodium, a mechanism that drives the transducer under liquid sodium and an image displaying section, it inspects the fuel assembly through its image in optically opaque high-temperature (300 0 C) liquid sodium. The results of its evaluation test are: (1) The transducer could continue satisfactory operation under 350 0 C (at the highest) sodium for more than a month. (2) The driving mechanism, though it was the first of the kind appearing in Japan, has been proved that it could continue operation for a week under 300 0 C sodium. (3) The image displaying section, in spite of the low speed of the transducer (below 20 rpm), could display stable and clear images. (4) The image in 300 0 C was as clear as that in room-temperature water. (auth.)

  15. 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.)

  16. 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.

  17. 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.

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. 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

  3. The Effects of Internal Components' Disposition on Thermal-Hydraulic Behaviors in Sodium Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Han, Ji Woong; Eoh, Jae Hyuk; Kim, Seong O

    2009-01-01

    Decay heat removal is very important in a nuclear power plant. The KALIMER-600, Korea Advanced Liquid MEtal Reactor, employs the PDRC(Passive Decay heat Removal Circuit) to remove the decay heat. However the cooling performance before the activation of DHX greatly depends on the natural circulation flow within the reactor pool. In the previous studies the effect of various design parameters such as coastdown flow, IHX(Intermediate Heat eXchanger) elevation and heat transfer via CCS (Cavity Cooling System) on the initial cooling performance has been analyzed. In the case of IHX elevation analysis the increase of IHX elevation was shown to enhance the initial cooling performance. However, the elevating the IHX is accompanied by the variation of hot or cold pool volume, the previous calculation was resulted from the combination of those effects. In order to analyze those effects qualitatively supplementary calculation conditions were prepared and related analyses have been done in this study. In those analyses the ratio between hot and cold pool volumes has been varied without elevating the IHX by changing the vertical position of separation plate and baffle plate. The COMMIX-1AR/P code is utilized as a tool to investigate overall transient behaviors within a pool. This study is expected to provide the basic information for the decision of internal components' layout in the sodium cooled fast reactor

  4. 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

  5. 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)

  6. 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)

  7. 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

  8. Apparatus for the removal of after heat in a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Cachera, P.C.

    1976-01-01

    In a fast reactor in which each cooling loop comprises the primary sodium circuit which exchanges heat with the sodium of a secondary circuit by means of an intermediate heat exchanger, each cooling loop comprises in parallel with the secondary sodium circuit an auxiliary secondary circuit in which a chemically inert gas extracts heat from the primary sodium, said secondary auxiliary circuit being equipped with at least one gas turbine which is supplied with the inert gas and operates in a closed energy-producing gas cycle

  9. 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

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

    Science.gov (United States)

    Modarres, Dariush

    1991-01-01

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

  11. 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

  12. 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)

  13. Advanced Spectral Library (ASTRAL): Cool stars edition

    Science.gov (United States)

    Ayres, T. R.

    2013-02-01

    ASTRAL is a project to create high-resolution, high-S/N UV (1150-3200 Å) atlases of bright stars utilizing {HST}/STIS. During Cycle 18 (2010-2011), eight cool star targets were observed, including key objects like Procyon and Betelgeuse, churning through 146 orbits in the process. The new spectral atlases are publically available through the project website. Data were obtained with the Hubble Space Telescope.

  14. 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)

  15. 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)

  16. 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

  17. Advances in rapid cooling treatment for heat stroke

    Directory of Open Access Journals (Sweden)

    Jia-jia ZHAO

    2014-10-01

    Full Text Available Heat stroke is a life-threatening disease characterized clinically by central nervous system dysfunction and severe hyperthermia (core temperature rises to higher than 40℃. The unchecked rise of body core temperature overwhelms intrinsic or extrinsic heat generation mechanism, thus overwhelms homoeostatic thermoregulation. Hyperthermia causes cellular and organ dysfunction with progressive exacerbation resulting in multi-organ failure and death. Rapid cooling to reduce core temperature as quickly as possible is the primary and most effective treatment, as it has been shown that the major determinant of outcome in heatstroke is the degree and duration of hyperthermia. If suppression of body temperature is delayed, the fatality rate will be elevated. Several cooling methods are available, including physical cooling by conduction, convection and evaporation with ice/cold water immersion, internal cooling by invasive methods such as hemofiltration, intravascular cooling, cold water gastric and rectal lavage, and cooling with drugs. It is crucial to formulate a scientific and reasonable strategy for the subsequent treatment in accordance with the patient's physical condition, the condition of cooling equipment, and the manipulator's proficiency in cooling methods and equipment usage. This article reviews the domestic and international advances in study of rapid and efficient cooling measures for heat stroke. DOI: 10.11855/j.issn.0577-7402.2014.10.17

  18. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Administrator

    of introducing innovative features towards further reduc- tion in unit energy cost and enhancing safety in these reactors. Clear strategies have been identified to simplify the design, reduce construction time, enhance the burnup and close the fuel cycle with minimum cooling and out- of-pile inventory, without sacrificing ...

  19. Note: A four-pass acousto-optic modulator system for laser cooling of sodium atoms.

    Science.gov (United States)

    Lu, Bo; Wang, Dajun

    2017-07-01

    We present a four-pass acousto-optic modulator (AOM) system for providing the repumping light for laser cooling of sodium atoms. With only one 400 MHz AOM, we achieve a tunable laser frequency shift around 1.6 GHz with total efficiency up to 30%. This setup provides an alternative over conventional methods to generate a sodium repumping light using more expensive high frequency AOMs or electro-optical modulators (EOMs) in the GHz domain. This compact and reliable setup can be easily adapted to other frequencies and may find applications in laser spectroscopy, laser cooling and trapping, and coherent manipulation of atomic quantum states.

  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. 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

  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. Corrosion and radioactivity in the primary circuit of a sodium cooled reactor. The computer code ''Corona''

    International Nuclear Information System (INIS)

    Costa, L.; Fremont, R. de; Mougniot, J.C.; Msika, D.

    1976-01-01

    In this paper the proble ms of corrosion and activity in the primary coolant circuit of a sodium cooled fast reactor are treated. The main features of Corona, a computer code which has been used to perform the calculations, are outlined

  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. Performance improvement of dry cooled advanced concentrating solar power plants using daytime radiative cooling

    International Nuclear Information System (INIS)

    Zeyghami, Mehdi; Khalili, Fardin

    2015-01-01

    Highlights: • Simple and recompression super critical carbon dioxide power cycles are analyzed. • Radiative cooling is investigated as the supplemental cooling for air cooled plants. • For simple super critical carbon dioxide cycle performance improvement is up to 5%. • For recompression super critical carbon dioxide cycle the improvement is up to 7.5%. - Abstract: In this study, utilization of daytime radiative cooling to enhance the performance of air-cooled concentrating solar thermal power plants is investigated. Water scarcity and environmental concerns are the driving forces for solar thermal power plants to use dry cooling systems. In order to overcome the energy conversion efficiency penalties associated with using air cooled technologies various supplemental cooling techniques have been proposed. Recent advancements in manufacturing structures with selective radiative properties have made the daytime radiative cooling to the cold outer space practical. In this work, the efficiency improvement of the air-cooled advanced supercritical carbon dioxide power cycles coupled with a radiative cooler is explored. It is shown that for the simple supercritical carbon dioxide cycle operating at hot source temperature equal 550 °C by employing 14.02 m 2 /kW e radiative cooler, it is possible to overcome the efficiency losses due to air cooling and the net output of the cycle improves by 5.0%. At hot source temperature equal 800 °C, the required radiative cooler area is 4.38 m 2 /kW e and respective performance improvement is equal 3.1%. For the recompression supercritical carbon dioxide cycle operating at hot source temperature equal 550 °C by employing 18.26 m 2 /kW e radiative cooler, it is possible to overcome the efficiency losses due to air cooling and the net output of the cycle improves by 7.5%. At hot source temperature equal 800 °C, the required radiative cooler area is 10.46 m 2 /kW e and respective performance improvement is equal 4.9%.

  6. Ferrittic steels sodium cooled fast reactor piping: an alternative to austenitic stainless steels

    International Nuclear Information System (INIS)

    Dubey, J.K.; Athmalingam, S.; Balasubramaniyan, V.; Srinivasan, G.

    2016-01-01

    Piping for Nuclear Steam Supply System (NSSS) in sodium cooled fast reactor constitutes a significant portion of the total plant cost. Optimal choice of piping material is therefore essential from the economy consideration. Material selection also plays an important role in reliable and safe operation of fast breeder reactor. The major factors considered in the selection of material include compatibility of material, operating conditions, availability of design data in nuclear codes, ease of fabrication, international experience, cost etc. Cost reduction is an important aspect for the future fast breeder reactor to be competitive. There are several components for which cheaper materials may satisfy the design requirements. Sodium piping in fast reactor is designed for low pressure and high temperature when compared to fossil power plant steam piping. Hence sodium piping is thin walled. Sodium piping has to be designed for normal, possible design basis events and transient load like seismic and sodium-water reaction pressure. This paper explores the various aspect of ferritic steel as alternative to austenitic stainless steel for piping of sodium cooled fast reactor

  7. 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.

  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. 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.

  10. 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

  11. 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.

  12. 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)

  13. 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.

  14. 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

  15. 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

  16. Steels for the primary circuits of the sodium cooled breeder reactors

    International Nuclear Information System (INIS)

    Weisz, M.

    1976-01-01

    The problems related to the utilization of austenitic stainless steels in the sodium cooled breeder reactors are discussed: consequences of the dispersion between different castings on the mechanical properties; effects of Na on the creep and fatigue behavior (particularly, the influence of carbon transfer); consequences of a long hold-time at high temperature on toughness, the intergranular sensibilization and the mechanical properties of the welded joints [fr

  17. 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

  18. French code system for a sodium cooled LMR inter-assembly thermal hydraulic analysis

    International Nuclear Information System (INIS)

    Kim, Young-Gyun; Lim, Hyun-Jin; Kim, Young-Il

    2005-03-01

    Sodium cooled LMR core is generally comprised of many ducted assemblies which have no flow exchanges between them. So, the required flow to each assembly corresponding to its power has to be allocated in thermal hydraulic design. Flow allocation facility, which is called orifice, is used for this purpose in an LMR core. In this context, flow grouping, assembly subchannel analysis and inter-assembly flow analysis have to be done in the LMR core thermal hydraulic design and analysis. This report describes this sodium cooled LMR core thermal hydraulic design procedure, in which are flow grouping, subchannel analysis and inter-assembly whole core analysis. And the French whole core analysis code system is described which is used for the domestic whole core thermal hydraulic analysis code system development. Firstly, sodium cooled LMR core thermal hydraulic conceptual design and analysis procedure is explained in chapter 2. Chapter 3 overviews the necessity and methodology of the whole core thermal hydraulic analysis, and the French whole core analysis system is described in chapter 4. Chapter 5 describes the domestic plan of the inter-assembly thermal hydraulic analysis system, and chapter 6 shows the conclusion and the future works

  19. French code system for a sodium cooled LMR inter-assembly thermal hydraulic analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young-Gyun; Lim, Hyun-Jin; Kim, Young-Il

    2005-03-01

    Sodium cooled LMR core is generally comprised of many ducted assemblies which have no flow exchanges between them. So, the required flow to each assembly corresponding to its power has to be allocated in thermal hydraulic design. Flow allocation facility, which is called orifice, is used for this purpose in an LMR core. In this context, flow grouping, assembly subchannel analysis and inter-assembly flow analysis have to be done in the LMR core thermal hydraulic design and analysis. This report describes this sodium cooled LMR core thermal hydraulic design procedure, in which are flow grouping, subchannel analysis and inter-assembly whole core analysis. And the French whole core analysis code system is described which is used for the domestic whole core thermal hydraulic analysis code system development. Firstly, sodium cooled LMR core thermal hydraulic conceptual design and analysis procedure is explained in chapter 2. Chapter 3 overviews the necessity and methodology of the whole core thermal hydraulic analysis, and the French whole core analysis system is described in chapter 4. Chapter 5 describes the domestic plan of the inter-assembly thermal hydraulic analysis system, and chapter 6 shows the conclusion and the future works.

  20. 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.

  1. 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)

  2. 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)

  3. Transient Response to Rapid Cooling of a Stainless Steel Sodium Heat Pipe

    Science.gov (United States)

    Mireles, Omar R.; Houts, Michael G.

    2011-01-01

    Compact fission power systems are under consideration for use in long duration space exploration missions. Power demands on the order of 500 W, to 5 kW, will be required for up to 15 years of continuous service. One such small reactor design consists of a fast spectrum reactor cooled with an array of in-core alkali metal heat pipes coupled to thermoelectric or Stirling power conversion systems. Heat pipes advantageous attributes include a simplistic design, lack of moving parts, and well understood behavior. Concerns over reactor transients induced by heat pipe instability as a function of extreme thermal transients require experimental investigations. One particular concern is rapid cooling of the heat pipe condenser that would propagate to cool the evaporator. Rapid cooling of the reactor core beyond acceptable design limits could possibly induce unintended reactor control issues. This paper discusses a series of experimental demonstrations where a heat pipe operating at near prototypic conditions experienced rapid cooling of the condenser. The condenser section of a stainless steel sodium heat pipe was enclosed within a heat exchanger. The heat pipe - heat exchanger assembly was housed within a vacuum chamber held at a pressure of 50 Torr of helium. The heat pipe was brought to steady state operating conditions using graphite resistance heaters then cooled by a high flow of gaseous nitrogen through the heat exchanger. Subsequent thermal transient behavior was characterized by performing an energy balance using temperature, pressure and flow rate data obtained throughout the tests. Results indicate the degree of temperature change that results from a rapid cooling scenario will not significantly influence thermal stability of an operating heat pipe, even under extreme condenser cooling conditions.

  4. Thermal hydraulic feasibility of supercritical carbon dioxide Brayton cycle power conversion for the KALIMER-150 sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Moisseytsev, A.; Cho, D.H.; Kim, Seong-O; Hahn, Dohee

    2004-01-01

    One possible approach to achieving a significant reduction in the overnight and operating costs of a sodium-cooled fast reactor is to replace the traditional Rankine steam cycle with an advanced power converter that consists of a gas turbine Brayton cycle that utilizes supercritical carbon dioxide (S-CO 2 ) as the working fluid. A joint project between Argonne National Laboratory and the Korea Atomic Energy Research Institute has been initiated to investigate the thermal-hydraulic feasibility of coupling the S-CO 2 Brayton cycle to the KALIMAR-150 sodium-cooled fast reactor conceptual design. As an initial step in investigating the system aspects of coupling the reactor to the S-CO 2 Brayton cycle, the case is investigated in which the intermediate heat transfer loop is eliminated in order to achieve additional cost reductions. The main objectives are to determine the potential gain in plant efficiency and to estimate the size of the key Brayton cycle components. A S-CO 2 Brayton cycle efficiency of 43.2% is calculated. Accounting for primary pump power and other in-house loads, a net plant efficiency of 40.8% is obtained, compared to 38.2% for the current (Rankine cycle) plant. If higher Na temperatures could be accommodated, then a 1% gain in plant efficiency could be obtained for each 20degC incremental increase in sodium core outlet temperature. Further investigation of the thermal sizing of the Na/S-Co 2 heat exchanger is also carried out; parametric sensitivity studies are performed for the case in which the intermediate heat transport system is retained as well as the case in which it is eliminated. (author)

  5. Comparison of advanced cooling technologies efficiency depending on outside temperature

    Energy Technology Data Exchange (ETDEWEB)

    Blaise Hamanaka; Haihua Zhao; Phil Sharpe

    2009-09-01

    In some areas, water availability is a serious problem during the summer and could disrupt the normal operation of thermal power plants which needs large amount of water to operate. Moreover, when water quantities are sufficient, there can still be problem created by the waste heat rejected into the water which is regulated in order to limit the impact of thermal pollution on the environment. All these factors can lead to a decrease of electricity production during the summer and during peak hours, when electricity is the most needed. In order to deal with these problems, advanced cooling technologies have been developed and implemented to reduce water consumption and withdrawals but with an effect in the plant efficiency. This report aims at analyzing the efficiency of several cooling technologies with a fixed power plant design and so to produce a reference to be able to compare them.

  6. Investigations of decay heat removal by natural convection with boiling in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Kaiser, A.; Peppler, W.; Strake, M.

    1979-03-01

    The safety analysis of a LMFBR indicates the requirement of safely removing the decay heat produced after a reactor shut-down, especially in the case of a failure of all primary circuits. To investigate the conditions under which power in the range of the decay heat can be transfered from a pin bundle to a sodium loop by natural convection, a series of experiments was carried out. Special attention was paid to the behaviour of the natural convection system when boiling occurs, and also to the limits of cooling capability. To apply the experimental results a computer program was made using a simplified model of the emergency cooling system of the SNR 300. With this program several cases of emergency cooling under the boundary conditions of in-tank natural convection were analyzed, assuming a breach of a primary circuit. As an example, the consequences of an increase of the flow resistances in a subassembly were investigated. It was demonstrated that under conditions of steady state boiling there will be only very low vapour qualities. Similar results were obtained from investigations when the sodium temperature at the inlet to the core was elevated, and when the flow resistances in the cold leg of the natural convection loop were increased by a factor of two. Further experiments gave evidence that the cooling of the bundle will substantially be maintained under conditions of low vapour qualities. In summary, it may be stated that even under very pessimistic assumptions concerning the progress of the in-tank natural circulation, the cooling will be maintained reliably, even if boiling occurs for some time. (orig.) [de

  7. 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

  8. 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)

  9. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Yun, Sunghwan; Kim, Sang Ji

    2014-01-01

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a 23 Na(n,g) 24 Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B 4 C shielding inside the subassembly

  10. 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

  11. The reviews of the FCI under CDAs in sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Zhang Zhigang; Liu Xingchao; Yang Zhi

    2017-01-01

    The molten fuel-coolant interaction (FCI) under core disruptive accidents (CDAs) in sodium fast breeder reactor (SFR) is a crucial problem on core safety study internationally. It involves multiphase, multicomponent, deformation and solidification, complex heat and mass transfer problems, which is a crux to assess the post-accident heat removal (PAHR) and core re-cooling ability. The latest experimental and numerical research status of FCI was introduced, especially the molten metal fragmentation mechanism. The sodium entrainment physical model, fragmentation induced by the thermal stress and fragmentation induced by solidification are the major breakthroughs in studying the fragmentation mechanism. Modeling of FCI process using MPS and ISPH numerical methods has made some progression. At the same time, the author summarized the work and the existing problems and made a general outlook about the research directions and trends in the future. (author)

  12. 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.

  13. 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.

  14. Development and performance of fuel elements for sodium-cooled breeder reactors in Germany

    International Nuclear Information System (INIS)

    Mayer, H.; Hoechel, J.

    1980-01-01

    The first sodium-cooled reactor commissioned in Germany, KNK, serves now as test facility for plutonium bearing oxide fuel elements. The target is to provide reliable fuel for the SNR-300 project (Kalkar Nuclear Power Plant). The long-range target is fuel for burnups above 100,000 MW d/t, which moreover can easily be fabricated and reprocessed. As in the U.K., the line of grid-spaced bundles is favorised, being promising as regards the possibility of replacement of a defected pin and reinsertion of the bundle. (orig.) [de

  15. 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

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

    OpenAIRE

    Chan Bock Lee; Jin Sik Cheon; Sung Ho Kim; Jeong-Yong Park; Hyung-Kook Joo

    2016-01-01

    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 pyroelectrochem...

  17. 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.

  18. 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

  19. 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.

  20. 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

  1. 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

  2. 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

  3. 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.

  4. 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.

  5. 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)

  6. 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

  7. 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.

  8. Water cooled metal optics for the Advanced Light Source

    International Nuclear Information System (INIS)

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

    1992-01-01

    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 midperiod error. We will briefly describe our improved version of the long trace profiler, and its importance to our metrology program. We have completely redesigned the mechanical, optical and computational parts of the profiler system with the cooperation of P. 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. This 15 in. long Glidcop T M mirror is coated with electroless nickel from Acteron Corporation in Redwood City, CA. 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. (orig.)

  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. 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

  11. 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

  12. 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...

  13. 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)

  14. 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

  15. Formation of metastabil liquid phases in the isotonic solution of sodium cloraide during cooling

    Directory of Open Access Journals (Sweden)

    A. T. Ходько

    2016-07-01

    Full Text Available In this paper the cooling process cryomicroscopy of 0.15 M of isotonic sodium chloride solution was conducted. It was shown that there is liquid – liquid phase change before the crystallization process. As a result, the coarse system (highly concentrated emulsion was formed. The dispersed phase and the disperse medium in a binary system with the same qualitative chemical composition differ in concentration. Therefore, the greater is the volume ratio of the coexisting phases, the greater is the difference in their quantitative compositions. The dispersed phase, that composes the main volume in the system under investigation, should have lower NaCl concentration than the disperse medium and the initial solution. In this case it will be hypotonic (and disperse medium – hypertonic in relation to cytoplasm of human internal environment. This physical-chemical factor, which hasn’t been considered previously, might be responsible for osmotic damage in living cells during cryopreservation of cell suspensions.

  16. Sodium cooled fast reactors being built or planned in the world

    International Nuclear Information System (INIS)

    Martin, L.

    2014-01-01

    This article reviews the status of sodium-cooled fast reactor programs throughout the world. For each country: Russia, India, Japan, Republic of Korea and China the national framework is recalled as well as the purposes of each fast reactor program. Main technological features are described and changes with current operating fast reactors are highlighted. The following programs are described: the Russian program involving BN 800, BN 1200 and MBIR reactors, the Indian program including PFBR and FBR reactors, the Japanese JSFR reactor, the Korean PGSFR reactor, the Chinese program involving CEFR and CFR 600 reactors. Concerning SMR (Small Modular Reactor), reactors whose power output is below 300 MWe, the USA and Japan are the most active countries, only the Japanese 4S reactor and the international SMFR program are described, the PRISM reactor and the 'Traveling Wave Reactor' are briefly quoted in the article. (A.C.)

  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. 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

  20. 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.

  1. Numerical approach of self-wastage phenomena in steam generator of sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Ohnishi, Yuki; Takata, Takashi; Yamaguchi, Akira; Uchibori, Akihiro; Kikuchi, Shin; Kurihara, Akikazu; Ohshima, Hiroyuki

    2012-01-01

    In the steam generator of sodium-cooled fast reactor (SFR), self-wastage phenomenon is a crack enlargement on the heat transfer tube itself caused by sodium-water reaction (SWR), which is triggered by the leakage of steam/water from the initial micro-crack. Therefore, a quantification of the self-wastage phenomenon is of importance from the viewpoint of safety assessment in the steam generator. In this study, we propose a numerical approach to evaluate the self-wastage phenomena and investigate an enlargement of the crack using a multi-dimensional-SWR code 'SERAPHIM'. In the analysis, two-dimensional initial crack is assumed based on SWAT-4 experiment carried out by Japan Atomic Energy Agency (JAEA). The wastage rate was estimated by Arrhenius type of the hypothetical equation, and remeshing arrangement was performed by changing solid-cells to fluid-cells with the estimated wastage amount on the heat exchanger tube in the initial (or former) model. After simulated again using the remeshing models, the resulting SWR products were distributed not only circumferential direction but also radial direction. The wastage region was formed invert triangle shape as the similar with experimental observation. (author)

  2. 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

  3. 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

  4. 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

  5. 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)

  6. 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)

  7. A prospective scenario of the French nuclear fleet growth based on sodium cooled fast reactor technology

    International Nuclear Information System (INIS)

    Garzenne, Claude; Le Mer, Joel; Lemasson, David; Hoang, Manh-Hung

    2011-01-01

    Generation IV Sodium cooled Fast Reactors (SFR) deployment would allow to optimize the use of the various available resources (natural, reprocessed and depleted uranium, plutonium) thanks to breeding capacities featuring a valuable advantage with respect to the fuel cycle flexibility and fissile material management. The complete replacement of the 60 GWe French nuclear fleet by GEN IV SFRs in 2100 would require around 1000 tons of plutonium. An accurate simulation of this prospective scenario shows that the amount of plutonium issued from the French PWRs spent fuel reprocessing would not be enough. The lacking amount of plutonium could be produced with fertile blankets during the transient SFR deployment phase. A more ambitious research scenario, aimed at doubling the nuclear French fleet installed power in 2100, would require to use SFRs at their maximum breeding capacity. However, it is not possible to deploy more than about 100 GWe of SFRs in 2100, meaning that the fleet growth would have to be partially supported by GEN III PWRs. Using the scenario simulation code TIRELIRE-STRATEGIE, we have optimized the main scenario parameters: the capacities of the fuel cycle facilities, the proportion of PWRs necessary for supporting the growth phase, the kinetics of SFRs deployment compatible with the plutonium build-up, etc., while respecting industrial constraints such as a realistic cooling time before reprocessing, a fuel cycle plants utilization rate constant over several decades, etc.. We illustrate the impact of this French fleet growth scenario over the nuclear material fluxes in the fuel cycle plants, the uranium consumption, and the waste production. (author)

  8. 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

  9. Advances in open-cycle solid desiccant cooling

    Energy Technology Data Exchange (ETDEWEB)

    Penney, T R; Maclaine-cross, I

    1985-05-01

    Of the solar cooling options available open cycle solid desiccant cooling looks very promising. A brief review of the experimental and analytical efforts to date shows that within the last 10 years thermal performance has doubled. Research centers have been developed to explore new materials and geometry options and to improve and validate mathematical models that can be used by design engineers to develop new product lines. Typical results from the Solar Energy Research Institute's (SERI) Desiccant Cooling Research Program are shown. Innovative ideas for new cycles and spinoff benefits provide incentives to continue research in this promising field.

  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. Safety Design and Evaluation in a Large-Scale Japan Sodium-Cooled Fast Reactor

    Directory of Open Access Journals (Sweden)

    H. Yamano

    2012-01-01

    Full Text Available As a next-generation plant, a large-scale Japan sodium-cooled fast reactor (JSFR adopts a number of innovative technologies in order to achieve economic competitiveness, enhanced reliability, and safety. This paper describes safety requirements for JSFR conformed to the defense-in-depth principle in IAEA. Specific design features of JSFR are a passive reactor shutdown system and a recriticality-free concept against anticipated transients without scram (ATWS in design extension conditions (DECs. A fully passive decay heat removal system with natural circulation is also introduced for design-basis events (DBEs and DECs. In this paper, the safety design accommodation in JSFR was validated by safety analyses for representative DBEs: primary pump seizure and long-term loss-of-offsite power accidents. The safety analysis also showed the effectiveness of the passive shutdown system against a typical ATWS. Severe accident analysis supported by safety experiments and phenomenological consideration led to the feasibility of in-vessel retention without energetic recriticality. Moreover, a probabilistic safety assessment indicated to satisfy the risk target.

  12. 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

  13. 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.

  14. 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)

  15. 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

  16. 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.

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

    Directory of Open Access Journals (Sweden)

    Blanchet David

    2017-01-01

    Full Text Available 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.

  18. 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.

  19. 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

  20. 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

  1. 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

  2. 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…

  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. 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

  5. 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.

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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 ...

  11. Sodium Advanced Fast Reactor (SAFR). Volume II. Summary plant description

    Energy Technology Data Exchange (ETDEWEB)

    None

    1985-09-01

    The balance of plant (BOP) includes all buildings and structures, all steam and water facilities, all power generation and transmission systems, and all auxiliary systems which do not contain or handle sodium. In addition, the BOP includes all site facilities such as roads, fences, etc.

  12. Advanced wet--dry cooling tower concept performance prediction

    Energy Technology Data Exchange (ETDEWEB)

    Snyder, T.; Bentley, J.; Giebler, M.; Glicksman, L.R.; Rohsenow, W.M.

    1977-01-01

    The purpose of this year's work has been to test and analyze the new dry cooling tower surface previously developed. The model heat transfer test apparatus built last year has been instrumented for temperature, humidity and flow measurement and performance has been measured under a variety of operating conditions. Tower Tests showed approximately 40 to 50% of the total energy transfer as taking place due to evaporation. This can be compared to approximately 80 to 85% for a conventional wet cooling tower. Comparison of the model tower test results with those of a computer simulation has demonstrated the validity of that simulation and its use as a design tool. Computer predictions have been made for a full-size tower system operating at several locations. Experience with this counterflow model tower has suggested that several design problems may be avoided by blowing the cooling air horizontally through the packing section. This crossflow concept was built from the previous counterflow apparatus and included the design and fabrication of new packing plates. Instrumentation and testing of the counterflow model produced data with an average experimental error of 10%. These results were compared to the predictions of a computer model written for the crossflow configuration. In 14 test runs the predicted total heat transfer differed from the measured total heat transfer by no more than 8% with most runs coming well within 5%. With the computer analogy's validity established, it may now be used to help predict the performance of fullscale wet-dry towers.

  13. 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

  14. 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

  15. Potential use of dry cooling in support of advanced energy generation systems

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, D.W.; Arnold, E.M.; Allemann, R.T.

    1979-09-01

    Advanced energy technologies were investigated for filling the energy supply and demand gap, including fuel cells, thermionic converters, and fusion. Technologies that have the potential for supplying energy in the future are solar, geothermal, coal gasification and liquefaction, clean solid fuel from coal, and oil shale. Results are presented of an analysis of the advanced energy generation systems, the potential for using dry cooling, and the waste heat generation characteristics of the advanced technologies. The magnitude of the waste heat expected to be generated indicates the following percentages of total cooling requirements would be needed by advanced energy technologies: (a) 1% to 2% in 1985, (b) 17% to 40% in 2000, and (c) 24% to 76% in 2025. Dry cooling could be required for flashed steam and dry steam geothermal plants if balancing withdrawal and reinjection of the geothermal fluid becomes a requirement. Binary cycle geothermal plants and plants using the hot dry rocks geothermmal resource are even more likely to require dry cooling since these plants will need an outside source of water. Solar central tower plants have a high potential for the use of dry cooling since they are likely to be located in the Southwest where water availability problems are already apparent. The high water consumption associated with the projected synthetic fuel production levels indicates that dry cooling will be desirable, perhaps even mandatory, to achieve a high level of synthetic fuel production. In the year 2000, between 2.5 and 13 GW of electrical energy produced by advanced power generation systems may require dry cooling. In the year 2025, this requirement may increase to between 4.5 and 81 GW/sub e/.

  16. 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

  17. 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

  18. 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)

  19. A comparison of core perturbation by coolant loss between sodium and lead-bismuth cooled reactor

    International Nuclear Information System (INIS)

    Kim, Yong Nam; Kim, Jong Kyung; Park, Won Seok

    2003-01-01

    This study performs a comparative analysis of the core perturbation caused by coolant loss between sodium and lead-bismuth eutectic. Considering the Zr-based and the U-based fuel in a 1,000MWth class reactor for TRU incineration, we investigate which coolant shows better performance for negative coolant loss reactivity in each case of fuel type. The calculation results show that in the case of U-based fuel, sodium gives rise to more positive coolant loss reactivity than lead-bismuth. However, when the Zr-based (U-free) fuel is considered, sodium offers negative coolant loss reactivity, whereas lead-bismuth makes the coolant loss reactivity positive. It is recommended to employ sodium coolant for the fertile-free fueled core and lead-bismuth for the core with fertile nuclides

  20. Advanced materials for magnetic cooling: Fundamentals and practical aspects

    Science.gov (United States)

    Balli, M.; Jandl, S.; Fournier, P.; Kedous-Lebouc, A.

    2017-06-01

    Over the last two decades, the research activities on magnetocalorics have been exponentially increased, leading to the discovery of a wide category of materials including intermetallics and oxides. Even though the reported materials were found to show excellent magnetocaloric properties on a laboratory scale, only a restricted family among them could be upscaled toward industrial levels and implemented as refrigerants in magnetic cooling devices. On the other hand, in the most of the reported reviews, the magnetocaloric materials are usually discussed in terms of their adiabatic temperature and entropy changes (ΔTad and ΔS), which is not enough to get more insight about their large scale applicability. In this review, not only the fundamental properties of the recently reported magnetocaloric materials but also their thermodynamic performance in functional devices are discussed. The reviewed families particularly include Gd1-xRx alloys, LaFe13-xSix, MnFeP1-xAsx, and R1-xAxMnO3 (R = lanthanide and A = divalent alkaline earth)-based compounds. Other relevant practical aspects such as mechanical stability, synthesis, and corrosion issues are discussed. In addition, the intrinsic and extrinsic parameters that play a crucial role in the control of magnetic and magnetocaloric properties are regarded. In order to reproduce the needed magnetocaloric parameters, some practical models are proposed. Finally, the concepts of the rotating magnetocaloric effect and multilayered magnetocalorics are introduced.

  1. 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

  2. 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

  3. 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)

  4. 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)

  5. 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.)

  6. Thermal and deformation analyses of a novel cryogenically cooled monochromator for an Advanced Photon Source beamline

    International Nuclear Information System (INIS)

    Wang, Zhibi; Yun, Wenbing; Kuzay, T.M.; Knapp, G.

    1994-01-01

    The analytical results and design considerations for a novel cryogenically cooled Advanced Photon Source (APS) monochromator are presented. Because the monochromator uses silicon crystal, cryogenic cooling enables one to take advantage of the high conductivity and low thermal expansion coefficient of silicon at cryogenic temperatures. The APS monochromator features a machined slot with variable thickness below the surface. With this configuration, only a fraction of the total undulator power is absorbed by the crystal; the remaining power is transmitted through the crystal and is absorbed b a second element that can be cooled by standard cooling techniques. A variety of analyses has been performed with different parameters and configurations to maximize the performance of the monochromator and minimize the total absorbed power by the crystal

  7. 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)

  8. 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

  9. 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)

  10. 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.

  11. 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)

  12. 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.

  13. 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.

  14. The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium cooled fast reactor

    International Nuclear Information System (INIS)

    Korkmaz, Mehmet E.; Agar, Osman

    2014-01-01

    In this research, we investigated the burnup characteristics and the conversion of fertile 232 Th into fissile 233 U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning 232 Th fuel (fuel pin 1) and 233 U 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.

  15. The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman [Karamanoglu Mehmetbey University, Faculty of Kamil Oezdag Science, Karaman (Turkmenistan)

    2014-06-15

    In this research, we investigated the burnup characteristics and the conversion of fertile {sup 232}Th into fissile {sup 233}U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning {sup 232}Th fuel (fuel pin 1) and {sup 233}U 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.

  16. 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.

  17. 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

  18. 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.)

  19. 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)

  20. 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)

  1. A review of inherent safety characteristics of metal alloy sodium-cooled fast reactor fuel against postulated accidents

    Directory of Open Access Journals (Sweden)

    Tanju Sofu

    2015-04-01

    Full Text Available The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, double-fault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel–coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.

  2. Preliminary design study of a board type radial fuel shuffling sodium cooled breed and burn reactor core

    International Nuclear Information System (INIS)

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

    2014-01-01

    Highlights: • A 1500MWt radial fuel shuffling sodium cooled breed and burn reactor core was designed. • The board type radial fuel shuffling strategy was applied and demonstrated. • Influences of the fuel height and core radius were investigated. - Abstract: In this paper, a preliminary board type radial fuel shuffling sodium cooled breed and burn reactor core is designed. In the current design, a number of breeding subassemblies are arranged in the center core to ensure enough breeding. A self-developed MCNP-ORIGEN coupled system with the ENDF/B-VI data library is applied to perform neutronics and burn-up calculations. For a 2.0 m radius and 2.5 m height core, the results demonstrate the feasibility of the board type radial fuel shuffling strategy. Breeding mainly occurs in the breeding subassemblies during the first 6 fuel cycles as they are moved to the burning/breeding region. The core will become asymptotically stable after about 24 years. The discharged burn-up of most subassemblies is about 15.0–30.0%. The influences of the core size on the major core parameters, such as initial k eff , steady k eff , maximum power density, peak burn-up and burn-up ratio between breeding and ignition subassemblies are calculated and investigated. The results indicate that the initial k eff increases with fuel height and core radius and finally reaches stability; the steady k eff increases with fuel height and core radius, then reaches peak value and finally decreases; the maximum power density, the peak burn-up and the burn-up ratio between breeding and ignition subassemblies decrease with the increase of fuel height and core radius; if core radius is less than 1.875 m, they increase sharply with the decrease of core radius

  3. 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

  4. Study and evaluation of innovative fuel handling systems for sodium-cooled fast reactors: fuel handling route optimization

    International Nuclear Information System (INIS)

    Dechelette, Franck; Morin, Franck; Laffont, Guy; Rodriguez, Gilles; Sanseigne, Emmanuel; Christin, Sebastien; Mognot, Xavier; Morcillo, Aurelien

    2014-01-01

    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 mutualization of one FHS coupled with two reactors. These three concepts are then inter-compared 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. (authors)

  5. Final Report THIBO-Experiments: Thermal-hydraulically excited fuel pin oscillations in sodium cooled fast reactors

    International Nuclear Information System (INIS)

    Bojarsky, E.; Deckers, H.; Lehning, H.; Piel, D.; Reiser, H.; Schmidt, L.

    1990-09-01

    The KNK II reactor in Karlsruhe experienced fuel element damages which could not be traced back to hydraulically excited vibrations. Instead, some indications pointed to low-frequency fuel rod oscillations caused by temperature differences over the circumference of the fuel rod as a result of the high specific rod power and the clearance of fuel rods in their spacers. In 1988, specific experiments were started in the sodium loop of the IMF III to investigate this phenomenon (THIBO experiments, THIBO standing for Thermal Induced pin Bowing). The rod movements were made visible and detected in a reproducible way. In 1989, another series of tests (THIBO II) have been run in a second test section. In this case, the cooling channel area was reduced so much that the thermal-hydraulic conditions approximated very closely those existing in the KNK II reactor. The experiments have shown that the fuel rods may start moving already at relatively low sodium temperature increases and low partial loads, respectively, even if the rod clearance in the spacer was set to realistically low levels [de

  6. 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

  7. The sodium cooled small sealed fast reactor (4S) with non-refueling

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Ueda, Nobuyuki; Kinoshita, Izumi; Nishimura, Satoshi; Minato, Akio

    2004-01-01

    CRIEPI has been developing the 4S reactor (Super Safe, Small and Simple reactor) for application to dispersed energy supply and multipurpose use. Electrical output of the 4S reactor is from 10 MW to 50 MW, and the core lifetime without refueling is from 10 to 30 years. 30 year core lifetime can be achieved with the 10 MWe 4S (4S-10M) reactor. All temperature feedback reactivity coefficients, including coolant void reactivity, of the 4S-10M are negative during the 30 year lifetime. The pressure loss of the reactor core is lower than 2 kg/cm 2 to enable effective utilization of the natural circulation force. To suppress the influence of the scale disadvantage, loop-type reactor design is proposed as the candidates for the 4S-10M. The size of the reactor vessel is miniaturized by adopting the loop type as a nuclear system, and 2.5 m in diameter and 14 m in height have been achieved (4S-10ML). An integrated equipment which includes primary and secondary electromagnetic pumps (EMPs), an intermediated heat exchanger (IHX) and a steam generator (SG) is proposed and is collocated by the reactor vessel. The decay heat removal systems of 4S-10ML consist of the reactor vessel air cooling system (RVACS) and SGACS (a similar system to the RVACS, air cooling of the outside of the integrated equipment vessel). They are completely passive systems. A step mat structure and the horizontal aseismatic structure are adopted to reduce the construction cost of the reactor building. 4S-10ML has unique features in the cooling systems such as integrated equipment and two separate passive decay heat removal systems which operate at the same time. To evaluate the design feasibility, the transition analyses were executed by the CERES code. The design concept of the 4S-10ML and the results of the plant transition analyses are described in this report. (author)

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

    International Nuclear Information System (INIS)

    Sinha, N.K.; Raj, Baldev

    2014-01-01

    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 −3 /10 −2 scc/s/m length of seal) and argon inflation gas (10 −3 scc/s/m length of seal) is described

  9. 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

  10. Experience with the commissioning of helically coiled advanced gas cooled reactor boilers

    International Nuclear Information System (INIS)

    Kettle, D.B.

    1984-01-01

    The paper describes aspects of the experience gained during commissioning of the helically coiled pod boilers for an advanced gas-cooled reactor. The boiler geometry is shown to be a factor contributing to gas-side and water-side convection phenomena encountered during commissioning. Detailed information on thermal performance and vibrational response was obtained from commissioning tests on specially instrumented boiler units. (author)

  11. 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 ...

  12. Passive Acoustic Leak Detection for Sodium Cooled Fast Reactors Using Hidden Markov Models

    Science.gov (United States)

    Marklund, A. Riber; Kishore, S.; Prakash, V.; Rajan, K. K.; Michel, F.

    2016-06-01

    Acoustic leak detection for steam generators of sodium fast reactors have been an active research topic since the early 1970s 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.

  13. 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)

  14. Clad failures detection on sodium-cooled fast reactors by high count rate gamma spectroscopy

    International Nuclear Information System (INIS)

    Rohee, Emmanuel

    2016-01-01

    The question of clad failures detection, through the monitoring of potentially released fission products in the primary sodium, is of utmost importance for SFR type generation IV reactors, and in particular for the future ASTRID industrial demonstrator. This thesis aims to propose ways to significantly improve the nuclear instrumentation dedicated to clad failures detection. A first study regarding the signal-to-noise ratio of the existing instrumentation, which is based on Delayed Neutron Detection (DND), is proposed. The experience feedback, as well as the Monte Carlo simulations, shows a disturbance of the useful neutron signal by photoneutrons produced in the polyethylene. A material replacement solution based on the use of graphite is proposed, and its performances are evaluated by Monte Carlo simulations and experiments. The second lead followed, technologically more ambitious, consists in enriching the existing instrumentation by introducing an innovative method, based on high count rate gamma spectroscopy on the primary sodium. In addition to bringing redundancy to the current instrumentation, such a spectroscopy would permit to enrich the clad failure diagnosis with an earlier and more complete measure (broader range of accessible fission products). First, a feasibility study is carried out using Monte Carlo simulations, by modeling a measurement station and simulating the HPGe detector response. Minimal detectable activities are calculated for the fission products of interest, and compared with results from simulations of release following a clad failure scenario. In a second stage, high count rate gamma spectroscopy using the ADONIS spectrometry system developed by CEA LIST was tested and validated in a real environment, by installing a measuring station on the ISABELLE 1 fuel pin irradiation loop at the OSIRIS reactor. Several irradiations have been monitored in this thesis, two of which allowed to follow a clad failure process using this approach

  15. Numerical investigation on heat transfer in an advanced new leading edge impingement cooling configuration

    Directory of Open Access Journals (Sweden)

    G. Lin

    2015-12-01

    Full Text Available It is known that the leading edge has the most critical heat transfer area of a gas turbine blade. The highest heat transfer rates on the airfoil can always be found on the stagnation region of the leading edge. In order to further improve the gas turbine thermal efficiency the development of more advanced internal cooling configurations at leading edge is very necessary. As the state of the art leading edge cooling configuration a concave channel with multi inline jets has been widely used in most of the blades. However, this kind of configuration also generates strong spent flow, which shifts the impingement off the stagnation point and weakens the impingement heat transfer. In order to solve this problem a new internal cooling configuration using double swirl chambers in gas turbine leading edge has been developed and introduced in this paper. The double swirl chambers cooling (DSC technology is introduced by the authors and contributes a significant enhancement of heat transfer due to the generation of two anti-rotated swirls. In DSC-cooling, the reattachment of the swirl flows always occurs in the middle of the chamber, which results in a linear impingement effect. Compared with the reference standard impingement cooling configuration this new cooling system provides a much more uniform heat transfer distribution in the chamber axial direction and also provides a much higher heat transfer rate. In this study, the influences of different geometrical parameters e.g. merging ratio of two cylinder channels, the jet inlet hole configurations and radius of blunt protuberances in DSC have been investigated numerically. The results show that in the DSC cooling system the jet inlet hole configurations have large influences on the thermal performance. The rectangular inlet holes, especially those with higher aspect ratios, show much better heat transfer enhancement than the round inlet holes. However, as the price for it the total pressure drop is

  16. 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)

  17. 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)

  18. 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

  19. 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.

  20. 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)

  1. Study on Structural Concept Design Improvement for a Gen-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Han; Joo, Young Sang; Lee, Hyeong Yeon; Kim, Jong Bum; Kim, Seok Hoon; Park, Chang Kyu; Koo, Gyeong Hoi

    2010-02-15

    An economic improvement is a hot issue as one of the Gen-IV nuclear plant goals. To secure economic competitiveness of a SFR compared to a pressurized water reactor, several structural design concepts are adapted in without loosing the reactor safety level. One is the increase of the plant capacity with the minimum number of component and loop, which leads the reduction of the plant maintenance, repair, and construction costs by a large-size scale effect. Another is the simple system arrangement, compact reactor size for only two loop system for a 1200MWe capacity of a pool type SFR, and the minimization of IHTS piping length through the properly locating the SG and secondary pump. Several researches are also studied to attain the economic improvement target of the NSSS in structural point of view; for example, an integrated concept of a refueling machine and inspection device with a long waveguide sensor for reactor internals, a high temperature LBB (leak before break) technology application, which can minimize the large protection facility against a large sodium leak like as a guillotine pipe break

  2. Technology of Fabrication for Sodium-cooled Fast Reactor Metallic Fuel

    International Nuclear Information System (INIS)

    Oh, S. J.; Kim, K. H.; Lee, C. T.; Ryu, H. J.; Ko, Y. M.; Woo, W. M.; Jang, S. J.; Lee, Y. S.; Lee, C. B.

    2008-02-01

    The fabrication process of metallic fuel for SFR(sodium fast reactor) of Generation-IV candidate reactors is composed of the fabrication of fuel pin, fuel rod, and fuel assembly. The key technology of the fabrication process for SFR can be referred to the fabrication technology of fuel pin. As SFR fuel contains MA(minor actinide) elements proceeding the recycling of actinide elements, it is so important to extinguish MA during irradiation in SFR, included in nuclear fuel through collection of volatile MA elements during fabrication of fuel pin. Hence, it is an imminent circumstance to develop the fabrication process of fuel pin. This report is an state-of art report related to the characteristics of irradiation performance for U-Zr- Pu metallic fuel, and the apparatus and the technology of conventional injection casting process. In addition, to overcome the drawbacks of the conventional injection casting and the U-Zr-Pu fuel, new fabrication technologies such as the gravity casting process, the casting of fuel pin to metal-barrier mold, the fabrication of particulate metallic fuel utilizing centrifugal atomization is surveyed and summarized. The development of new U-10Mo-X metallic fuel as nuclear fuel having a single phase in the temperature range between 550 and 950 .deg. C, reducing the re-distribution of the fuel elements and improving the compatibility between fuel and cladding, is also surveyed and summarized

  3. 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.

  4. 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

  5. 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

  6. A fundamental approach to specify thermal and pressure loadings on containment buildings of sodium cooled fast reactors during a core disruptive accident

    International Nuclear Information System (INIS)

    Velusamy, K.; Chellapandi, P.; Satpathy, K.; Verma, Neeraj; Raviprasan, G.R.; Rajendrakumar, M.; Chetal, S.C.

    2011-01-01

    Highlights: → An approach to quantify thermal and pressure loadings on RCB is presented. → Scaling laws to determine sodium release from water experiments are proposed. → Potential of in-vessel sodium fire after a CDA is assessed. → The proposed approach is applied to Indian Prototype Fast Breeder Reactor. - Abstract: Reactor Containment Building (RCB) is the ultimate barrier to the environment against activity release in any nuclear power plant. It has to be designed to withstand both positive and negative pressures that are credible. Core Disruptive Accident (CDA) is an important event that specifies the design basis for RCB in sodium cooled fast reactors. In this paper, a fundamental approach towards quantification of thermal and pressure loadings on RCB during a CDA, has been described. Mathematical models have been derived from fundamental conservation principles towards determination of sodium release during a CDA, subsequent sodium fire inside RCB, building up of positive and negative pressures inside RCB, potential of in-vessel sodium fire due to failed seals and temperature evolution in RCB walls during extended period of containment isolation. Various heating sources for RCB air and RCB wall and their potential have been identified. Scaling laws for conducting CDA experiments in small-scale water models by chemical explosives and the rule for extrapolation of water leak to quantify sodium leak in reactor are proposed. Validation of the proposed models and experimental simulation rules has been demonstrated by applying them to Indian prototype fast breeder reactor. Finally, it is demonstrated that in-vessel sodium fire potential is very weak and no special containment cooling system is essential.

  7. 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

  8. 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

  9. 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.

  10. 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)

  11. 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.

  12. 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

  13. 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

  14. 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)

  15. 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.

  16. 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.

  17. 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.

  18. 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)

  19. 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)

  20. 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.

  1. 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.

  2. 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

  3. SPARC fast reactor design : Design of two passively safe metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control

    OpenAIRE

    Lindström, Tobias

    2015-01-01

    In this master thesis a small modular sodium-cooled metal-fuelled pool-type fast reactor design, called SPARC - Safe and Passive with Autonomous Reactivity control, has been designed. The long term reactivity changes in the SPARC are managed by implementation of the the Autonomous Reactivity Control (ARC) system, which is the novelty of the design. The overall design is mainly based on the Integral Fast Reactor project (IFR), which experimentally demonstrated the passive safety characteristic...

  4. 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.

  5. 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

  6. 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.

  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. 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)

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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

  16. Annular flow induced vibration associated with on-load refuelling of advanced gas cooled reactors

    International Nuclear Information System (INIS)

    Fox, M.J.H.; Hodson, D.E.; Parkin, M.W.

    1987-01-01

    On-load refuelling of Advanced Gas Cooled Reactors results in a long, slender, articulated fuel assembly being suspended within a fuel channel, up which flows the high density gaseous coolant. The gas flow in the fuel assembly-channel annulus can cause vibration of the fuel assembly. This paper reports on continuing studies of this phenomenon. In particular it outlines the latest findings on the excitation mechanism, flow instabilities in an annular diffuser; successful developments in finite element modelling of the fuel assembly vibration which now include flow effects and non linearities caused by fuel assembly-channel impact; and finally experimental demonstration of the beneficial effect of introducing friction dampers into the fuel assembly. (author)

  17. 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)

  18. 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

  19. 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

  20. 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

  1. 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.

  2. 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.

  3. The development of SIMMER-III, an advanced computer program for LMFR safety analysis, and its application to sodium experiments

    Energy Technology Data Exchange (ETDEWEB)

    Tobita, Y.; Kondo, S.A.; Yamano, H. [Japan Nuclear Cycle Development Institute, ATD0OEC, 4002 Narita O-arai, Ibaraki, 311-1393 (Japan); Morita, K. [Kyusyu University, Institute of Enviromental Systems 6-10-1 Hakozaki, Higashi-ku, Fukuoka 812-8581 (Japan); Maschek, W. [Forschungszentrum Karlsruhe, IKET, Postfach 3640 D-76021, Karlsruhe (Germany); Coste, P. [CEA, DRN/DTP/SMTH, CE de Grenoble 38054 Grenoble CEDEX 9 (France); Cadiou, T. [CEA, DENCAD/DER/SERI, CE de Cadarache 13108 Saint Paul lez Durance CEDEX (France)

    2006-07-01

    generalized and flexible code framework, along with improved numerical stability and accuracy, allows us to apply it to a variety of simple and complex multiphase flow problems. The code assessment program is an ongoing effort. Two major milestones have been achieved in the past by completing two assessment campaigns, Phase 1 and Phase 2: Phase 1 for fundamental code assessment of individual models and Phase 2 for integral code assessment for key phenomena relevant to liquid-metal-cooled fast reactor safety. Through this systematic code assessment program, comprehensive validation of the physical models has been conducted step-by-step. The assessment program has demonstrated that SIMMER-III is a state-of-the-art code with advanced models sufficiently flexible for simulating transient multiphase phenomena occurring during core disruptive accidents. This paper concentrates on the specifics of the code, mainly reflected in its application to sodium experiments related to the safety of liquid-metal-cooled fast reactors. (authors)

  4. Core Thermal-Hydraulic Design of a Sodium Cooled Fast Reactor for the U/TRU Fuel Modification

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Sun Rock; Cho, Chung Ho; Kim, Young Gyun; Song, Hoon; Park, Won Seok; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    The Korea Atomic energy Research Institute (KAERI) has been developing an advanced SFR design technology with the final goal of constructing a demonstration plant by 2028. The main objective of the SFR demonstration plant is to verify TRU metal fuel performance, large-scale reactor operation, and transmutation ability of high-level wastes. However, in the early stage, the SFR will run on low enriched uranium fuel due to a lack of TRU fuel qualification. After sequential evaluations of the fuel performance, the fissile fuel material will transform from uranium to LTRU (LWR-TRU), and then finally to MTRU (Mixed TRU of LTRU and recycled TRU). At the same time, the core configurations will be modified to meet the nuclear design requirements. Therefore, there is also a strong need to ensure a proper cooling capability during modifications of the entire core. In this work, the core thermal-hydraulic design for U/TRU fuel modification is performed using the SLTHEN (Steady-State LMR Thermal-Hydraulic Analysis Code Based on ENERGY Model) code. As the power distribution in a reactor core is not uniform, it requires a suitable flow allocation to each assembly. There are two ways of allocating the flow rates depending on the orifice positions. The inner officering scheme locates orifice plates in the lower part of the fuel assembly. Therefore, it is possible that the flow distribution is redesigned according to the core configurations. On the other hand, the outer officering scheme fixes orifice plates within the receptacle body throughout the entire plant lifetime. This has the advantage lower of fabrication costs and operating errors but included insufficient design flexibility. This paper provides comparative studies of orifice position for the core thermal-hydraulic design

  5. 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

  6. Experience and topical problems of surveillance and diagnosis of sodium-cooled fast breeders in the period of introducing prototype units

    International Nuclear Information System (INIS)

    Kochetkov, L.A.; Petrenko, A.A.

    1984-01-01

    The solution of the problem of increasing the safety and economy of sodium-cooled fast reactors is impossible unless appropriate surveillance and diagnostic systems have been developed. In the past, improvement of surveillance and diagnostic systems took the following directions: centralization of surveillance, increase of safety, coupling to computer and control systems. It is reported on experience gained in developing and operating individual surveillance and diagnostic systems for fast breeders. Basic objectives of further developing methods and measuring instruments, diagnostic procedures and standards of surveillance equipment are presented. (author)

  7. 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

  8. 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

  9. 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

  10. 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.

  11. 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

  12. 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

  13. 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)

  14. 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.

  15. 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

  16. 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.

  17. 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.

  18. 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.

  19. Report of the consultancy on review of thermophysical properties of materials for advanced water-cooled reactors. Working material

    International Nuclear Information System (INIS)

    1996-01-01

    Since 1990 the IAEA's Nuclear Power Technology Development Section has carried out a coordinated research programme on thermophysical properties of materials for advanced water cooled reactors. The objective of this activity has been to collect and systematize a thermophysical properties data base for light and heavy water reactor materials under normal operating and transient conditions. This activity has been organized within the frame of IAEA's International Working Group on Advanced Technologies for Water-cooled Reactors. The important thermophysical properties include thermal conductivity, thermal diffusivity, specific heat capacity, enthalpy, thermal expansion and others. Several organizations involved in this CRP have suggested establishment of a new programme to extend the database to include properties in the liquid region applicable to severe accidents, to critically assess and peer review the property data and correlations, and to recommend the most appropriate data. The purpose of the consultancy was to examine the interest in further cooperation, and, if appropriate, to prepare the scope and approach for a potential new international collaborative programme to collect and review thermophysical properties data for advanced water cooled reactors and to recommend the most appropriate data. Figs

  20. 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

  1. A carbon dioxide partial condensation direct cycle for advanced gas cooled fast and thermal reactors

    International Nuclear Information System (INIS)

    Yasuyoshi, Kato; Takeshi, NItawaki; Yoshio, Yoshizawa

    2001-01-01

    A carbon dioxide partial condensation direct cycle concept has been proposed for gas cooled fast and thermal reactors. The fast reactor with the concept are evaluated to be a potential alternative option to liquid metal cooled fast reactors, providing comparable cycle efficiency at the same core outlet temperature, eliminating the safety problems, simplifying the heat transport system and making easier plant maintenance. The thermal reactor with the concept is expected to be an alternative solution to current high temperature gas cooled reactors (HTGRs) with helium gas turbines, allowing comparable cycle efficiency at the moderate temperature of 650 C instead of 800 C in HTGRs. (author)

  2. 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

  3. VELM61 and VELM22: Multigroup cross-section libraries for sodium-cooled reactor shield analysis

    Energy Technology Data Exchange (ETDEWEB)

    Fu, C.Y.; Ingersoll, D.T.

    1987-04-01

    Two coupled neutron and photon multigroup cross-section libraries, derived from ENDF/B-V nuclear data, are described. The energy group structures, 61n/23..gamma.. and 22n/10..gamma.., are subsets of the Vitamin-E 174n/38..gamma.. group structure, and are tailored to the iron and sodium resonances, windows, and capture gamma-ray spectra. Each of the two libraries are available in two formats, the AMPX master format and the ANISN format. Cross sections for all materials in the Vitamin-E library were collapsed using a standard energy weighting function, and in addition, several cross-section sets for each of the major constituents of commercial grade sodium, stainless steel (types 304 and 316), and carbon steel were derived using several problem-dependent weighting functions for averaging the fine groups. Effects of various group structures and weighting functions on the accuracy of the broad group libraries are studied by ANISN analysis of a typical sodium-iron shield configuration.

  4. 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

  5. Advanced heat pump cycle for district heating and cooling systems. Second quarterly progress report

    Energy Technology Data Exchange (ETDEWEB)

    Radermacher, R.

    1991-10-01

    A new scheme to significantly improve the performance of the two stage vapor compression cycle by eliminating the rectifier was first investigated with the help of computer simulation, and then incorporated in the experimental setup. Simulation results show that the cycle with a bleed line (modified cycle without the rectifier) has 20 to 30% higher cooling COP as compared to the cycle with the rectifier. It is important to note that this improvement in COP is accompanied by 10 to 15% increase in cooling load. Initial experimental results along with operating experience and description of the data acquisition program are presented here. Results show that hear can be pumped from an average temperature of 0{degrees}C to an average temperature of 100{degrees}C with a pressure ratio as low as 7.1. Cooling COPs up to 1.0 were obtained for cooling loads of about 4.17 kW.

  6. 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)

  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. 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

  9. 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.

  10. 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

  11. Advanced qualification methodology for actively cooled high heat flux plasma facing components

    International Nuclear Information System (INIS)

    Durocher, A.

    2006-01-01

    High heat flux plasma facing components (PFCs) in steady state fusion devices require high reliability. This can be only guaranteed by a very high level of qualification obtained with a rigorous acceptance inspection protocol. These components have to withstand heat fluxes from the plasma in the range of 10-20 MW/m 2 involving a number of severe engineering constraints: (i) the armour materials must be refractory and compatible with plasma wall interaction requirements; (ii) the heat sink should have a high thermal conductivity, high mechanical resistance and sufficient weldability behaviour; (iii) the cooling system, which is generally based on a circulation of pressurized water in the PFCs heat sink, must offer a high thermal efficiency; (iv) the joint of the refractory armour material onto the metallic heat sink,. To meet the power exhaust needs of PFCs during plasma operation requires control of their thermal and mechanical integrity. The first step is to detect defects in the element, such as material discontinuities like cracks and debondings. These will cause hot spots on the armour materiel and may even lead to the destruction of the PFC e.g. critical flux event. As the heat exhaust capability and the PFCs lifetime during plasma operation will stem from the manufacturing quality, a set of qualification activities should be performed during the component development and subsequent manufacturing phases. The major progress brought by this methodology stems from the combination and the correlation of three techniques: thermomechanical modelling, high heat flux testing and advanced non-destructive techniques, such as active infrared thermography. The scheme is applied during all the qualification activities: research and development phase, prototype manufacture including damage study for high heat flux, first series fabrication to define acceptance criteria and commissioning of the series fabrication. The paper describes the qualification route, which has been

  12. 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.

  13. 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

  14. 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

  15. 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

  16. 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)

  17. Microgravity experiments on boiling and applications: research activity of advanced high heat flux cooling technology for electronic devices in Japan.

    Science.gov (United States)

    Suzuki, Koichi; Kawamura, Hiroshi

    2004-11-01

    Research and development on advanced high heat flux cooling technology for electronic devices has been carried out as the Project of Fundamental Technology Development for Energy Conservation, promoted by the New Energy and Industrial Technology Development Organization of Japan (NEDO). Based on the microgravity experiments on boiling heat transfer, the following useful results have obtained for the cooling of electronic devices. In subcooled flow boiling in a small channel, heat flux increases considerably more than the ordinary critical heat flux with microbubble emission in transition boiling, and dry out of the heating surface is disturbed. Successful enhancement of heat transfer is achieved by a capillary effect from grooved surface dual subchannels on the liquid supply. The critical heat flux increases 30-40 percent more than for ordinary subchannels. A self-wetting mechanism has been proposed, following investigation of bubble behavior in pool boiling of binary mixtures under microgravity. Ideas and a new concept have been proposed for the design of future cooling system in power electronics.

  18. 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

  19. Preliminary Test Requirements for the Performance Test of Passive Decay Heat Removal System of Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Tae Ho; Kwon, Young Min; Kim, Tae Joon; Eoh, Jae Hyuk; Lee, Yong Bum; Ha, Kwi Seok; Hwang, In Koo

    2009-06-15

    In order to verify the concept of safety grade passive decay removal system PDRC (Passive Decay heat Removal Circuit) of KALIMER-600 and the design features to resolve the design issues for securing the cooling performance, the performance test is implemented. In this report, the preliminary test requirements for using as a guideline to the design of the experimental facility were established. Since the experimental facility should be designed so as to simulate the various thermal- hydraulic phenomena, as closely as possible, to be occurred in reference reactor during the decay heat removal operation, the design characteristics of the reference reactor (KALIMER-600) were analyzed for drawing major constitutive elements to be simulated in the facility. The preliminary test matrix was set up by the analysis of various design basis events and then the key test matrix was determined. Also, the priority for various thermal hydraulic phenomena which should be considered in the design of the experimental facility was determined by analyzing the phenomena for each key test matrix. Based on the analysis, the general design requirements for experimental facility were prepared and the design requirements for fluid systems and instrumentation were established. The test requirements in this report will be reflected in the scaling analysis and the basic design of the experimental facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

  20. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. A review. Pt. I. Key areas

    International Nuclear Information System (INIS)

    Sinha, Nilay K.; Raj, Baldev

    2013-01-01

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative studies of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton registered GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or ready-made supply. Part I addresses key areas of design shortlisting, strategy, development and unification with a backdrop of international evolution. (orig.)

  1. 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

  2. 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.

  3. 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

  4. 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

  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. 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.

  7. 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.)

  8. 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.

  9. 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)

  10. 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.

  11. Review of Phenomenological Models for the Initial Phase HCDA Analysis in a Metal-Fueled Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kwon, Young Min; Lee, Ki Rim; Ha, Kwi Seok; Chang, Won Pyo; Suk, Soo Dong

    2009-03-01

    The safety aspects of the KALIMER design results from the advanced safety performance characteristics of its ternary alloy metallic fuel. The superior thermal, mechanical, and neutronic performance of the metal-fueled core assures inherent safety response to unprotected and multiple fault accidents which are HCDA initiating events. HCDA has received great attentions because of its significant consequence, leading to substantial core disruption, although its probability of occurrence is very low. The SAS4A code provides an integrated quantitative framework for examining the phenomenological behaviors under HCDA conditions. Various phenomenological models such as prefailure characterization, transient pin response, margins to cladding failure, axial in-pin fuel relocation prior to cladding breach, and molten fuel relocation after cladding breach are required for the HCDA analysis. The important mechanisms which introduce negative reactivity during HCDA are fuel extrusion and in-pin fuel relocation, and structural feedback through thermal-mechanical neutronic effects. This report describes the safety performance characteristics of the metal fuel as observed in ex-pile and in-pile tests, and describes associated theoretical models employed into the SAS4A HCDA analysis code. Most of such tests and experiments, and development of theoretical models have been performed for the IFR program by ANL. This report provides a phenomenological basis for gaining an understanding of the metal fuel performance characteristics that obtained from expile experiments and in-pile tests. This report will provide insight and direction for planning HCDA experiments and developing theoretical models in Korea later

  12. 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.

  13. Advanced Liquid Cooling for a Traction Drive Inverter Using Jet Impingement and Microfinned Enhanced Surfaces: Preprint

    Energy Technology Data Exchange (ETDEWEB)

    Waye, S. K.; Narumanchi, S.; Mihalic, M.; Moreno, G.; Bennion, K.; Jeffers, J.

    2014-08-01

    Jet impingement on plain and micro-finned enhanced surfaces was compared to a traditional channel flow configuration. The jets provide localized cooling to areas heated by the insulated-gate bipolar transistor and diode devices. Enhanced microfinned surfaces increase surface area and thermal performance. Using lighter materials and designing the fluid path to manage pressure losses increases overall performance while reducing weight, volume, and cost. Powering four diodes in the center power module of the inverter and computational fluid dynamics (CFD) modeling was used to characterize the baseline as well as jet-impingement-based heat exchangers. CFD modeling showed the thermal performance improvements should hold for a fully powered inverter. Increased thermal performance was observed for the jet-impingement configurations when tested at full inverter power (40 to 100 kW output power) on a dynamometer. The reliability of the jets and enhanced surfaces over time was also investigated. Experimentally, the junction-to- coolant thermal resistance was reduced by up to 12.5% for jet impingement on enhanced surfaces s compared to the baseline channel flow configuration. Base plate-to-coolant (convective) resistance was reduced by up to 37.0% for the jet-based configuration compared to the baseline, suggesting that while improvements to the cooling side reduce overall resistance, reducing the passive stack resistance may contribute to lowering overall junction-to-coolant resistance. Full inverter power testing showed reduced thermal resistance from the middle of the module baseplate to coolant of up to 16.5%. Between the improvement in thermal performance and pumping power, the coefficient of performance improved by up to 13% for the jet-based configuration.

  14. 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.

  15. 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.

  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. Building on knowledge base of sodium cooled fast breeder reactors to develop materials technology for fusion reactors

    International Nuclear Information System (INIS)

    Raj, B.; Rao, B.S.

    2007-01-01

    component degradation behaviour under fusion reactor operating conditions. Finally, the strategies in the development of human resources and knowledge management for advancement of nuclear technology in India will be highlighted. (authors)

  19. Modelling 3D crack propagation in ageing graphite bricks of Advanced Gas-cooled Reactor power plant

    Directory of Open Access Journals (Sweden)

    Thi-Tuyet-Giang Vo

    2015-10-01

    Full Text Available In this paper, crack propagation in Advanced Gas-cooled Reactor (AGR graphite bricks with ageing properties is studied using the eXtended Finite Element Method (X-FEM. A parametric study for crack propagation, including the influence of different initial crack shapes and propagation criteria, is conducted. The results obtained in the benchmark study show that the crack paths from X-FEM are similar to the experimental ones. The accuracy of the strain energy release rate computation in a heterogeneous material is also evaluated using a finite difference approach. Planar and non-planar 3D crack growth simulations are presented to demonstrate the robustness and the versatility of the method utilized. Finally, this work contributes to the better understanding of crack propagation behaviour in AGR graphite bricks and so contributes to the extension of the AGR plants’ lifetimes in the UK by reducing uncertainties.

  20. 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)

  1. Mandate a Man to Fish?: Technological Advance in Cooling Systems at U.S. Thermal Electric Plants

    OpenAIRE

    Victor M. Peredo-Alvarez; Allen S. Bellas; Ian Lange

    2015-01-01

    Steam-based electrical generating plants use large quantities of water for cooling. The potential environmental impacts of water cooling systems have resulted in their inclusion in the Clean Water Act's (CWA) Sections 316(a), related to thermal discharges and 316(b), related to cooling water intake. The CWA mandates a technological standard for water cooling systems. This analysis examines how the performance-adjusted rates of thermal emissions and water withdrawals for cooling units have cha...

  2. Integral design concepts of advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-11-01

    Under the sub-programme on non-electrical applications of advanced reactors, the International Atomic Energy Agency has been providing a worldwide forum for exchange of information on integral reactor concepts. Two Technical Committee meetings were held in 1994 and 1995 on the subject where state-of-the-art developments were presented. Efforts are continuing for the development of advanced nuclear reactors of both evolutionary and innovative design, for electricity, co-generation and heat applications. While single purpose reactors for electricity generation may require small and medium sizes under certain conditions, reactors for heat applications and co-generation would be necessary in the small and medium range and need to be located closer to the load centres. The integral design approach to the development of advanced light water reactors has received special attention over the past few years. Several designs are in the detailed design stage, some are under construction, one prototype is in operation. A need has been felt for guidance on a number of issues, ranging from design objectives to the assessment methodology needed to show how integral designs can meet these objectives, and also to identify their advantages and problem areas. The technical document addresses the current status of the design, safety and operational issues of integral reactors and recommends areas for future development

  3. Study on transient hydrogen behavior and effect on passive containment cooling system of the advanced PWR

    International Nuclear Information System (INIS)

    Wang Yan

    2014-01-01

    A certain amount of hydrogen will be generated due to zirconium-steam reaction or molten corium concrete interaction during severe accidents in the pressurized water reactor (PWR). The generated hydrogen releases into the containment, and the formed flammable mixture might cause deflagration or detonation to produce high thermal and pressure loads on the containment, which may threaten the integrity of the containment. The non-condensable hydrogen in containment may also reduce the steam condensation on the containment surface to affect the performance of the passive containment cooling system (PCCS). To study the transient hydrogen behavior in containment with the PCCS performance during the accidents is significant for the further study on the PCCS design and the hydrogen risk mitigation. In this paper, a new developed PCCS analysis code with self-reliance intellectual property rights, which had been validated by comparison on the transients in the containment during the design basis accidents with other developed PCCS analysis code, is brief introduced and used for the transient simulation in the containment under a postulated small break LOCA of cold-leg. The results show that the hydrogen will flow upwards with the coolant released from the break and spread in the containment by convection and diffusion, and it results in the increase of the pressure in the containment due to reducing the heat removal capacity of the PCCS. (author)

  4. CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Eshita [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400019 India (India); Nayak, Arun K. [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Vijayan, Pallippattu K., E-mail: vijayanp@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India)

    2015-10-15

    Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

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

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

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

  6. 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

  7. Application of Pulsed Electrical Fields for Advanced Cooling and Water Recovery in Coal-Fired Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Young Cho; Alexander Fridman

    2009-04-02

    The overall objective of the present work was to develop technologies to reduce freshwater consumption in a cooling tower of coal-based power plant so that one could significantly reduce the need of make-up water. The specific goal was to develop a scale prevention technology based an integrated system of physical water treatment (PWT) and a novel filtration method so that one could reduce the need for the water blowdown, which accounts approximately 30% of water loss in a cooling tower. The present study investigated if a pulsed spark discharge in water could be used to remove deposits from the filter membrane. The test setup included a circulating water loop and a pulsed power system. The present experiments used artificially hardened water with hardness of 1,000 mg/L of CaCO{sub 3} made from a mixture of calcium chloride (CaCl{sub 2}) and sodium carbonate (Na{sub 2}CO{sub 3}) in order to produce calcium carbonate deposits on the filter membrane. Spark discharge in water was found to produce strong shockwaves in water, and the efficiency of the spark discharge in cleaning filter surface was evaluated by measuring the pressure drop across the filter over time. Results showed that the pressure drop could be reduced to the value corresponding to the initial clean state and after that the filter could be maintained at the initial state almost indefinitely, confirming the validity of the present concept of pulsed spark discharge in water to clean dirty filter. The present study also investigated the effect of a plasma-assisted self-cleaning filter on the performance of physical water treatment (PWT) solenoid coil for the mitigation of mineral fouling in a concentric counterflow heat exchanger. The self-cleaning filter utilized shockwaves produced by pulse-spark discharges in water to continuously remove scale deposits from the surface of the filter, thus keeping the pressure drop across the filter at a relatively low value. Artificial hard water was used in the

  8. Advanced Fuel Pellet Materials and Fuel Rod Design for Water Cooled Reactors. Proceedings of a Technical Committee Meeting

    International Nuclear Information System (INIS)

    2010-10-01

    The economics of current nuclear power plants have improved through increased fuel burnup and longer fuel cycles, i.e. increasing the effective time that fuel remains in the reactor core and the amount of energy it generates. Efficient consumption of fissile material in the fuel element before it is discharged from the reactor means that less fuel is required over the reactor's life cycle, which results in lower amounts of fresh fuel, lower spent fuel storage costs, and less waste for ultimate disposal. Better utilization of fissile nuclear materials, as well as more flexible power manoeuvring, place challenging operational demands on materials used in reactor components, and first of all, on fuel and cladding materials. It entails increased attention to measures ensuring desired in-pile fuel performance parameters that require adequate improvements in fuel material properties and fuel rod designs. These are the main reasons that motivated the IAEA Technical Working Group on Fuel Performance and Technology (TWG-FPT) to recommend the organization of a Technical Committee Meeting on Advanced Fuel Pellet Materials and Fuel Rod Designs for Power Reactors. The proposal was supported by the IAEA TWGs on Advanced Technologies for Light and Heavy Water-Cooled Reactors (TWG-LWR and TWG-HWR), and the meeting was held at the invitation of the Government of Switzerland at the Paul Scherrer Institute in Villigen, from 23 to 26 November 2009. This was the third IAEA meeting on these subjects (the first was held in 1996 in Tokyo, Japan, and the second in 2003 in Brussels, Belgium), which reflects the continuous interest in the above issues among Member States. The purpose of the meeting was to review the current status in the development of fuel pellet materials and to explore recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting was attended by 45 specialists representing fuel vendors, nuclear utilities, research and development

  9. Development of heat transfer enhancement techniques for external cooling of an advanced reactor vessel

    Science.gov (United States)

    Yang, Jun

    Nucleate boiling is a well-recognized means for passively removing high heat loads (up to ˜106 W/m2) generated by a molten reactor core under severe accident conditions while maintaining relatively low reactor vessel temperature (Critical Heat Flux (CHF), becomes the key to the success of external passive cooling of reactor vessel undergoing core disrupture accidents. In the present study, two boiling heat transfer enhancement methods have been proposed, experimentally investigated and theoretically modelled. The first method involves the use of a suitable surface coating to enhance downward-facing boiling rate and CHF limit so as to substantially increase the possibility of reactor vessel surviving high thermal load attack. The second method involves the use of an enhanced vessel/insulation design to facilitate the process of steam venting through the annular channel formed between the reactor vessel and the insulation structure, which in turn would further enhance both the boiling rate and CHF limit. Among the various available surface coating techniques, metallic micro-porous layer surface coating has been identified as an appropriate coating material for use in External Reactor Vessel Cooling (ERVC) based on the overall consideration of enhanced performance, durability, the ease of manufacturing and application. Since no previous research work had explored the feasibility of applying such a metallic micro-porous layer surface coating on a large, downward facing and curved surface such as the bottom head of a reactor vessel, a series of characterization tests and experiments were performed in the present study to determine a suitable coating material composition and application method. Using the optimized metallic micro-porous surface coatings, quenching and steady-state boiling experiments were conducted in the Sub-scale Boundary Layer Boiling (SBLB) test facility at Penn State to investigate the nucleate boiling and CHF enhancement effects of the surface

  10. INVESTIGATION OF FUNDAMENTAL THERMAL-HYDRAULIC PHENOMENA IN ADVANCED GAS-COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    INVESTIGATION OF FUNDAMENTAL THERMAL-HYDRAULIC PHE

    2006-09-01

    INL LDRD funded research was conducted at MIT to experimentally characterize mixed convection heat transfer in gas-cooled fast reactor (GFR) core channels in collaboration with INL personnel. The GFR for Generation IV has generated considerable interest and is under development in the U.S., France, and Japan. One of the key candidates is a block-core configuration first proposed by MIT, has the potential to operate in Deteriorated Turbulent Heat Transfer (DTHT) regime or in the transition between the DTHT and normal forced or laminar convection regime during post-loss-of-coolant accident (LOCA) conditions. This is contrary to most industrial applications where operation is in a well-defined and well-known turbulent forced convection regime. As a result, important new need emerged to develop heat transfer correlations that make possible rigorous and accurate predictions of Decay Heat Removal (DHR) during post LOCA in these regimes. Extensive literature review on these regimes was performed and a number of the available correlations was collected in: (1) forced laminar, (2) forced turbulent, (3) mixed convection laminar, (4) buoyancy driven DTHT and (5) acceleration driven DTHT regimes. Preliminary analysis on the GFR DHR system was performed and using the literature review results and GFR conditions. It confirmed that the GFR block type core has a potential to operate in the DTHT regime. Further, a newly proposed approach proved that gas, liquid and super critical fluids all behave differently in single channel under DTHT regime conditions, thus making it questionable to extrapolate liquid or supercritical fluid data to gas flow heat transfer. Experimental data were collected with three different gases (nitrogen, helium and carbon dioxide) in various heat transfer regimes. Each gas unveiled different physical phenomena. All data basically covered the forced turbulent heat transfer regime, nitrogen data covered the acceleration driven DTHT and buoyancy driven DTHT

  11. A feasibility and optimization study to determine cooling time and burnup of advanced test reactor fuels using a nondestructive technique

    Energy Technology Data Exchange (ETDEWEB)

    Navarro, Jorge [Univ. of Utah, Salt Lake City, UT (United States)

    2013-12-01

    The goal of this study presented is to determine the best available non-destructive technique necessary to collect validation data as well as to determine burn-up and cooling time of the fuel elements onsite at the Advanced Test Reactor (ATR) canal. This study makes a recommendation of the viability of implementing a permanent fuel scanning system at the ATR canal and leads3 to the full design of a permanent fuel scan system. The study consisted at first in determining if it was possible and which equipment was necessary to collect useful spectra from ATR fuel elements at the canal adjacent to the reactor. Once it was establish that useful spectra can be obtained at the ATR canal the next step was to determine which detector and which configuration was better suited to predict burnup and cooling time of fuel elements non-destructively. Three different detectors of High Purity Germanium (HPGe), Lanthanum Bromide (LaBr3), and High Pressure Xenon (HPXe) in two system configurations of above and below the water pool were used during the study. The data collected and analyzed was used to create burnup and cooling time calibration prediction curves for ATR fuel. The next stage of the study was to determine which of the three detectors tested was better suited for the permanent system. From spectra taken and the calibration curves obtained, it was determined that although the HPGe detector yielded better results, a detector that could better withstand the harsh environment of the ATR canal was needed. The in-situ nature of the measurements required a rugged fuel scanning system, low in maintenance and easy to control system. Based on the ATR canal feasibility measurements and calibration results it was determined that the LaBr3 detector was the best alternative for canal in-situ measurements; however in order to enhance the quality of the spectra collected using this scintillator a deconvolution method was developed. Following the development of the deconvolution method

  12. Assessment of Silicon Carbide Composites for Advanced Salt-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [ORNL; Wilson, Dane F [ORNL; Forsberg, Charles W [ORNL

    2007-09-01

    The Advanced High-Temperature Reactor (AHTR) is a new reactor concept that uses a liquid fluoride salt coolant and a solid high-temperature fuel. Several alternative fuel types are being considered for this reactor. One set of fuel options is the use of pin-type fuel assemblies with silicon carbide (SiC) cladding. This report provides (1) an initial viability assessment of using SiC as fuel cladding and other in-core components of the AHTR, (2) the current status of SiC technology, and (3) recommendations on the path forward. Based on the analysis of requirements, continuous SiC fiber-reinforced, chemically vapor-infiltrated SiC matrix (CVI SiC/SiC) composites are recommended as the primary option for further study on AHTR fuel cladding among various industrially available forms of SiC. Critical feasibility issues for the SiC-based AHTR fuel cladding are identified to be (1) corrosion of SiC in the candidate liquid salts, (2) high dose neutron radiation effects, (3) static fatigue failure of SiC/SiC, (4) long-term radiation effects including irradiation creep and radiation-enhanced static fatigue, and (5) fabrication technology of hermetic wall and sealing end caps. Considering the results of the issues analysis and the prospects of ongoing SiC research and development in other nuclear programs, recommendations on the path forward is provided in the order or priority as: (1) thermodynamic analysis and experimental examination of SiC corrosion in the candidate liquid salts, (2) assessment of long-term mechanical integrity issues using prototypical component sections, and (3) assessment of high dose radiation effects relevant to the anticipated operating condition.

  13. Palliative treatment for advanced biliary adenocarcinomas with combination dimethyl sulfoxide-sodium bicarbonate infusion and S-adenosyl-L-methionine.

    Science.gov (United States)

    Hoang, Ba X; Tran, Hung Q; Vu, Ut V; Pham, Quynh T; Shaw, D Graeme

    2014-09-01

    Adenocarcinoma of the gallbladder and cholangiocarcinoma account for 4% and 3%, respectively, of all gastrointestinal cancers. Advanced biliary tract carcinoma has a very poor prognosis with all current available modalities of treatment. In this pilot open-label study, the authors investigated the efficacy and safety of a combination of dimethyl sulfoxide-sodium bicarbonate (DMSO-SB) infusion and S-adenosyl-L-methionine (ademetionine) oral supplementation as palliative pharmacotherapy in nine patients with advanced nonresectable biliary tract carcinomas (ABTCs). Patients with evidence of biliary obstruction with a total serum bilirubin ≤300 μmol/L were allowed to join the study. The results of this 6-month study and follow-up of all nine patients with ABTC indicated that the investigated combination treatment improved pain control, blood biochemical parameters, and quality of life for the patients. Moreover, this method of treatment has led to a 6-month progression-free survival for all investigated patients. The treatment was well tolerated for all patients without major adverse reactions. Given that ABTC is a highly fatal malignancy with poor response to chemotherapy and targeted drugs, the authors consider that the combination of DMSO-SB and ademetionine deserves further research and application as a palliative care and survival-enhancing treatment for this group of patients.

  14. Advanced fuel pellet materials and designs for water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2004-10-01

    with the technological advances attempted in doping of fuel pellets with the primary objective of obtaining larger grains. While most of the papers gave an account of the experimental studies on addition of various dopants in different fuel materials, some of them outlined the behaviour of such pellets at sintering process. Papers dealing with 'Fission gas release from fuel pellets under high burnup conditions were presented in Session 3. Session 4 was devoted to the evolution of fuel pellet structure and thermal properties at high burnup. Session 5 was dealing with fuel pellet-cladding interaction (PCI) being a complex phenomenon that may lead to cladding failure and subsequent release of fission products into the reactor coolant. Research efforts to understand better the PCI phenomenon and minimize it with design solutions are considered necessary

  15. Advances in conceptual design of a gas-cooled accelerator driven system (ADS) transmutation devices to sustainable nuclear energy development

    Energy Technology Data Exchange (ETDEWEB)

    Garcia, Rosales; Fajardo, Garcia; Curbelo, Perez; Oliva, Munoz; Hernandez, Garcia, E-mail: jrosales@instec.cu [Higher Institute of Technologies and Applied Sciences, Habana City (Cuba); Castells, Escriva [Energetic Engeniering Institute, Politechnical University of Valencia, Valencia (Spain); Abanades [Department of Simulation of Termoenergetic Systems, Politechnical University of Madrid, Madrid (Spain)

    2011-07-01

    The possibilities of a nuclear energy development are considerably increasing with the world energetic demand increment. However, the management of nuclear waste from conventional nuclear power plants and its inventory minimization are the most important issues that should be addressed. Fast reactors and Accelerator Driven Systems (ADS) are the main options to reduce the long-lived radioactive waste inventory. Pebble Bed Very High Temperature advanced systems have great perspectives to assume the future nuclear energy development challenges. The conceptual design of a Transmutation Advanced Device for Sustainable Energy Applications (TADSEA) has been made in preliminary studies. The TADSEA is an ADS cooled by helium and moderated by graphite that uses as fuel small amounts of transuranic elements in the form of TRISO particles, confined in 3 cm radius graphite pebbles forming a pebble bed configuration. It would be used for nuclear waste transmutation and energy production. In this paper, the results of a method for calculating the number of whole pebbles fitting in a volume according to its size are showed. From these results, the packing fraction influence on the TADSEAs main work parameters is studied. In addition, a redesign of the previous configuration, according to the established conditions in the preliminary design, i.e. the exit thermal power, is made. On the other hand, the heterogeneity of the TRISO particles inside the pebbles can not be negligible. In this paper, a study of the power density distribution inside the pebbles by means of a detailed simulation of the TRISO fuel particles and using an homogeneous composition of the fuel is addressed. (author)

  16. Advanced Gas Cooled Fast Reactor Preliminary Design - 300 MWe Project Status And Trends For a Higher Unit Power Selection

    International Nuclear Information System (INIS)

    Poette, C.; Garnier, J.C.; Conti, A.; Bosq, J.C.; Mathieu, B.; Gaillard, J.P.; Bassi, C.

    2004-01-01

    The requirements for future nuclear energy systems, among which sustainability by reduction of long term radioactive waste and full utilization of fuel resources, are strong incentives for the development of advanced fast reactors. Moreover, self-sustaining fuel cycle and homogeneous actinides recycling are searched for to provide a high proliferation resistance. This paper presents the current design status of the 600 MWth (300 MWe) direct cycle helium cooled fast reactor including core, primary system layout and preliminary safety options. Impact of unit power in the range of 600 MWth - 3000 MWth on core design, safety options and fuel development issues is also presented and discussed. Results indicate the feasibility of a higher unit power selection that could lead to the deployment of a large fleet of electricity generating reactors. As a key step in the gas fast 'reactor development, the Experimental Test and Demonstration Reactor (ETDR) to be built in CEA-Cadarache for operation in 2015 is necessary for the qualification of gas fast reactor fuel and materials technology as well as core physics tools validation. The paper details the ETDR objectives and current design options. (authors)

  17. Recent Advances in the Use of Sodium Borohydride as a Solid State Hydrogen Store

    Directory of Open Access Journals (Sweden)

    Jianfeng Mao

    2015-01-01

    Full Text Available The development of new practical hydrogen storage materials with high volumetric and gravimetric hydrogen densities is necessary to implement fuel cell technology for both mobile and stationary applications. NaBH4, owing to its low cost and high hydrogen density (10.6 wt%, has received extensive attention as a promising hydrogen storage medium. However, its practical use is hampered by its high thermodynamic stability and slow hydrogen exchange kinetics. Recent developments have been made in promoting H2 release and tuning the thermodynamics of the thermal decomposition of solid NaBH4. These conceptual advances offer a positive outlook for using NaBH4-based materials as viable hydrogen storage carriers for mobile applications. This review summarizes contemporary progress in this field with a focus on the fundamental dehydrogenation and rehydrogenation pathways and properties and on material design strategies towards improved kinetics and thermodynamics such as catalytic doping, nano-engineering, additive destabilization and chemical modification.

  18. Sodium fire suppression

    International Nuclear Information System (INIS)

    Malet, J.C.

    1979-01-01

    Ignition and combustion studies have provided valuable data and guidelines for sodium fire suppression research. The primary necessity is to isolate the oxidant from the fuel, rather than to attempt to cool the sodium below its ignition temperature. Work along these lines has led to the development of smothering tank systems and a dry extinguishing powder. Based on the results obtained, the implementation of these techniques is discussed with regard to sodium fire suppression in the Super-Phenix reactor. (author)

  19. Mandate a Man to Fish?: Technological advance in cooling systems at U.S. thermal electric plants

    Science.gov (United States)

    Peredo-Alvarez, Victor M.; Bellas, Allen S.; Trainor-Guitton, Whitney J.; Lange, Ian

    2016-02-01

    Steam-based electrical generating plants use large quantities of water for cooling. The potential environmental impacts of water cooling systems have resulted in their inclusion in the Clean Water Act's (CWA) Sections 316(a), related to thermal discharges and 316(b), related to cooling water intake. The CWA mandates a technological standard for water cooling systems. This analysis examines how the performance-adjusted rates of thermal emissions and water withdrawals for cooling units have changed over their vintage and how these rates of change were impacted by imposition of the CWA. Results show that the rate of progress increased for cooling systems installed after the CWA whilethere was no progress previous to it.

  20. Contribution to the study of the transmission of ultrasound at a solid - gas - liquid interface. Application to non-destructive testing of the fourth generation of liquid sodium cooled reactors

    International Nuclear Information System (INIS)

    Paumel, K.

    2008-01-01

    One of the ways envisaged for the ultrasonic inspection of the fourth generation of liquid sodium cooled reactors is to use a transducer immersed in sodium. A good acoustic coupling of the transducer with sodium is needed. However, without special precautions, it is not obtained in all situations. The goal is to study the conditions for the appearance of a very bad acoustic coupling. Under certain conditions, the non wetting of the surface of the transducer by sodium causes trapping gas pockets in the roughness. Moreover, increasing amounts of surface gas fraction induces a sharp drop in the transmission of ultrasound. A first quasi-static analysis based on the crevice model allows to study the dependence of the stability of these gas pockets on the temperature, the hydrostatic pressure, and the level of dissolved gas saturation of the liquid. Modelling the dynamic behaviour of a simple gas pocket geometry and conducting an in-water viewing experience show that the gas surface fraction does not increase as a result of sound pressure transducer. In order to develop a parametric study based on the size and gas surface fraction, several samples are made. An ultrasonic experiment using various frequencies can measure the transmission through these samples. Meanwhile, three different models describing the experimental setup are proposed. The comparison of experimental and analytical results (of the last model) show a similar pattern of the dependence of the transmission on the various parameters. (author) [fr

  1. Methodology for Extraction of Remaining Sodium of Used Sodium Containers

    International Nuclear Information System (INIS)

    Jung, Minhwan; Kim, Jongman; Cho, Youngil; Jeong, Jiyoung

    2014-01-01

    Sodium used as a coolant in the SFR (Sodium-cooled Fast Reactor) reacts easily with most elements due to its high reactivity. If sodium at high temperature leaks outside of a system boundary and makes contact with oxygen, it starts to burn and toxic aerosols are produced. In addition, it generates flammable hydrogen gas through a reaction with water. Hydrogen gas can be explosive within the range of 4.75 vol%. Therefore, the sodium should be handled carefully in accordance with standard procedures even though there is a small amount of target sodium remainings inside the containers and drums used for experiment. After the experiment, all sodium experimental apparatuses should be dismantled carefully through a series of draining, residual sodium extraction, and cleaning if they are no longer reused. In this work, a system for the extraction of the remaining sodium of used sodium drums has been developed and an operation procedure for the system has been established. In this work, a methodology for the extraction of remaining sodium out of the used sodium container has been developed as one of the sodium facility maintenance works. The sodium extraction system for remaining sodium of the used drums was designed and tested successfully. This work will contribute to an establishment of sodium handling technology for PGSFR. (Prototype Gen-IV Sodium-cooled Fast Reactor)

  2. Technical meeting on advanced fuel pellet materials and fuel rod designs for water cooled reactors. Book of abstracts

    International Nuclear Information System (INIS)

    2009-01-01

    Heavy Water-Cooled Reactors (TWGLWR and TWGHWR) with a proposal to hold it at the Paul Scherrer Institute, Switzerland. The purpose of the meeting is to provide an overview on the status and perspective of fuel pellet materials development and recent improvements in fuel rod designs for light and heavy water cooled power reactors. The meeting will cover both light and heavy water reactor fuels with the following main objectives: - Consideration of modern technological and design tools enabling reliable performance of fuels and rod columns in current and planned operational environments; - Analysis of high burnup fuel structure and properties, including RIM effects, thermal behaviour, fission gas release, PCI and PCMI; - Discussion on specific features of MOX fuel, as well as perspectives on advanced fuels like Vibro-pack, Thorium fuel and others. Each of the papers in this book of abstracts has been indexed separately

  3. Low-Cost Manufacturing Technique for Advanced Regenerative Cooling for In-Space Cryogenic Engines, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — The goal of the proposed effort is to use selective laser melting (SLM, an additive manufacturing technique) to manufacture a hot fire-capable, water-cooled spool...

  4. Advances in nuclear science and technology

    CERN Document Server

    Greebler, Paul

    1968-01-01

    Advances in Nuclear Science and Technology Volume 4 provides information pertinent to the fundamental aspects of advanced reactor concepts. This book discusses the advances in various areas of general applicability, including modern perturbation theory, optimal control theory, and industrial application of ionizing radiations.Organized into seven chapters, this volume begins with an overview of the technology of sodium-cooled fast breeder power reactors and gas-cooled power reactors. This text then examines the key role of reactor safety in the development of fast breeder reactors. Other chapt

  5. Synergy between scientific advancement and technological innovation, illustrated by a mechanism-based model characterizing sodium-glucose cotransporter-2 inhibition.

    Science.gov (United States)

    Zhang, Liping; Ng, Chee M; List, James F; Pfister, Marc

    2010-09-01

    Advances in experimental medicine and technological innovation during the past century have brought tremendous progress in modern medicine and generated an ever-increasing amount of data from bench and bedside. The desire to extend scientific knowledge motivates effective data integration. Technological innovation makes this possible, which in turn accelerates the advancement in science. This mutually beneficial interaction is illustrated by the development of an expanded mechanism-based model for understanding a novel mechanism, sodium-glucose cotransporter-2 SGLT2 inhibition for potential treatment of type 2 diabetes mellitus.

  6. Use of oxygen dosing to prevent flow accelerated corrosion in British Energy's Advanced Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Quirk, G.P.; Woolsey, I.S.; Rudge, A.

    2010-01-01

    Flow accelerated corrosion (FAC) was recognized as major threat to the carbon steel feed and economizer tubing of the once-through boilers of the UK's Advanced Gas-cooled Reactors (AGRs) following the observation of FAC damage of the boiler inlet orifice assemblies at two plants in 1977, and subsequent review of the likelihood of further damage elsewhere within the boilers of all AGRs. In most cases, replacement of susceptible tubing was not feasible; due to the inaccessibility of the boiler components within the reactor concrete pressure vessel. Preventing further FAC damage within the boilers therefore had to rely largely on changes to the boiler feedwater chemistry. Following extensive research programs carried out in the late 1970s and early 1980s two main feedwater chemistry regimes were adopted to suppress FAC in different AGRs. The four units found to be at greatest risk of FAC damage adopted an oxygen dosed All Volatile Treatment (AVT) regime during commissioning, while four other units retained the original deoxygenated ammonia dosed AVT regime, but with an increased feedwater pH. The deoxygenated ammonia dosed chemistry regime was also adopted in four AGR units subsequently built, which used 1%Cr0.5%Mo feed and economizer tubing in their once-through boilers. The oxygen dosed AVT chemistry regime adopted in four units having helical once-through boilers has proved highly effective in preventing FAC, with no evidence of damage after around 150,000 hours of operation. However, FAC damage was eventually found in some of the other units operating with a deoxygenated feedwater chemistry regime, in spite of having adopted an elevated feedwater pH. These units have now successfully converted to an oxygen dosed AVT feedwater chemistry regime to prevent further FAC damage, with the result that all 14 AGR reactors now operate with variants of the original oxygen dosed feedwater chemistry regime developed during the 1980s. The paper outlines the development of

  7. Laser-cooling and electromagnetic trapping of neutral atoms

    International Nuclear Information System (INIS)

    Phillips, W.D.; Migdall, A.L.; Metcalf, H.J.

    1986-01-01

    Until recently it has been impossible to confine and trap neutral atoms using electromagnetic fields. While many proposals for such traps exist, the small potential energy depth of the traps and the high kinetic energy of available atoms prevented trapping. We review various schemes for atom trapping, the advances in laser cooling of atomic beams which have now made trapping possible, and the successful magnetic trapping of cold sodium atoms

  8. Experimental tests and qualification of analytical methods to address thermohydraulic phenomena in advanced water cooled reactors. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2000-05-01

    Worldwide there is considerable experience in nuclear power technology, especially in water cooled reactor technology. Of the operating plants, in September 1998, 346 were light water reactors (LWRs) totalling 306 GW(e) and 29 were heavy water reactors (HWRs) totalling 15 GW(e). The accumulated experience and lessons learned from these plants are being incorporated into new advanced reactor designs. Utility requirements documents have been formulated to guide these design activities by incorporating this experience, and results from research and development programmes, with the aim of reducing costs and licensing uncertainties by establishing the technical bases for the new designs. Common goals for advanced designs are high availability, user-friendly features, competitive economics and compliance with internationally recognized safety objectives. Large water cooled reactors with power outputs of 1300 MW(e) and above, which possess inherent safety characteristics (e.g. negative Doppler moderator temperature coefficients, and negative moderator void coefficient) and incorporate proven, active engineered systems to accomplish safety functions are being developed. Other designs with power outputs from, for example, 220 MW(e) up to about 1300 MW(e) which also possess inherent safety characteristics and which place more emphasis on utilization of passive safety systems are being developed. Passive systems are based on natural forces and phenomena such as natural convection and gravity, making safety functions less dependent on active systems and components like pumps and diesel generators. In some cases, further experimental tests for the thermohydraulic conditions of interest in advanced designs can provide improved understanding of the phenomena. Further, analytical methods to predict reactor thermohydraulic behaviour can be qualified for use by comparison with the experimental results. These activities should ultimately result in more economical designs. The

  9. Effect of Water-Alcohol Injection and Maximum Economy Spark Advance on Knock-Limited Performance and Fuel Economy of a Large Air-Cooled Cylinder

    Science.gov (United States)

    Heinicke, Orville H.; Vandeman, Jack E.

    1945-01-01

    An investigation was conducted to determine the effect of a coolant solution of 25 percent ethyl alcohol, 25 percent methyl alcohol, and 50 percent water by volume and maximum-economy spark advance on knock-limited performance and fuel economy of a large air-cooled cylinder. The knock-limited performance of the cylinder at engine speeds of 2100 and 2500 rpm was determined for coolant-fuel ratios of 0.0, 0.2, and 0.4. The effect of water-alcohol injection on fuel economy was determined in constant charge-air flow tests. The tests were conducted at a spark advance of 20 deg B.T.C. and maximum-economy spark advance.

  10. Laser cooling of solids

    OpenAIRE

    Nemova, Galina

    2009-01-01

    Parallel to advances in laser cooling of atoms and ions in dilute gas phase, which has progressed immensely, resulting in physics Nobel prizes in 1997 and 2001, major progress has recently been made in laser cooling of solids. I compare the physical nature of the laser cooling of atoms and ions with that of the laser cooling of solids. I point out all advantages of this new and very promising area of laser physics. Laser cooling of solids (optical refrigeration) at the present time can be lar...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-04-01

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

  12. Safety and radiation-enhancing effect of sodium glycididazole in loco regionally advanced laryngeal cancers previously treated with platinum-containing chemotherapy regimens: a preliminary report

    International Nuclear Information System (INIS)

    Zeng, Y.C.; Wu, R.; Xu, Z.G.; Zhang, X.Y.; Wu, L.N.; Wang, Y.M.; Zheng, W.; Chen, X.D.; Chi, F.; Zhang, Z.Y.; Li, X.; Jin, X.Y.; Chen, W.; Wang, S.L.; Xiao, F.D.; Wang, E.Y.; Dong, X.Q.; Jia, M.X.; Li, Y.; Fan, G.L.; Hao, S.H.; Zhang, L.B.; Zhang, H.B.; Xia, H.H.X.

    2010-01-01

    Purpose: To determine the safety and radiation-enhancing effect of sodium glycididazole in laryngeal squamous cell carcinoma (stage T3-4,N0-3,M0) with conventional radiotherapy. Patients and methods: Patients with locoregional advanced laryngeal cancer (stage T3-4,N0-3,M0) were included: group 1(control, n = 30)were not administered of sodium glycididazole; group 2 (test, n = 30) received sodium glycididazole at a dose of 700 mg/m2 intravenous infusion 30 minutes before radiotherapy three times a week. Surrogate end-points of efficacy were tumor and nodal size. Safety parameters were vomiting, nausea, mucositis, laryngeal edema, esophagus and skin reaction, dysphagia, dyspnea, neurological deficit. Patients were evaluated weekly during treatment for 7 weeks and thereafter monthly for 3 months. Results: In the test, the overall response rate was 88.89% (95% CI, 71.00-97.00%) at 7 weeks and 92.59% (95% CI, 76.00 to 99.00%) at 1 month of follow-up. In the control, the overall response rate was 62.5% (95% CI, 41.00 to 81.00%) at 7 weeks and 58.33% (95% CI, 37.00 to 78.00%) at 1 month of follow-up. The short-term locoregional response rate was better in the test group at 7 weeks (p = 0.027) and at 1 month (p = 0.005) of follow-up. The test group had significantly more nausea and vomiting in weeks 1 (p = 0.047), 2 (p = 0.007), and 3 (p = 0.01) of treatment. Conclusions: The study indicates sodium glycididazole is an effective radiation-enhancing agent that improves short-term locoregional control and is well tolerated in patients with loco regionally advanced laryngeal cancer. (authors)

  13. NASA Microclimate Cooling Challenges

    Science.gov (United States)

    Trevino, Luis A.

    2004-01-01

    The purpose of this outline form presentation is to present NASA's challenges in microclimate cooling as related to the spacesuit. An overview of spacesuit flight-rated personal cooling systems is presented, which includes a brief history of cooling systems from Gemini through Space Station missions. The roles of the liquid cooling garment, thermal environment extremes, the sublimator, multi-layer insulation, and helmet visor UV and solar coatings are reviewed. A second section is presented on advanced personal cooling systems studies, which include heat acquisition studies on cooling garments, heat rejection studies on water boiler & radiators, thermal storage studies, and insulation studies. Past and present research and development and challenges are summarized for the advanced studies.

  14. Advancement of compressible multiphase flows and sodium-water reaction analysis program SERAPHIM. Validation of a numerical method for the simulation of highly underexpanded jets

    International Nuclear Information System (INIS)

    Uchibori, Akihiro; Ohshima, Hiroyuki; Watanabe, Akira

    2010-01-01

    SERAPHIM is a computer program for the simulation of the compressible multiphase flow involving the sodium-water chemical reaction under a tube failure accident in a steam generator of sodium cooled fast reactors. In this study, the numerical analysis of the highly underexpanded air jets into the air or into the water was performed as a part of validation of the SERAPHIM program. The multi-fluid model, the second-order TVD scheme and the HSMAC method considering a compressibility were used in this analysis. Combining these numerical methods makes it possible to calculate the multiphase flow including supersonic gaseous jets. In the case of the air jet into the air, the calculated pressure, the shape of the jet and the location of a Mach disk agreed with the existing experimental results. The effect of the difference scheme and the mesh resolution on the prediction accuracy was clarified through these analyses. The behavior of the air jet into the water was also reproduced successfully by the proposed numerical method. (author)

  15. Advanced simulations of energy demand and indoor climate of passive ventilation systems with heat recovery and night cooling

    DEFF Research Database (Denmark)

    Hviid, Christian Anker; Svendsen, Svend

    with little energy consumption and with satisfying indoor climate. The concept is based on using passive measures like stack and wind driven ventilation, effective night cooling and low pressure loss heat recovery using two fluid coupled water-to-air heat exchangers developed at the Technical University......In building design the requirements for energy consumption for ventilation, heating and cooling and the requirements for increasingly better indoor climate are two opposing factors. This paper presents the schematic layout and simulation results of an innovative multifunctional ventilation concept...... simulation program ESP-r to model the heat and air flows and the results show the feasibility of the proposed ventilation concept in terms of low energy consumption and good indoor climate....

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

    International Nuclear Information System (INIS)

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

    1996-06-01

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

  17. Beam cooling

    OpenAIRE

    Danared, H

    2006-01-01

    Beam cooling is the technique of reducing the momentum spread and increasing the phase-space density of stored particle beams. This paper gives an introduction to beam cooling and Liouville’s theorem, and then it describes the three methods of active beam cooling that have been proven to work so far, namely electron cooling, stochastic cooling, and laser cooling. Ionization cooling is also mentioned briefly.

  18. The influence of cooling on the advance of lava flows: insights from analogue experiments on the feedbacks between flow dynamics and thermal structure

    Science.gov (United States)

    Garel, F.; Kaminski, E.; Tait, S.; Limare, A.

    2012-12-01

    During an effusive volcanic eruption, the crisis management is mainly based on the prediction of lava flows advance and its velocity. The spreading of a lava flow, seen as a gravity current, depends on its "effective rheology" and the eruptive mass flux. These two parameters are not known a priori during an eruption and a key question is how to evaluate them in near real-time (rather than afterwards.) There is no generic macroscopic model for the rheology of an advancing lava flow, and analogue modelling is a precious tool to empirically estimate the rheology of a complex flow. We investigate through laboratory experiments the simultaneous spreading and cooling of horizontal currents fed at constant rate from a point source. The materials used are silicone oil (isoviscous), and poly-ethylene glycol (PEG) wax injected in liquid state and solidiying during its advance. In the isoviscous case, the temperature field is a passive tracer of the flow dynamics, whereas in the PEG experiments there is a feedback between the cooling of the flow and its effective rheology. We focus on the evolution of the current area and of the surface thermal structure, imaged with an infrared camera, to assess how the thermal structure can be related to the flow rate. The flow advance is continuous in the viscous case, and follows the predictions of Huppert (1982); in that case the surface temperature become steady after a transient time and the radiated heat flux is shown to be proportional to the input rate. For the PEG experiments, the spreading occurs through an alternation of stagnation and overflow phases, with a mean spreading rate decreasing as the experiment goes on. As in the case of lava flows, these experiments can exhibit a compound flow field, solid levees, thermal erosion, liquid overflows and channelization. A key observation is that the effective rheology of the solifying PEG material depends on the input flow rate, with high input rates yielding a rheology closer to the

  19. SODIUM DEUTERIUM REACTOR

    Science.gov (United States)

    Oppenheimer, E.D.; Weisberg, R.A.

    1963-02-26

    This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

  20. Sodium purification in Rapsodie

    International Nuclear Information System (INIS)

    Giraud, B.

    1968-01-01

    This report is one of a series of publications presenting the main results of tests carried out during the start-up of the first french fast neutron reactor: Rapsodie. The article presents the sodium purification techniques used in the reactor cooling circuits both from the constructional point of view and with respect to results obtained during the first years working. (author) [fr

  1. Influence of fuel pin bowing on the temperature distribution in fuel pin cladding tubes in case of sodium cooling; experimental results

    International Nuclear Information System (INIS)

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

    1978-09-01

    The influence of rod bowing on the local temperature distribution was measured with turbulent sodium flow in the cladding tubes of a 19-rod bundle mock-up of the SNR 300 Mark Ia fuel element. Such measurements have been carried out for the first time. The results presented in this report are part 1 of the experimental evaluation not yet completed. The major results are: 1. When a rod on the first ring gets deformed towards a neighbour on the second ring with a gap reduction from the nominal value of 100 % down to 20 %, the maximum azimuthal temperature difference of the outer rod increases by about 60 %. 2. The maximum azimuthal temperature difference of a rod on the first ring increases by a factor of 2, if it is approached by a neighbour on the same ring. 3. The reduction in cross section of a subchannel by rod bowing results only locally in distinct temperature rises, i.e. in the adjacent cladding tubes. Rods of the next but one row are no more subject to noticeable changes in temperature [de

  2. Natural circulation data and methods for advanced water cooled nuclear power plant designs. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    2002-04-01

    The complex set of physical phenomena that occur in a gravity environment when a geometrically distinct heat sink and heat source are connected by a fluid flow path can be identified as natural circulation (NC). No external sources of mechanical energy for the fluid motion are involved when NC is established. Within the present context, natural convection is used to identify the phenomena that occur when a heat source is put in contact with a fluid. Therefore, natural convection characterizes a heat transfer regime that constitutes a subset of NC phenomena. This report provides the presented papers and summarizes the discussions at an IAEA Technical Committee Meeting (TCM) on Natural Circulation Data and Methods for innovative Nuclear Power Plant Design. While the planned scope of the TCM involved all types of reactor designs (light water reactors, heavy water reactors, gas-cooled reactors and liquid metal-cooled reactors), the meeting participants and papers addressed only light water reactors (LWRs) and heavy water reactors (HWRs). Furthermore, the papers and discussion addressed both evolutionary and innovative water cooled reactors, as defined by the IAEA. The accomplishment of the objectives of achieving a high safety level and reducing the cost through the reliance on NC mechanisms, requires a thorough understanding of those mechanisms. Natural circulation systems are usually characterized by smaller driving forces with respect to the systems that use an external source of energy for the fluid motion. For instance, pressure drops caused by vertical bends and siphons in a given piping system, or heat losses to environment are a secondary design consideration when a pump is installed and drives the flow. On the contrary, a significant influence upon the overall system performance may be expected due to the same pressure drops and thermal power release to the environment when natural circulation produces the coolant flow. Therefore, the level of knowledge for

  3. The dismantling of fast reactors: sodium processing

    International Nuclear Information System (INIS)

    Rodriguez, G.; Berte, M.; Serpante, J.P.

    1999-01-01

    Fast reactors require a coolant that does not slow down neutrons so water can not be used. Metallic sodium has been chosen because of its outstanding neutronic and thermal properties but sodium reacts easily with air and water and this implies that sodium-smeary components can not be considered as usual nuclear wastes. A stage of sodium neutralizing is necessary in the processing of wastes from fast reactors. Metallic sodium is turned into a chemically stable compound: soda, carbonates or sodium salts. This article presents several methods used by Framatome in an industrial way when dismantling sodium-cooled reactors. (A.C.)

  4. Advanced helium cooled pebble bed blanket with SiC{sub f}/SiC as structural material

    Energy Technology Data Exchange (ETDEWEB)

    Boccaccini, L.V. E-mail: lorenzo.boccaccini@iket.fzk.de; Fischer, U.; Gordeev, S.; Malang, S

    2000-11-01

    The Helium Cooled Pebble Bed blanket concept developed in the frame of the EPB-programme is based on the use of low activation ferritic/martensitic steel (EUROFER-97) as structural material. As the maximum allowable temperature of this steel is 550 deg. C, the coolant helium temperature can not exceed 450-500 deg. C, resulting in a relatively low thermal efficiency of the power generation system. The use of a ceramic material like SiC{sub f}/SiC with a maximum allowable temperature of 1300 deg. C allows to increase the maximum helium temperatures in the blanket, with the possibility to adopt more efficient power conversion systems. SiC{sub f}/SiC provides some other attractive features from the neutronic point of view (low neutron absorber in comparison to EUROFER) and safety (low activation). To take full advantage of the potential of this structural material, a new blanket design has been proposed. The pebble beds have been arranged in parallel to the first wall -- by this configuration it was possible to reduce the required amount of beryllium, to improve the tritium breeding ratio and increasing the allowable neutron fluence. Finally, the adopted flow scheme results in a decisive reduction of the coolant pressure drop. On the basis of this design thermo-mechanic, thermo-hydraulic and neutronic calculations have been performed to optimise the design parameters (number and thickness of the beds, {sup 6}Li enrichment, helium temperatures and pressure, etc.). An assessment of the limitation of this concept in term of maximum neutron wall, surface heating, achievable tritium breeding ratio, thermal efficiency in the power conversion system, pumping power for the blanket cooling loops has been performed.

  5. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. Pt. II. R and D necessities and development across the world. A review

    International Nuclear Information System (INIS)

    Sinha, Nilay K.; Raj, Baldev

    2013-01-01

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative study of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton registered GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or readymade supply. R and D necessities for inflatable seals and their development across the world are given closer look in Part II of the review in continuation of Part I. (orig.)

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

    International Nuclear Information System (INIS)

    Yurmanov, Victor; Lemekhov, Vadim; Smykov, Vladimir

    2012-09-01

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

  7. Wear behavior of 2-1/4 Cr-1Mo tubing against alloy 718 tube-support material in sodium-cooled steam generators

    International Nuclear Information System (INIS)

    Wilson, W.L.

    1983-05-01

    A series of prototypic steam generator 2-1/4 Cr-1 Mo tube/alloy 718 tube support plate wear tests were conducted in direct support of the Westinghouse Nuclear Components Division -- Breeder Reactor Components Project Large Scale steam Generator design. The initial objective was to verify the acceptable wear behavior of softer, ''over-aged'' alloy 718 support plate material. For all interfaces under all test conditions, resultant wear damage was adhesive in nature with varying amounts of 2-1/4 Cr-1 Mo tube material being adhesively transferred to the alloy 718 tube supports. Maximum tube wear depths exceeded the initially established design allowable limit of 127 μm (.005 in.) at 17 of the 18 interfaces tested. A decrease in contact stresses produced acceptable tube wear depths below a readjusted maximum design allowable value of 381 μm (.015 in.). Additional conservatisms associated with the simulation of a 40-year lifetime of rubbing in a one-week laboratory test provided further confidence that the 381 μm maximum tube wear allowance would not be exceeded in service. Softer, ''over-aged'' alloy 718 material was found to produce slightly less wear damage on 2-1/4 Cr-1 Mo tubing than fully age hardened material. Also, air formed oxide films on the alloy 718 reduced initial tube wear and delayed the onset of adhesive surface damage. However, at high surface stress levels, these films were not sufficiently stable to provide adequate long term protection from adhesive wear. The results of the present work and those of previous test programs suggest that the successful in-sodium tribological performance of 2-1/4 Cr-1 Mo/alloy 718 rubbing couples is dependent upon the presence of lubricative surface films, such as oxides and/or surface reaction or deposition products. 11 refs., 13 figs., 4 tabs

  8. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  9. High temperature sodium-concrete interactions

    International Nuclear Information System (INIS)

    Chasanov, M.G.; Staahl, G.E. Sr.

    1977-01-01

    Concrete specimens were immersed in sodium at 500 0 C, and the sodium-concrete interactions were studied. At this temperature the important reaction between CO 2 , produced by the limestone aggregate concrete, and sodium is 4Na+CO 2 → 2Na 2 O+C. This reaction is of interest for reactor safety analysis as it could act as a means of reducing gas pressures arising from CO 2 release by the concrete, in sodium cooled reactors. (B.D.)

  10. The influence of sodium fires on LMFBRs safety analysis

    International Nuclear Information System (INIS)

    Justin, F.

    1979-01-01

    In a sodium cooled reactor, sodium fires are accidental conditions to be taken into account in safety analysis. For the various sodium categories, fire conditions, associated risks, safety analysis objectives and detailed corresponding issues are indicated, An experimental research program can be deduced from these considerations. This report covers the following: safety analysis methodology; primary sodium fires; secondary sodium fires; auxiliary sodium fires, and related experimental research programs

  11. Stochastic cooling at Fermilab

    International Nuclear Information System (INIS)

    Marriner, J.

    1986-08-01

    The topics discussed are the stochastic cooling systems in use at Fermilab and some of the techniques that have been employed to meet the particular requirements of the anti-proton source. Stochastic cooling at Fermilab became of paramount importance about 5 years ago when the anti-proton source group at Fermilab abandoned the electron cooling ring in favor of a high flux anti-proton source which relied solely on stochastic cooling to achieve the phase space densities necessary for colliding proton and anti-proton beams. The Fermilab systems have constituted a substantial advance in the techniques of cooling including: large pickup arrays operating at microwave frequencies, extensive use of cryogenic techniques to reduce thermal noise, super-conducting notch filters, and the development of tools for controlling and for accurately phasing the system

  12. Advanced Sodium Ion Battery Anode Constructed via Chemical Bonding between Phosphorus, Carbon Nanotube, and Cross-Linked Polymer Binder.

    Science.gov (United States)

    Song, Jiangxuan; Yu, Zhaoxin; Gordin, Mikhail L; Li, Xiaolin; Peng, Huisheng; Wang, Donghai

    2015-12-22

    Maintaining structural stability is a great challenge for high-capacity conversion electrodes with large volume change but is necessary for the development of high-energy-density, long-cycling batteries. Here, we report a stable phosphorus anode for sodium ion batteries by the synergistic use of chemically bonded phosphorus-carbon nanotube (P-CNT) hybrid and cross-linked polymer binder. The P-CNT hybrid was synthesized through ball-milling of red phosphorus and carboxylic group functionalized carbon nanotubes. The P-O-C bonds formed in this process help maintain contact between phosphorus and CNTs, leading to a durable hybrid. In addition, cross-linked carboxymethyl cellulose-citric acid binder was used to form a robust electrode. As a result, this anode delivers a stable cycling capacity of 1586.2 mAh/g after 100 cycles, along with high initial Coulombic efficiency of 84.7% and subsequent cycling efficiency of ∼99%. The unique electrode framework through chemical bonding strategy reported here is potentially inspirable for other electrode materials with large volume change in use.

  13. Untersuchungen zur Optimiening der Kueh11uft-Ausblasekonfiguration fortschrittlicher Turbinenbeschaufelungen (Experiments on the Optimization of Cool Air Blow-Off Configurations of Advanced Turbine Blades )

    National Research Council Canada - National Science Library

    Ganzert, Wolfgang

    2000-01-01

    Deftly building upon a strong literature on film cooling in aerodynamics, the author observes various film cooling configurations in experiments designed to provide a better understanding of the TRACE...

  14. Impingement jet cooling in gas turbines

    CERN Document Server

    Amano, R S

    2014-01-01

    Due to the requirement for enhanced cooling technologies on modern gas turbine engines, advanced research and development has had to take place in field of thermal engineering. Impingement jet cooling is one of the most effective in terms of cooling, manufacturability and cost. This is the first to book to focus on impingement cooling alone.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-04-01

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

  16. High energy resolution and high count rate gamma spectrometry measurement of primary coolant of generation 4 sodium-cooled fast reactor; Spectrometrie gamma haute resolution et hauts taux de comptage sur primaire de reacteur de type generation 4 au sodium liquide

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R.

    2010-11-10

    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) [French] Les reacteurs a neutrons rapides refroidis au sodium sont en developpement en vue d'assurer une quatrieme generation de reacteurs repondant a la demande energetique, tout en assurant la preservation des ressources d'uranium par un fonctionnement en surgenerateur. L'objectif de la filiere est egalement d'ameliorer la gestion de la radiotoxicite des dechets produits par transmutation des actinides mineurs et de controler la non-proliferation par un fonctionnement en cycle ferme. Une instrumentation de surveillance et de controle de ce type de reacteur a ete etudiee dans cette these. La spectrometrie gamma de nouvelle generation permet, par les hauts taux de traitement aujourd'hui accessibles, d'envisager de nouvelles approches pour suivre avec une precision accrue la puissance neutronique et de detecter plus precocement des ruptures de gaine combustible. Des simulations numeriques ont ete realisees et une campagne d'essai a ete menee a bien sur le reacteur Phenix de Marcoule. Des perspectives prometteuses ont ete mises en exergue pour ces deux problematiques

  17. Characterization and management of radioactive sodium and other reactor components as input data for the decommissioning of liquid metal-cooled fast reactors. A compilation of data produced of data produced by members of the IAEA technical working group on fast reactors (TWG-FR) at two consultancies and one technical committee meeting. Working material

    International Nuclear Information System (INIS)

    2002-01-01

    A number of liquid metal cooled fast reactors (LMFRs) are in operation and, some have already been shut down; other reactors will reach the end of their design lifetime in a few years and become candidates for decommissioning. It is unfortunate that little consideration was devoted to decommissioning of reactors at the plant design and construction stage. It is with this focus that the Technical Working Group on Fast Reactors (TWGFR) recommended that the IAEA organize the exchange of information on LMFRs decommissioning technology. It was pointed out that the decommissioning of small sodium-cooled reactors has shown that there are two basic differences between thermal and fast reactors decommissioning: on the one side, the treatment and disposal of radioactive sodium coolant, and on the other side, the management of reactor components, for which the structural materials are activated in depth by fast neutrons. To this end, a Technical Committee Meeting on Sodium Removal and Disposal from LMFRs in Normal Operation and in the framework of Decommissioning (Aix-en-Provence, France, November 1997) and two Consultancies on Decommissioning of the Kazakh BN-350 LMFR (Vienna, Austria, October 1996; Obninsk, Russian Federation, February 1998) were convened by the IAEA. These Meetings brought together a group of experts from France, Russia, Kazakhstan, the UK, and the USA to exchange information on, and to review current technical knowledge and experience in the management of radioactive coolant and reactor components following closing of LMFRs, as well as their design features and operating experience relevant for decommissioning procedures. The report provides general and detailed information on activation characteristics of the primary coolant; treatment and disposal of the spent sodium; removal of the residual sodium deposits and decontamination; the activation characteristics of the reactor components and the management of the latter. The recurring theme is finding

  18. Understanding and Predicting Effect of Sodium Exposure on Microstructure of Grade 91 Steel

    Energy Technology Data Exchange (ETDEWEB)

    Li, Meimei [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [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 effect of sodium exposures on microstructure and tensile properties of Grade 91 (G91) steel in support of the design and operation of G91 components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY17 (M3AT-17AN1602018), under the Work Package AT-17AN160201, “SFR Materials Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program.

  19. A Synergistic Combination of Advanced Separation and Chemical Scale Inhibitor Technologies for Efficient Use of Imparied Water As Cooling Water in Coal-based Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Jasbir Gill

    2010-08-30

    Nalco Company is partnering with Argonne National Laboratory (ANL) in this project to jointly develop advanced scale control technologies that will provide cost-effective solutions for coal-based power plants to operate recirculating cooling water systems at high cycles using impaired waters. The overall approach is to use combinations of novel membrane separations and scale inhibitor technologies that will work synergistically, with membrane separations reducing the scaling potential of the cooling water and scale inhibitors extending the safe operating range of the cooling water system. The project started on March 31, 2006 and ended in August 30, 2010. The project was a multiyear, multi-phase project with laboratory research and development as well as a small pilot-scale field demonstration. In Phase 1 (Technical Targets and Proof of Concept), the objectives were to establish quantitative technical targets and develop calcite and silica scale inhibitor chemistries for high stress conditions. Additional Phase I work included bench-scale testing to determine the feasibility of two membrane separation technologies (electrodialysis ED and electrode-ionization EDI) for scale minimization. In Phase 2 (Technology Development and Integration), the objectives were to develop additional novel scale inhibitor chemistries, develop selected separation processes, and optimize the integration of the technology components at the laboratory scale. Phase 3 (Technology Validation) validated the integrated system's performance with a pilot-scale demonstration. During Phase 1, Initial evaluations of impaired water characteristics focused on produced waters and reclaimed municipal wastewater effluents. Literature and new data were collected and evaluated. Characteristics of produced waters vary significantly from one site to another, whereas reclaimed municipal wastewater effluents have relatively more uniform characteristics. Assessment to date confirmed that calcite and silica

  20. Characterizing the effects of elevated temperature on the air void pore structure of advanced gas-cooled reactor pressure vessel concrete using x-ray computed tomography

    Directory of Open Access Journals (Sweden)

    Withers P.J.

    2013-07-01

    Full Text Available X-ray computed tomography (X-ray CT has been applied to nondestructively characterise changes in the microstructure of a concrete used in the pressure vessel structure of Advanced Gas-cooled Reactors (AGR in the UK. Concrete specimens were conditioned at temperatures of 105 °C and 250 °C, to simulate the maximum thermal load expected to occur during a loss of coolant accident (LOCA. Following thermal treatment, these specimens along with an unconditioned control sample were characterised using micro-focus X-ray CT with a spatial resolution of 14.6 microns. The results indicate that the air void pore structure of the specimens experienced significant volume changes as a result of the increasing temperature. The increase in the porous volume was more prevalent at 250 °C. Alterations in air void size distributions were characterized with respect to the unconditioned control specimen. These findings appear to correlate with changes in the uni-axial compressive strength of the conditioned concrete.

  1. Sodium Oxybate

    Science.gov (United States)

    Sodium oxybate is used to prevent attacks of cataplexy (episodes of muscle weakness that begin suddenly and ... urge to sleep during daily activities, and cataplexy). Sodium oxybate is in a class of medications called ...

  2. Sodium Phosphate

    Science.gov (United States)

    Sodium phosphate is used in adults 18 years of age or older to empty the colon (large intestine, bowel) ... view of the walls of the colon. Sodium phosphate is in a class of medications called saline ...

  3. Passive Safety Systems in Advanced Water Cooled Reactors (AWCRS). Case Studies. A Report of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-09-01

    This report presents the results from the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) collaborative project (CP) on Advanced Water Cooled Reactor Case Studies in Support of Passive Safety Systems (AWCR), undertaken under the INPRO Programme Area C. INPRO was launched in 2000 - on the basis of a resolution of the IAEA General Conference (GC(44)/RES/21) - to ensure that nuclear energy is available in the 21st century in a sustainable manner, and it seeks to bring together all interested Member States to consider actions to achieve innovation. An important objective of nuclear energy system assessments is to identify 'gaps' in the various technologies and corresponding research and development (R and D) needs. This programme area fosters collaboration among INPRO Member States on selected innovative nuclear technologies to bridge technology gaps. Public concern about nuclear reactor safety has increased after the Fukushima Daiichi nuclear power plant accident caused by the loss of power to pump water for removing residual heat in the core. As a consequence, there has been an increasing interest in designing safety systems for new and advanced reactors that are passive in nature. Compared to active systems, passive safety features do not require operator intervention, active controls, or an external energy source. Passive systems rely only on physical phenomena such as natural circulation, thermal convection, gravity and self-pressurization. Passive safety features, therefore, are increasingly recognized as an essential component of the next-generation advanced reactors. A high level of safety and improved competitiveness are common goals for designing advanced nuclear power plants. Many of these systems incorporate several passive design concepts aimed at improving safety and reliability. The advantages of passive safety systems include simplicity, and avoidance of human intervention, external power or signals. For these reasons, most

  4. Insect sodium channels and insecticide resistance

    OpenAIRE

    Dong, Ke

    2007-01-01

    Voltage-gated sodium channels are essential for the generation and propagation of action potentials (i.e., electrical impulses) in excitable cells. Although most of our knowledge about sodium channels is derived from decades of studies of mammalian isoforms, research on insect sodium channels is revealing both common and unique aspects of sodium channel biology. In particular, our understanding of the molecular dynamics and pharmacology of insect sodium channels has advanced greatly in recent...

  5. Nuclear data sensitivity and uncertainty assessment of sodium voiding reactivity coefficients of an ASTRID-like sodium fast reactor

    Science.gov (United States)

    Nuria, García-Herranz; Anne-Laurène, Panadero; Ana, Martinez; Sandro, Pelloni; Konstantin, Mikityuk; Andreas, Pautz

    2017-09-01

    The EU 7th Framework ESNII+ project was launched in 2013 with the strategic orientation of preparing ESNII for Horizon 2020. ESNII stands for the European Industrial Initiative on Nuclear Energy, created by the European Commission in 2010 to promote the development of a new generation of nuclear systems in order to provide a sustainable solution to cope with Europe's growing energy needs while meeting the greenhouse gas emissions reduction target. The designs selected by the ESNII+ project are technological demonstrators of Generation-IV systems. The prototype for the sodium cooled fast reactor technology is ASTRID (standing for Advanced Sodium Technological Reactor for Industrial Demonstration), which detailed design phase is foreseen to be initiated in 2019. The ASTRID core has a peculiar design which was created in order to tackle the main neutronic challenge of sodium cooled fast reactors: the inherent overall positive reactivity feedback in case of sodium voiding occurring in the core. Indeed, the core is claimed by its designers to have an overall negative reactivity feedback in this scenario. This feature was demonstrated for an ASTRID-like core within the ESNII+ framework studies performed by nine European institutions. In order to shift the paradigm towards best-estimate plus uncertainties, the nuclear data sensitivity analysis and uncertainty propagation on reactivity coefficients has to be carried out. The goal of this work is to assess the impact of nuclear data uncertainties on sodium voiding reactivity feedback coefficients in order to get a more complete picture of the actual safety margins of the ASTRID low void-core design. The nuclear data sensitivity analysis is performed in parallel using SCALE TSUNAMI-3D and the newly developed GPT SERPENT 2 module. A comparison is carried out between the two methodologies. Uncertainty on the sodium reactivity feedbacks is then calculated using TSAR module of SCALE and the necessary safety margins conclusions

  6. Nuclear data sensitivity and uncertainty assessment of sodium voiding reactivity coefficients of an ASTRID-like sodium fast reactor

    Directory of Open Access Journals (Sweden)

    Nuria García-Herranz

    2017-01-01

    Full Text Available The EU 7th Framework ESNII+ project was launched in 2013 with the strategic orientation of preparing ESNII for Horizon 2020. ESNII stands for the European Industrial Initiative on Nuclear Energy, created by the European Commission in 2010 to promote the development of a new generation of nuclear systems in order to provide a sustainable solution to cope with Europe’s growing energy needs while meeting the greenhouse gas emissions reduction target. The designs selected by the ESNII+ project are technological demonstrators of Generation-IV systems. The prototype for the sodium cooled fast reactor technology is ASTRID (standing for Advanced Sodium Technological Reactor for Industrial Demonstration, which detailed design phase is foreseen to be initiated in 2019. The ASTRID core has a peculiar design which was created in order to tackle the main neutronic challenge of sodium cooled fast reactors: the inherent overall positive reactivity feedback in case of sodium voiding occurring in the core. Indeed, the core is claimed by its designers to have an overall negative reactivity feedback in this scenario. This feature was demonstrated for an ASTRID-like core within the ESNII+ framework studies performed by nine European institutions. In order to shift the paradigm towards best-estimate plus uncertainties, the nuclear data sensitivity analysis and uncertainty propagation on reactivity coefficients has to be carried out. The goal of this work is to assess the impact of nuclear data uncertainties on sodium voiding reactivity feedback coefficients in order to get a more complete picture of the actual safety margins of the ASTRID low void-core design. The nuclear data sensitivity analysis is performed in parallel using SCALE TSUNAMI-3D and the newly developed GPT SERPENT 2 module. A comparison is carried out between the two methodologies. Uncertainty on the sodium reactivity feedbacks is then calculated using TSAR module of SCALE and the necessary safety

  7. Spray cooling

    International Nuclear Information System (INIS)

    Rollin, Philippe.

    1975-01-01

    Spray cooling - using water spraying in air - is surveyed as a possible system for make-up (peak clipping in open circuit) or major cooling (in closed circuit) of the cooling water of the condensers in thermal power plants. Indications are given on the experiments made in France and the systems recently developed in USA, questions relating to performance, cost and environmental effects of spray devices are then dealt with [fr

  8. Na2.5Fe1.75(SO4)3/Ketjen/rGO: An advanced cathode composite for sodium ion batteries

    Science.gov (United States)

    Goñi, A.; Iturrondobeitia, A.; Gil de Muro, I.; Lezama, L.; Rojo, T.

    2017-11-01

    An advanced cathode composite Na2.5Fe1.75(SO4)3/Ketjen/rGO for sodium ion batteries has been prepared, joining together the excellent electrochemical properties of the three components: off stoichiometric iron sulfate alluaudite, Ketjen Black carbon and reduced graphene oxide (rGO). This electrode material has been exhaustively characterized by XRD, thermogravimetric analysis, Raman spectroscopy and SEM and TEM microscopy. The study has demonstrated that a high quality electrode material has been designed containing a porous sulfate core properly coated by interweaved rGO fibers and Ketjen Black nanoparticles. The electrochemical study has revealed an excellent performance providing specific capacities close to the theoretical one at 1C. Additionally, this composite has shown a very good rate capability and a great cycling stability for at least 200 cycles maintaining a coulombic efficiency of 96%. The post mortem analysis, which includes EPR and XPS measurements, has demonstrated that the carbonaceous coating on the composite generates a stable and protective SEI layer over the active material guaranteeing a successful performance during a long cycle life.

  9. Parametric Study on an Initial Cooling Performance in the KALIMER-600

    International Nuclear Information System (INIS)

    Han, Ji-Woong; Eoh, Jae-Hyuk; Lee, Tae-Ho; Kim, Seong-O

    2009-01-01

    Decay heat removal is very important in a nuclear power plant. The KALIMER-600, Korea Advanced Liquid MEtal Reactor, employs the PDRC(Passive Decay heat Removal Circuit) to remove the decay heat. DHX(Decay Heat eXchanger) in the PDRC of KALIMER-600 is disposed in the DHX support barrel located in the hot pool region. Each DHX support barrel has the lower end communicating with the cold pool such that the sodium free surface inside the barrel is maintained with the same level of the cold pool using the pumping head of the PHTS(Primary Heat Transport System) pumps. Consequently, DHX is not in direct contact with the cold pool sodium during a normal plant operation. Under transient conditions such as the loss of a normal heat sink accident, free surface outside the barrel rises up due to the expansion of the sodium induced by the core decay heat during the initial stage cooling. When it overflows into the cold pool through the DHX support barrel the heat removal via DHX is initiated and the second stage cooling begins. In order to secure the safety of a reactor until the activation of a second stage cooling by PDRC, it is very important to suppress the core temperature rising by an enhancement of the initial cooling performance. In this study the parametric investigations have been applied to reveal the effect of various design parameters on the initial cooling performance. The various design parameters such as coastdown flow, IHX(Intermediate Heat eXchanger) elevation, heat transfer via CCS (Cavity Cooling System) were considered. The numerical approaches based on a multidimensional analysis can be utilized as a useful tool to investigate overall transient behaviors within a pool. In this research the COMMIX-1AR/P code is utilized as a transient analysis tool in KALIMER-600 after a shut down. This study will provide the basic design information to improve the initial cooling performance in the KALIMER-600

  10. Friction and wear in sodium

    International Nuclear Information System (INIS)

    Hoffman, N.J.; Droher, J.J.

    1973-01-01

    In the design of a safe and reliable sodium-cooled reactor one of the more important problem areas is that of friction and wear of components immersed in liquid sodium or exposed to sodium vapor. Sodium coolant at elevated temperatures may severely affect most oxide-bearing surface layers which provide corrosion resistance and, to some extent, lubrication and surface hardness. Consequently, accelerated deterioration may be experienced on engaged-motion contact surfaces, which could result in unexpected reactor shutdown from component malfunction or failure due to galling and seizure. An overall view of the friction and wear phenomena encountered during oscillatory rubbing of surfaces in high-temperature, liquid-sodium environments is presented. Specific data generated at the Liquid Metal Engineering Center (LMEC) on this subject is also presented. (U.S.)

  11. Cooling tower

    International Nuclear Information System (INIS)

    Baer, E.; Dittrich, H.; Ernst, G.; Roller, W.

    1975-01-01

    The task on which the invention is based is to design a cooling tower in such a way that the negative influences of the wind, in particular strong side winds (wind velocities of over 10 m/s), on the functioning of the cooling tower are reduced or eliminated altogether. (orig./TK) [de

  12. Dimethylamine as a Replacement for Ammonia Dosing in the Secondary Circuit of an Advanced Gas-Cooled Reactor (AGR) Power Station

    International Nuclear Information System (INIS)

    Armstrong, C.; Mitchell, M.; Bull, A.; Quirk, G.P.; Rudge, A.

    2012-09-01

    Increasing flow resistance observed over recent years within the helical once-through boilers in the four Advanced Gas-Cooled Reactors (AGRs) at Hartlepool and Heysham 1 Power stations have reduced boiler performance, resulting in reductions in feedwater flow, steam temperatures, power output and the need to carry out periodic chemical cleaning. The root cause is believed to be the development of magnetite deposits with high flow impedance in the 9%Cr evaporator section of the boiler tubing. To prevent continued increases in boiler flow resistance, dimethylamine is being trialled, in one of the four affected units, as a replacement to the conventional ammonia dosing. Dimethylamine increases the pH at temperature around the secondary circuit and, based on full scale boiler rig simulations, is expected to reduce iron transport and prevent flow resistance increases within the evaporator section of the boiler. The dimethylamine plant trial commenced in January 2011 and is ongoing. The feedwater concentration of dimethylamine has been increased progressively towards a final target value of 900 μg kg -1 and its effect on iron transport and boiler pressure loss is being closely monitored. The high steam temperatures (>500 deg. C) of the secondary circuit lead to some decomposition of dimethylamine, which is being carefully monitored at various locations around the circuit. The decomposition products identified with dimethylamine dosing include ammonia, methylamine, formic acid, carbon dioxide and, as yet, unidentified neutral organic species. The effect of dimethylamine dosing on iron transport, boiler pressure drops and its decomposition behaviour around the secondary circuit during the plant trial will be presented in this paper. (authors)

  13. Sodium Variable Conductance Heat Pipe for Radioisotope Stirling Systems

    Science.gov (United States)

    Tarau, Calin; Anderson, William G.; Walker, Kara

    2009-01-01

    In a Stirling radioisotope system, heat must continually be removed from the General Purpose Heat Source (GPHS) modules to maintain the modules and surrounding insulation at acceptable temperatures. Normally, the Stirling convertor provides this cooling. If the converter stops in the current system, the insulation is designed to spoil, preventing damage to the GPHS, and also ending the mission. An alkali-metal Variable Conductance Heat Pipe (VCHP) has been designed to allow multiple stops and restarts of the Stirling convertor in an Advanced Stirling Radioisotope Generator (ASRG). When the Stirling convertor is turned off, the VCHP will activate when the temperatures rises 30 C above the setpoint temperature. A prototype VCHP with sodium as the working fluid was fabricated and tested in both gravity aided and against gravity conditions for a nominal heater head temperature of 790 C. The results show very good agreement with the predictions and validate the model. The gas front was located at the exit of the reservoir when heater head temperature was 790 C while cooling was ON, simulating an operating Advanced Stirling Converter (ASC). When cooling stopped, the temperature increased by 30 C, allowing the gas front to move past the radiator, which transferred the heat to the case. After resuming the cooling flow, the front returned at the initial location turning OFF the VCHP. The against gravity working conditions showed a colder reservoir and faster transients.

  14. Carbon and nitrogen transport in sodium systems

    International Nuclear Information System (INIS)

    Schrock, S.L.; Shiels, S.A.; Bagnall, C.

    1976-01-01

    Materials for the liquid metal cooled fast breeder reactor will be exposed to high temperature sodium for time periods up to 30 years. One consequence of this exposure will be changes in the interstitial element concentrations of the alloys and concomitant alterations in their mechanical behavior characteristics. Several ongoing technology programs have as their objective a quantitative definition of the rate and extent of this interstitial movement. The paper summarizes the status of these programs and reports in detail on the results of a recently completed, USERDA funded program at the Advanced Reactors Division of Westinghouse. These results, while substantiating earlier reported trends on interstitial movement, indicate the problem is not as severe as initially estimated. Moreover, the present wastage allowance for most reactor components contains sufficient conservatism to compensate for changes in mechanical strength resulting from this change in interstitial concentration

  15. Randomized in vivo trial evaluating plaque inhibition benefits of an advanced stannous-containing sodium fluoride dentifrice used in conjunction with power brush technology

    Science.gov (United States)

    Bellamy, PG; Boulding, A; Farmer, S; Day, TN; Barker, ML; Harris, R; Mussett, AJ

    2014-01-01

    Objective To compare the plaque inhibition efficacy of a novel stannous-containing sodium fluoride test dentifrice to a standard anticavity negative control dentifrice, when both were used in conjunction with an advanced oscillating–rotating (O/R) power toothbrush. Methods This was a randomized, two-treatment, three-period, double-blind crossover study conducted in a population using an O/R power brush. Subjects brushed twice per day with their assigned dentifrice during the three-treatment periods, each lasting for 17 consecutive days. Each period was separated by a 4-day washout period during which subjects continued to use their O/R power toothbrush. Plaque levels were assessed and averaged amongst three assessments taken on days 15, 16 and 17 at the end of each treatment period using digital plaque imaging analysis. Assessments were carried out on the facial anterior tooth surfaces in the morning before brushing (A.M. prebrush) following whole-mouth brushing (30 s per quadrant) with the assigned dentifrice (A.M. post-brush) and in the afternoon (P.M.). Results Twenty-seven subjects were randomized and completed the study. During the 17-day usage period, the stannous-containing test NaF dentifrice demonstrated a statistically significant lower mean plaque area versus the negative control dentifrice at each assessment timepoint; overnight A.M. prebrush was 33.8% lower (P toothbrush users had significantly less plaque coverage for all three measurements when using a stannous-containing NaF dentifrice than when using a negative control (fluoride) dentifrice. PMID:23844867

  16. Sodium in diet

    Science.gov (United States)

    Diet - sodium (salt); Hyponatremia - sodium in diet; Hypernatremia - sodium in diet; Heart failure - sodium in diet ... The body uses sodium to control blood pressure and blood volume. Your body also needs sodium for your muscles and nerves to work ...

  17. Hybrid Cooling Loop Technology for Robust High Heat Flux Cooling, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Advanced Cooling Technologies, Inc. (ACT) proposes to develop a hybrid cooling loop and cold plate technology for space systems thermal management. The proposed...

  18. Hybrid Cooling Loop Technology for Robust High Heat Flux Cooling Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Advanced Cooling Technologies, Inc. proposes to develop a hybrid cooling loop technology for space thermal control. The proposed technology combines the high heat...

  19. Ventilative Cooling

    DEFF Research Database (Denmark)

    Heiselberg, Per Kvols; Kolokotroni, Maria

    This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state-of-the-art of ventil......This report, by venticool, summarises the outcome of the work of the initial working phase of IEA ECB Annex 62 Ventilative Cooling and is based on the findings in the participating countries. It presents a summary of the first official Annex 62 report that describes the state......-of-the-art of ventilative cooling potentials and limitations, its consideration in current energy performance regulations, available building components and control strategies and analysis methods and tools. In addition, the report provides twenty six examples of operational buildings using ventilative cooling ranging from...

  20. Skeletal muscle sodium channelopathies.

    Science.gov (United States)

    Nicole, Sophie; Fontaine, Bertrand

    2015-10-01

    This is an update on skeletal muscle sodium channelopathies since knowledge in the field have dramatically increased in the past years. The relationship between two phenotypes and SCN4A has been confirmed with additional cases that remain extremely rare: severe neonatal episodic laryngospasm mimicking encephalopathy, which should be actively searched for since patients respond well to sodium channel blockers; congenital myasthenic syndromes, which have the particularity to be the first recessive Nav1.4 channelopathy. Deep DNA sequencing suggests the contribution of other ion channels in the clinical expressivity of sodium channelopathies, which may be one of the factors modulating the latter. The increased knowledge of channel molecular structure, the quantity of sodium channel blockers, and the availability of preclinical models would permit a most personalized choice of medication for patients suffering from these debilitating neuromuscular diseases. Advances in the understanding of the molecular structure of voltage-gated sodium channels, as well as availability of preclinical models, would lead to improved medical care of patients suffering from skeletal muscle, as well as other sodium channelopathies.

  1. Hidden Sodium

    Centers for Disease Control (CDC) Podcasts

    2013-03-04

    In this podcast, learn about reducing sodium intake by knowing what to eat and the main sources of sodium in the diet. It's important for a healthy lifestyle.  Created: 3/4/2013 by National Center for Chronic Disease Prevention and Health Promotion (NCCDPHP).   Date Released: 3/4/2013.

  2. Dietary sodium

    DEFF Research Database (Denmark)

    Graudal, Niels

    2015-01-01

    The 2013 Institute of Medicine (IOM) report "Sodium Intake in Populations: Assessment of Evidence" did not support the current recommendations of the IOM and the American Heart Association (AHA) to reduce daily dietary sodium intake to below 2,300 mg. The report concluded that the population......-based health outcome evidence was not sufficient to define a safe upper intake level for sodium. Recent studies have extended this conclusion to show that a sodium intake below 2,300 mg/day is associated with increased mortality. In spite of this increasing body of evidence, the AHA, Centers for Disease...... Control (CDC), other public health advisory bodies, and major medical journals have continued to support the current policy of reducing dietary sodium....

  3. C-Scan Performance Test of Under-Sodium ultrasonic Waveguide Sensor in Sodium

    International Nuclear Information System (INIS)

    Joo, Young Sang; Bae, Jin Ho; Kim, Jong Bum

    2011-01-01

    Reactor core and in-vessel structures of a sodium-cooled fast (SFR) are submerged in opaque liquid sodium in the reactor vessel. The ultrasonic inspection techniques should be applied for observing the in-vessel structures under hot liquid sodium. Ultrasonic sensors such as immersion sensors and rod-type waveguide sensors have developed in order to apply under-sodium viewing of the in-vessel structures of SFR. Recently the novel plate-type ultrasonic waveguide sensor has been developed for the versatile application of under-sodium viewing in SFR. In previous studies, the ultrasonic waveguide sensor module was designed and manufactured, and the feasibility study of the ultrasonic waveguide sensor was performed. To improve the performance of the ultrasonic waveguide sensor in the under-sodium application, a new concept of ultrasonic waveguide sensors with a Be coated SS304 plate is suggested for the effective generation of a leaky wave in liquid sodium and the non-dispersive propagation of A 0 -mode Lamb wave in an ultrasonic waveguide sensor. In this study, the C-scan performance of the under-sodium ultrasonic waveguide sensor in sodium has been investigated by the experimental test in sodium. The under-sodium ultrasonic waveguide sensor and the sodium test facility with a glove box system and a sodium tank are designed and manufactured to carry out the performance test of under-sodium ultrasonic waveguide sensor in sodium environment condition

  4. A numerical investigation of the sCO2 recompression cycle off-design behaviour, coupled to a sodium cooled fast reactor, for seasonal variation in the heat sink temperature

    International Nuclear Information System (INIS)

    Floyd, J.; Alpy, N.; Moisseytsev, A.; Haubensack, D.; Rodriguez, G.; Sienicki, J.; Avakian, G.

    2013-01-01

    Highlights: • Year-round behaviour of the supercritical CO 2 recompression cycle is simulated. • Behaviour of the system was uncertain due to large changes in the fluid properties. • Cycle thermodynamic optimisation and component preliminary designs were performed. • No off design cycle stability issues, compressors operate away from surge region. • Independent speed control of compressors maintains power and cycle efficiency. -- Abstract: Supercritical CO 2 cycles are particularly attractive for Generation IV Sodium-Cooled Fast Reactors (SFRs) as they can be simple and compact, but still offer steam-cycle equivalent efficiency while also removing potential for Na/H 2 O reactions. However, CO 2 thermophysical properties are very sensitive close to the critical point which raises, in particular, questions about the compressor and so cycle off-design behaviour when subject to inevitable temperature increases that result from seasonal variations in the heat sink temperature. This publication reports the numerical investigation of such an issue that has been performed using the Plant Dynamics Code (ANL, USA), the cycle being optimised for the next French SFR, ASTRID (1500 MW th ), as a test-case. On design, the net plant efficiency is 42.2% for a high pressure (25 MPa) turbine with an inlet temperature of 515 °C and considering a cycle low temperature of 35 °C. The off-design cycle behaviour is studied based on preliminary designs for the main components and assuming the use of a fixed heat sink flow rate. First results obtained using a common fixed shaft speed for all turbomachines, without any other active control, show no stability issues and roughly constant density (and volumetric flow rate) at the main compressor inlet for the range of heat sink temperature considered (21–40 °C). This occurs because the new stationary states are found without requiring a significant shift of mass to the higher pressure level, meaning the compressor inlet pressure

  5. Alkali Metal Backup Cooling for Stirling Systems - Experimental Results

    Science.gov (United States)

    Schwendeman, Carl; Tarau, Calin; Anderson, William G.; Cornell, Peggy A.

    2013-01-01

    In a Stirling Radioisotope Power System (RPS), heat must be continuously removed from the General Purpose Heat Source (GPHS) modules to maintain the modules and surrounding insulation at acceptable temperatures. The Stirling convertor normally provides this cooling. If the Stirling convertor stops in the current system, the insulation is designed to spoil, preventing damage to the GPHS at the cost of an early termination of the mission. An alkali-metal Variable Conductance Heat Pipe (VCHP) can be used to passively allow multiple stops and restarts of the Stirling convertor. In a previous NASA SBIR Program, Advanced Cooling Technologies, Inc. (ACT) developed a series of sodium VCHPs as backup cooling systems for Stirling RPS. The operation of these VCHPs was demonstrated using Stirling heater head simulators and GPHS simulators. In the most recent effort, a sodium VCHP with a stainless steel envelope was designed, fabricated and tested at NASA Glenn Research Center (GRC) with a Stirling convertor for two concepts; one for the Advanced Stirling Radioisotope Generator (ASRG) back up cooling system and one for the Long-lived Venus Lander thermal management system. The VCHP is designed to activate and remove heat from the stopped convertor at a 19 degC temperature increase from the nominal vapor temperature. The 19 degC temperature increase from nominal is low enough to avoid risking standard ASRG operation and spoiling of the Multi-Layer Insulation (MLI). In addition, the same backup cooling system can be applied to the Stirling convertor used for the refrigeration system of the Long-lived Venus Lander. The VCHP will allow the refrigeration system to: 1) rest during transit at a lower temperature than nominal; 2) pre-cool the modules to an even lower temperature before the entry in Venus atmosphere; 3) work at nominal temperature on Venus surface; 4) briefly stop multiple times on the Venus surface to allow scientific measurements. This paper presents the experimental

  6. Cardiac sodium channelopathies.

    Science.gov (United States)

    Amin, Ahmad S; Asghari-Roodsari, Alaleh; Tan, Hanno L

    2010-07-01

    Cardiac sodium channel are protein complexes that are expressed in the sarcolemma of cardiomyocytes to carry a large inward depolarizing current (INa) during phase 0 of the cardiac action potential. The importance of INa for normal cardiac electrical activity is reflected by the high incidence of arrhythmias in cardiac sodium channelopathies, i.e., arrhythmogenic diseases in patients with mutations in SCN5A, the gene responsible for the pore-forming ion-conducting alpha-subunit, or in genes that encode the ancillary beta-subunits or regulatory proteins of the cardiac sodium channel. While clinical and genetic studies have laid the foundation for our understanding of cardiac sodium channelopathies by establishing links between arrhythmogenic diseases and mutations in genes that encode various subunits of the cardiac sodium channel, biophysical studies (particularly in heterologous expression systems and transgenic mouse models) have provided insights into the mechanisms by which INa dysfunction causes disease in such channelopathies. It is now recognized that mutations that increase INa delay cardiac repolarization, prolong action potential duration, and cause long QT syndrome, while mutations that reduce INa decrease cardiac excitability, reduce electrical conduction velocity, and induce Brugada syndrome, progressive cardiac conduction disease, sick sinus syndrome, or combinations thereof. Recently, mutation-induced INa dysfunction was also linked to dilated cardiomyopathy, atrial fibrillation, and sudden infant death syndrome. This review describes the structure and function of the cardiac sodium channel and its various subunits, summarizes major cardiac sodium channelopathies and the current knowledge concerning their genetic background and underlying molecular mechanisms, and discusses recent advances in the discovery of mutation-specific therapies in the management of these channelopathies.

  7. Specialists meeting on sodium removal and decontamination. Summary report

    International Nuclear Information System (INIS)

    1978-08-01

    This report covers experiences on sodium removal techniques developed or gained in a number of countries running sodium cooled reactors. This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Cleaning of sodium wetted components or fuel assemblies was achieved by applying different techniques including vacuum distillation, using different alcohols or evaporation processes

  8. Safety and radiation-enhancing effect of sodium glycididazole in loco regionally advanced laryngeal cancers previously treated with platinum-containing chemotherapy regimens: a preliminary report; Tolerance et effet radiosensibilisateur du sodium glycididazole chez des patients atteints de cancer du larynx localement evolue ayant prealablement recu une chimiotherapie a base de cisplatine: rapport preliminaire

    Energy Technology Data Exchange (ETDEWEB)

    Zeng, Y.C.; Wu, R.; Xu, Z.G.; Zhang, X.Y.; Wu, L.N.; Wang, Y.M.; Zheng, W.; Chen, X.D.; Chi, F.; Zhang, Z.Y.; Li, X.; Jin, X.Y.; Chen, W.; Wang, S.L.; Xiao, F.D.; Wang, E.Y.; Dong, X.Q.; Jia, M.X.; Li, Y. [China Medical Univ., Dept. of Medical Oncology, Shengjing Hospital, Shenyang, PR (China); Fan, G.L. [Harbin First Hospital, Dept. of Otorhinolaryngology, PR (China); Hao, S.H.; Zhang, L.B.; Zhang, H.B. [General Hospital of Shenyang Military Region, Dept. of Nuclear Medicine and Radiology, Shenyang, PR (China); Xia, H.H.X. [Novartis Pharmaceuticals Corporation, East Hanover, New Jersey (United States)

    2010-01-15

    Purpose: To determine the safety and radiation-enhancing effect of sodium glycididazole in laryngeal squamous cell carcinoma (stage T3-4,N0-3,M0) with conventional radiotherapy. Patients and methods: Patients with locoregional advanced laryngeal cancer (stage T3-4,N0-3,M0) were included: group 1(control, n = 30)were not administered of sodium glycididazole; group 2 (test, n = 30) received sodium glycididazole at a dose of 700 mg/m2 intravenous infusion 30 minutes before radiotherapy three times a week. Surrogate end-points of efficacy were tumor and nodal size. Safety parameters were vomiting, nausea, mucositis, laryngeal edema, esophagus and skin reaction, dysphagia, dyspnea, neurological deficit. Patients were evaluated weekly during treatment for 7 weeks and thereafter monthly for 3 months. Results: In the test, the overall response rate was 88.89% (95% CI, 71.00-97.00%) at 7 weeks and 92.59% (95% CI, 76.00 to 99.00%) at 1 month of follow-up. In the control, the overall response rate was 62.5% (95% CI, 41.00 to 81.00%) at 7 weeks and 58.33% (95% CI, 37.00 to 78.00%) at 1 month of follow-up. The short-term locoregional response rate was better in the test group at 7 weeks (p = 0.027) and at 1 month (p = 0.005) of follow-up. The test group had significantly more nausea and vomiting in weeks 1 (p = 0.047), 2 (p = 0.007), and 3 (p = 0.01) of treatment. Conclusions: The study indicates sodium glycididazole is an effective radiation-enhancing agent that improves short-term locoregional control and is well tolerated in patients with loco regionally advanced laryngeal cancer. (authors)

  9. Non-aqueous removal of sodium from reactor components

    International Nuclear Information System (INIS)

    Welch, F.H.; Steele, O.P.

    1978-01-01

    Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component

  10. A randomized clinical trial to assess anti-plaque effects of an oral hygiene regimen with a stannous-containing sodium fluoride dentifrice, advanced manual toothbrush, and CPC rinse.

    Science.gov (United States)

    Feng, Xiping; He, Tao; Cao, Michelle; He, Yanyan; Ji, Nelson

    2016-04-01

    To assess the anti-plaque efficacy of an oral hygiene regimen comprised of a stannous-containing sodium fluoride dentifrice, advanced manual toothbrush, and a cetylpyridinium chloride (CPC) rinse compared to a negative control regimen. This was a 4-week randomized and controlled, parallel group, single-center, single- blind, clinical trial in generally healthy Chinese adults with existing dental plaque. Following a 1-week acclimation period and cessation of overnight oral hygiene prior to the baseline examination, overnight pre-brushing plaque levels were assessed via digital plaque imaging analysis (DPIA). Subjects were randomly assigned to either: (1) the test regimen of a stannous-containing sodium fluoride dentifrice (Crest Pro-Health Expert), an advanced manual toothbrush with CrissCross bristles (Crest Pro-Health manual toothbrush), and a 0.07% CPC rinse (Crest Pro-Health Multi-Protection); or (2) the negative control regimen group, a 0.243% sodium fluoride dentifrice (Crest Cavity Protection) and a soft flat trim manual toothbrush (Crest MeiLiLiangJie). Subjects returned at Week 2 and Week 4 following twice daily use of their assigned products, again following pre-visit cessation of overnight oral hygiene, for DPIA evaluation of overnight plaque levels. 35 fully evaluable subjects completed the trial. At Week 2, the pre-brushing overnight average DPIA plaque scores for the advanced products regimen group were 72.4% statistically significantly lower relative to the control group (P< 0.0001). At Week 4, the mean plaque inhibition benefit provided by the regimen group was 76.8% greater than the control group (P< 0.0001). All products were well-tolerated. (Am J Dent 2016;29:120-124)

  11. The development of a model to study the thermal behaviour of the coolant in the blind elements of a fast sodium-cooled breeder in the case of a severe hypothetical accident during the initial phase

    International Nuclear Information System (INIS)

    Genter, G.

    1981-03-01

    The enthalpy level of the coolant is studied in the interior of gaps and special elements of a fast sodium coded breeder reactor during the initial and the final stages of a hypothetical accident. For this purpose numerical models are presented to calculate the heat transport in the special element on the basis of heat conduction and axial convection. (orig./RW) [de

  12. A very cool cooling system

    CERN Multimedia

    Antonella Del Rosso

    2015-01-01

    The NA62 Gigatracker is a jewel of technology: its sensor, which delivers the time of the crossing particles with a precision of less than 200 picoseconds (better than similar LHC detectors), has a cooling system that might become the precursor to a completely new detector technique.   The 115 metre long vacuum tank of the NA62 experiment. The NA62 Gigatracker (GTK) is composed of a set of three innovative silicon pixel detectors, whose job is to measure the arrival time and the position of the incoming beam particles. Installed in the heart of the NA62 detector, the silicon sensors are cooled down (to about -20 degrees Celsius) by a microfluidic silicon device. “The cooling system is needed to remove the heat produced by the readout chips the silicon sensor is bonded to,” explains Alessandro Mapelli, microsystems engineer working in the Physics department. “For the NA62 Gigatracker we have designed a cooling plate on top of which both the silicon sensor and the...

  13. Synthesis of results obtained on sodium components and technology through the Generation IV International Forum SFR Component Design and Balance-of-Plant Project

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Rodriguez, G.; Kisohara, N.; Kim, J. B.; Gerber, A.; Ashurko, Y.; Toyama, S.

    2013-01-01

    Status: The viability of designing SFR components and BOP has been demonstrated with design, construction and operation of previous sodium-cooled reactors. The main objective of this R&D project is related to system performance, or by development on the use of AECS in the BOP that could allow further cost improvements. Objective: To conduct collaborative research and development of components and BOP for the SFR System. The Project has to satisfy the GIF’s criteria of safety, economy, sustainability, proliferation resistance and physical protection. Activities within this Project are addressing experimental and analytical evaluation of advanced ISI&R, LBB assessment, development of AECS with Brayton cycles, advanced SG technologies. Project activities will be based in part on the extensive historical R&D experience with component design and balance of plant for sodium-cooled fast reactors

  14. Cooling systems

    International Nuclear Information System (INIS)

    Coutant, C.C.

    1978-01-01

    Progress on the thermal effects project is reported with regard to physiology and distribution of Corbicula; power plant effects studies on burrowing mayfly populations; comparative thermal responses of largemouth bass from northern and southern populations; temperature selection by striped bass in Cherokee Reservoir; fish population studies; and predictive thermoregulation by fishes. Progress is also reported on the following; cause and ecological ramifications of threadfin shad impingement; entrainment project; aquaculture project; pathogenic amoeba project; and cooling tower drift project

  15. Electron Cooling of RHIC

    International Nuclear Information System (INIS)

    Ben-Zvi, I.; Barton, D.S.; Beavis, D.B.; Blaskiewicz, M.; Brennan, J.M.; Burrill, A.; Calaga, R.; Cameron, P.; Chang, X.Y.; Connolly, R.; Eidelman, Yu.I.; Fedotov, A.V.; Fischer, W.; Gassner, D.M.; Hahn, H.; Harrison, M.; Hershcovitch, A.; Hseuh, H.-C.; Jain, A.K.; Johnson, P.D.J.; Kayran, D.; Kewisch, J.; Lambiase, R.F.; Litvinenko, V.; MacKay, W.W.; Mahler, G.J.; Malitsky, N.; McIntyre, G.T.; Meng, W.; Mirabella, K.A.M.; Montag, C.; Nehring, T.C.N.; Nicoletti, T.; Oerter, B.; Parzen, G.; Pate, D.; Rank, J.; Rao, T.; Roser, T.; Russo, T.; Scaduto, J.; Smith, K.; Trbojevic, D.; Wang, G.; Wei, J.; Williams, N.W.W.; Wu, K.-C.; Yakimenko, V.; Zaltsman, A.; Zhao, Y.; Abell, D.T.; Bruhwiler, D.L.; Bluem, H.; Burger, A.; Cole, M.D.; Favale, A.J.; Holmes, D.; Rathke, J.; Schultheiss, T.; Todd, A.M.M.; Burov, A.V.; Nagaitsev, S.; Delayen, J.R.; Derbenev, Y.S.; Funk, L. W.; Kneisel, P.; Merminga, L.; Phillips, H.L.; Preble, J.P.; Koop, I.; Parkhomchuk, V.V.; Shatunov, Y.M.; Skrinsky, A.N.; Koop, I.; Parkhomchuk, V.V.; Shatunov, Y.M.; Skrinsky, A.N.; Sekutowicz, J.S.

    2005-01-01

    We report progress on the R and D program for electron-cooling of the Relativistic Heavy Ion Collider (RHIC). This electron cooler is designed to cool 100 GeV/nucleon at storage energy using 54 MeV electrons. The electron source will be a superconducting RF photocathode gun. The accelerator will be a superconducting energy recovery linac. The frequency of the accelerator is set at 703.75 MHz. The maximum electron bunch frequency is 9.38 MHz, with bunch charge of 20 nC. The R and D program has the following components: The photoinjector and its photocathode, the superconducting linac cavity, start-to-end beam dynamics with magnetized electrons, electron cooling calculations including benchmarking experiments and development of a large superconducting solenoid. The photoinjector and linac cavity are being incorporated into an energy recovery linac aimed at demonstrating ampere class current at about 20 MeV. A Zeroth Order Design Report is in an advanced draft state, and can be found on the web at http://www.agsrhichome.bnl.gov/eCool/

  16. Electron Cooling of RHIC

    Energy Technology Data Exchange (ETDEWEB)

    I. Ben-Zvi; D.S. Barton; D.B. Beavis; M. Blaskiewicz; J.M. Brennan; A. Burrill; R. Calaga; P. Cameron; X.Y. Chang; R. Connolly; Yu.I. Eidelman; A.V. Fedotov; W. Fischer; D.M. Gassner; H. Hahn; M. Harrison; A. Hershcovitch; H.-C. Hseuh; A.K. Jain; P.D.J. Johnson; D. Kayran; J. Kewisch; R.F. Lambiase; V. Litvinenko; W.W. MacKay; G.J. Mahler; N. Malitsky; G.T. McIntyre; W. Meng; K.A.M. Mirabella; C. Montag; T.C.N. Nehring; T. Nicoletti; B. Oerter; G. Parzen; D. Pate; J. Rank; T. Rao; T. Roser; T. Russo; J. Scaduto; K. Smith; D. Trbojevic; G. Wang; J. Wei; N.W.W. Williams; K.-C. Wu; V. Yakimenko; A. Zaltsman; Y. Zhao; D.T. Abell; D.L. Bruhwiler; H. Bluem; A. Burger; M.D. Cole; A.J. Favale; D. Holmes; J. Rathke; T. Schultheiss; A.M.M. Todd; A.V. Burov; S. Nagaitsev; J.R. Delayen; Y.S. Derbenev; L. W. Funk; P. Kneisel; L. Merminga; H.L. Phillips; J.P. Preble; I. Koop; V.V. Parkhomchuk; Y.M. Shatunov; A.N. Skrinsky; I. Koop; V.V. Parkhomchuk; Y.M. Shatunov; A.N. Skrinsky; J.S. Sekutowicz

    2005-05-16

    We report progress on the R&D program for electron-cooling of the Relativistic Heavy Ion Collider (RHIC). This electron cooler is designed to cool 100 GeV/nucleon at storage energy using 54 MeV electrons. The electron source will be a superconducting RF photocathode gun. The accelerator will be a superconducting energy recovery linac. The frequency of the accelerator is set at 703.75 MHz. The maximum electron bunch frequency is 9.38 MHz, with bunch charge of 20 nC. The R&D program has the following components: The photoinjector and its photocathode, the superconducting linac cavity, start-to-end beam dynamics with magnetized electrons, electron cooling calculations including benchmarking experiments and development of a large superconducting solenoid. The photoinjector and linac cavity are being incorporated into an energy recovery linac aimed at demonstrating ampere class current at about 20 MeV. A Zeroth Order Design Report is in an advanced draft state, and can be found on the web at http://www.agsrhichome.bnl.gov/eCool/.

  17. Performance Evaluation of the Concept of Hybrid Heat Pipe as Passive In-core Cooling Systems for Advanced Nuclear Power Plant

    International Nuclear Information System (INIS)

    Jeong, Yeong Shin; Kim, Kyung Mo; Kim, In Guk; Bang, In Cheol

    2015-01-01

    As an arising issue for inherent safety of nuclear power plant, the concept of hybrid heat pipe as passive in-core cooling systems was introduced. Hybrid heat pipe has unique features that it is inserted in core directly to remove decay heat from nuclear fuel without any changes of structures of existing facilities of nuclear power plant, substituting conventional control rod. Hybrid heat pipe consists of metal cladding, working fluid, wick structure, and neutron absorber. Same with working principle of the heat pipe, heat is transported by phase change of working fluid inside metal cask. Figure 1 shows the systematic design of the hybrid heat pipe cooling system. In this study, the concept of a hybrid heat pipe was introduced as a Passive IN-core Cooling Systems (PINCs) and demonstrated for internal design features of heat pipe containing neutron absorber. Using a commercial CFD code, single hybrid heat pipe model was analyzed to evaluate thermal performance in designated operating condition. Also, 1-dimensional reactor transient analysis was done by calculating temperature change of the coolant inside reactor pressure vessel using MATLAB. As a passive decay heat removal device, hybrid heat pipe was suggested with a concept of combination of heat pipe and control rod. Hybrid heat pipe has distinct feature that it can be a unique solution to cool the reactor when depressurization process is impossible so that refueling water cannot be injected into RPV by conventional ECCS. It contains neutron absorber material inside heat pipe, so it can stop the reactor and at the same time, remove decay heat in core. For evaluating the concept of hybrid heat pipe, its thermal performance was analyzed using CFD and one-dimensional transient analysis. From single hybrid heat pipe simulation, the hybrid heat pipe can transport heat from the core inside to outside about 18.20 kW, and total thermal resistance of hybrid heat pipe is 0.015 .deg. C/W. Due to unique features of long heat

  18. Cooling Grapple System for FMEF hot cell

    International Nuclear Information System (INIS)

    Semmens, L.S.; Frandsen, G.B.; Tome, R.

    1983-01-01

    A Cooling Grapple System was designed and built to handle fuel assemblies within the FMEF hot cell. The variety of functions for which it is designed makes it unique from grapples presently in use. The Cooling Grapple can positively grip and transport assemblies vertically, retrieve assemblies from molten sodium where six inches of grapple tip is submerged, cool 7 kw assemblies in argon, and service an in-cell area of 372 m 2 (4000 ft 2 ). Novel and improved operating and maintenance features were incorporated in the design including a shear pin and mechanical catcher system to prevent overloading the grapple while allowing additional reaction time for crane shutdown

  19. ATLAS - Liquid Cooling Systems

    CERN Multimedia

    Bonneau, P.

    1998-01-01

    Photo 1 - Cooling Unit - Side View Photo 2 - Cooling Unit - Detail Manifolds Photo 3 - Cooling Unit - Rear View Photo 4 - Cooling Unit - Detail Pump, Heater and Exchanger Photo 5 - Cooling Unit - Detail Pump and Fridge Photo 6 - Cooling Unit - Front View

  20. Laser cooling, evaporative cooling and Bose-Einstein condensation

    International Nuclear Information System (INIS)

    Ghosh, Pradip N.

    2002-01-01

    Laser radiations are used to slow down atoms by the process of momentum transfer. This leads to reducing the temperature to micro kelvin region. Gas phase atoms are trapped by using magnetic fields. The recent advances have led to the realization of the dream of physicists of confining the atoms and reducing their velocities to the limit imposed by quantum mechanics. A number of new experiments are possible with the cooled and trapped atoms and ions that would be useful to solve many problems of theoretical physics. Further cooling by the process of evaporative technique has led to the observation of Bose-Einstein Condensation predicted by Einstein and Bose nearly seventy-five years ago. A brief review of the method of laser cooling, magnetic trapping and evaporative cooling methods used for obtaining ultracold atoms are discussed. It is possible to obtain temperature in the nano kelvin region without using cryogenic methods thus simplifying the experimental methods to a great extent. (author)

  1. Test Your Sodium Smarts

    Science.gov (United States)

    ... You may be surprised to learn how much sodium is in many foods. Sodium, including sodium chloride ... foods with little or no salt. Test your sodium smarts by answering these 10 questions about which ...

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

    International Nuclear Information System (INIS)

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

    1980-08-01

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

  3. Ten years of sodium cooled steam generator tests on the C.G.V.S. Synthesis of the results obtained on these equipments and operation experiments of an industrial size test facility

    International Nuclear Information System (INIS)

    Fontaine, J.P.; Llory, M.; Quinet, J.L.

    1984-04-01

    From 1970 to 1980, Electricite de France carried out tests on four steam generators of the fast neutron reactor series on an industrial size testing equipment, the C.G.V.S. (large power testing Circuit for Steam Generators heated by Sodium). After a presentation of the testing installation, types of tests carried out and tested apparatus, a balance of lessons drawn from the circuit exploitation, and from the main results obtained on the tested equipments and on the means of calculation COPI and SICLE codes developed or adopted to simulate steam generator operation. 33 figs., 50 refs [fr

  4. Avances en la reducción del consumo de sal y sodio en Costa Rica Advances in reducing salt and sodium intake in Costa Rica

    Directory of Open Access Journals (Sweden)

    Adriana Blanco-Metzler

    2012-10-01

    Full Text Available En el presente artículo se describen los avances logrados en Costa Rica -así como los desafíos y limitaciones- en la reducción del consumo de sal. El establecimiento del Plan Nacional para la Reducción del Consumo de Sal/sodio en la Población de Costa Rica 2011 - 2021 se complementó con programas y proyectos multisectoriales específicos dirigidos a: 1 conocer la ingesta de sodio y el contenido de sal o sodio en los alimentos de mayor consumo; identificar los conocimientos, actitudes y comportamientos del consumidor respecto a la sal/sodio, su relación con la salud y el etiquetado nutricio-nal; evaluar la relación costo-efectividad de las medidas dirigidas a reducir la prevalencia de hipertensión arterial; 2 implementar estrategias para disminuir el contenido de sal/sodio en los alimentos procesados y los preparados en casa; 3 promover cambios de conducta en la población para reducir el consumo de sal en la alimentación; y 4 monitorear y evaluar las acciones dirigidas a reducir el consumo de sal o sodio en la población. Para alcanzar las metas propuestas se debe lograr una exitosa coordinación interinstitucional con los actores estratégicos, negociar compromisos con la industria alimentaria y los servicios de alimentación, y mejorar la regulación de los nutrientes críticos asociados con las enfermedades crónicas no transmisibles, en los alimentos. Se espera que a partir de los avances logrados durante la ejecución del Plan Nacional, Costa Rica logre alcanzar la meta internacional de reducción del consumo de sal.This article describes the progress-as well as the challenges and limitations-in reducing salt intake in Costa Rica. The National Plan to Reduce Public Consumption of Salt/Sodium in Costa Rica 2011 - 2021 was complemented with multisectoral programs and projects specifically designed to: 1 determine sodium intake and the salt/sodium content of the most widely consumed foods; identify the consumer knowledge, attitudes

  5. Cool Snacks

    DEFF Research Database (Denmark)

    Krogager, Stinne Gunder Strøm; Grunert, Klaus G; Brunsø, Karen

    2016-01-01

    Young people snack and their snacking habits are not always healthy. We address the questions whether it is possible to develop a new snack product that adolescents will find attractive, even though it is based on ingredients as healthy as fruits and vegetables, and we argue that developing...... such a product requires an interdisciplinary effort where researchers with backgrounds in psychology, anthropology, media science, philosophy, sensory science and food science join forces. We present the COOL SNACKS project, where such a blend of competences was used first to obtain thorough insight into young...... people's snacking behaviour and then to develop and test new, healthier snacking solutions. These new snacking solutions were tested and found to be favourably accepted by young people. The paper therefore provides a proof of principle that the development of snacks that are both healthy and attractive...

  6. Cool visitors

    CERN Multimedia

    2006-01-01

    Pictured, from left to right: Tim Izo (saxophone, flute, guitar), Bobby Grant (tour manager), George Pajon (guitar). What do the LHC and a world-famous hip-hop group have in common? They are cool! On Saturday, 1st July, before their appearance at the Montreux Jazz Festival, three members of the 'Black Eyed Peas' came on a surprise visit to CERN, inspired by Dan Brown's Angels and Demons. At short notice, Connie Potter (Head of the ATLAS secretariat) organized a guided tour of ATLAS and the AD 'antimatter factory'. Still curious, lead vocalist Will.I.Am met CERN physicist Rolf Landua after the concert to ask many more questions on particles, CERN, and the origin of the Universe.

  7. Overflow type sodium sampler for FBTR circuits

    International Nuclear Information System (INIS)

    Muralidaran, P.; Ganesan, V.; Chandran, K.; Periaswami, G.

    1996-01-01

    Obtaining a representative sample is crucial for getting reliable results in sodium analysis. Sampling liquid sodium reliability is complicated since impurities segregate while cooling. Selective sorption of certain elements calls for use of different crucible materials for various sodium impurities. Sampling methods currently in use such as flow through sampling and dip sampling are not the proper methods as they can not take care of the above problems. An overflow type sampler where the entire sample contained in a crucible can be used for analysis thus obviating problems due to segregation has been developed for use in Fast Breeder Test Reactor (FBTR). This report describes the construction and operation of this sampler. (author)

  8. Objective Provision Tree (OPT) in sodium cooled fast reactors; Objective Provision Tree (OPT) en reactores rapidos refrigerados por sodio. Aplicacion a la funcion de seguridad de evacuacion de calor residual

    Energy Technology Data Exchange (ETDEWEB)

    Queral, C.; Montero-Mayorga, J.; Gonzalez-Cadelo, J.

    2013-07-01

    Application to the safety function of residual heat removal As part of the project {sup S}afety Assessment for Reactor of GEN-IV (SARGEN IV) has been implemented the methodology ISAM from the IAEA to the safety assessment of new sodium reactor designs. Within the ISAM, a new tool to facilitate this assessment is the Objective Provision Tree (OPT) which documents the provisions necessary for each of the levels of defense in depth, as well as for each critical function of security. Due to the design innovations that have sodium reactors, the evaluation of safety and licensing of these reactors requires special considerations. In this work we have analyzed the mechanisms of failure of the safety function concerning the evacuation of waste heat, and have been proposed different provisions for each of the first three levels of defense in depth. The main result of this work is reflected in the elaboration of the OPTs, one for each of the first three levels of defense in depth for the safety of evacuation of residual heat function. These trees represent in a schematic way the provisions necessary to comply with the objectives of each level which are respectively: 1) deviations from normal operation, 2) control of abnormal operation and fault detection and 3) incidental control.

  9. Unconventional liquid metal cooled fast reactors

    International Nuclear Information System (INIS)

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

    1989-06-01

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

  10. Electron Cooling of RHIC

    CERN Document Server

    Ben-Zvi, Ilan; Barton, Donald; Beavis, Dana; Blaskiewicz, Michael; Bluem, Hans; Brennan, Joseph M; Bruhwiler, David L; Burger, Al; Burov, Alexey; Burrill, Andrew; Calaga, Rama; Cameron, Peter; Chang, Xiangyun; Cole, Michael; Connolly, Roger; Delayen, Jean R; Derbenev, Yaroslav S; Eidelman, Yury I; Favale, Anthony; Fedotov, Alexei V; Fischer, Wolfram; Funk, L W; Gassner, David M; Hahn, Harald; Harrison, Michael; Hershcovitch, Ady; Holmes, Douglas; Hseuh Hsiao Chaun; Johnson, Peter; Kayran, Dmitry; Kewisch, Jorg; Kneisel, Peter; Koop, Ivan; Lambiase, Robert; Litvinenko, Vladimir N; MacKay, William W; Mahler, George; Malitsky, Nikolay; McIntyre, Gary; Meng, Wuzheng; Merminga, Lia; Meshkov, Igor; Mirabella, Kerry; Montag, Christoph; Nagaitsev, Sergei; Nehring, Thomas; Nicoletti, Tony; Oerter, Brian; Parkhomchuk, Vasily; Parzen, George; Pate, David; Phillips, Larry; Preble, Joseph P; Rank, Jim; Rao, Triveni; Rathke, John; Roser, Thomas; Russo, Thomas; Scaduto, Joseph; Schultheiss, Tom; Sekutowicz, Jacek; Shatunov, Yuri; Sidorin, Anatoly O; Skrinsky, Aleksander Nikolayevich; Smirnov, Alexander V; Smith, Kevin T; Todd, Alan M M; Trbojevic, Dejan; Troubnikov, Grigory; Wang, Gang; Wei, Jie; Williams, Neville; Wu, Kuo-Chen; Yakimenko, Vitaly; Zaltsman, Alex; Zhao, Yongxiang; ain, Animesh K

    2005-01-01

    We report progress on the R&D program for electron-cooling of the Relativistic Heavy Ion Collider (RHIC). This electron cooler is designed to cool 100 GeV/nucleon at storage energy using 54 MeV electrons. The electron source will be a superconducting RF photocathode gun. The accelerator will be a superconducting energy recovery linac. The frequency of the accelerator is set at 703.75 MHz. The maximum electron bunch frequency is 9.38 MHz, with bunch charge of 20 nC. The R&D program has the following components: The photoinjector and its photocathode, the superconducting linac cavity, start-to-end beam dynamics with magnetized electrons, electron cooling calculations including benchmarking experiments and development of a large superconducting solenoid. The photoinjector and linac cavity are being incorporated into an energy recovery linac aimed at demonstrating ampere class current at about 20 MeV. A Zeroth Order Design Report is in an advanced draft state, and can be found on the web at http://www.ags...

  11. Effectiveness of Chlorinated Water, Sodium Hypochlorite, Sodium ...

    African Journals Online (AJOL)

    This study evaluated the efficacy of chlorinated water, sodium hypochlorite solution, sodium chloride solution and sterile distilled water in eliminating pathogenic bacteria on the surfaces of raw vegetables. Lettuce vegetables were dipped in different concentrations of chlorinated water, sodium hypochlorite solution, sodium ...

  12. Atmospheric dispersion of sodium aerosol due to a sodium leak in a fast breeder reactor complex

    International Nuclear Information System (INIS)

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

    2008-01-01

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

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

    International Nuclear Information System (INIS)

    2013-05-01

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

  14. WORKSHOP: Beam cooling

    International Nuclear Information System (INIS)

    Anon.

    1994-01-01

    Cooling - the control of unruly particles to provide well-behaved beams - has become a major new tool in accelerator physics. The main approaches of electron cooling pioneered by Gersh Budker at Novosibirsk and stochastic cooling by Simon van der Meer at CERN, are now complemented by additional ideas, such as laser cooling of ions and ionization cooling of muons

  15. Compositional variations for small-scale gamma prime (γ′) precipitates formed at different cooling rates in an advanced Ni-based superalloy

    International Nuclear Information System (INIS)

    Chen, Y.Q.; Francis, E.; Robson, J.; Preuss, M.; Haigh, S.J.

    2015-01-01

    Size-dependent compositional variations under different cooling regimes have been investigated for ordered L1 2 -structured gamma prime (γ′) precipitates in the commercial powder metallurgy Ni-based superalloy RR1000. Using scanning transmission electron microscope imaging combined with absorption-corrected energy-dispersive X-ray spectroscopy, we have discovered large differences in the Al, Ti and Co compositions for γ′ precipitates in the size range 10–300 nm. Our experimental results, coupled with complementary thermodynamic calculations, demonstrate the importance of kinetic factors on precipitate composition in Ni-based superalloys. In particular, these results provide new evidence for the role of elemental diffusion kinetics and aluminium antisite atoms on the low-temperature growth kinetics of fine-scale γ′ precipitates. Our findings have important implications for understanding the microstructure and precipitation behaviour of Ni-based superalloys, suggesting a transition in the mechanism of vacancy-mediated diffusion of Al from intrasublattice exchange at high temperatures to intersublattice antisite-assisted exchange at low temperatures

  16. Amorphous MoS 3 Infiltrated with Carbon Nanotubes as an Advanced Anode Material of Sodium-Ion Batteries with Large Gravimetric, Areal, and Volumetric Capacities

    Energy Technology Data Exchange (ETDEWEB)

    Ye, Hualin [Institute of Functional Nano and Soft Materials (FUNSOM), Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou 215123 China; Wang, Lu [Institute of Functional Nano and Soft Materials (FUNSOM), Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou 215123 China; Deng, Shuo [Institute of Functional Nano and Soft Materials (FUNSOM), Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou 215123 China; Zeng, Xiaoqiao [Chemical Sciences and Engineering Division, Argonne National Laboratory, Lemont IL 60439 USA; Nie, Kaiqi [Institute of Functional Nano and Soft Materials (FUNSOM), Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou 215123 China; Duchesne, Paul N. [Department of Chemistry, Dalhousie University, Halifax NS B3H 4R2 Canada; Wang, Bo [Institute of Functional Nano and Soft Materials (FUNSOM), Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou 215123 China; Liu, Simon [Department of Chemical Engineering, University of Waterloo, Ontario N2L 3G1 Canada; Zhou, Junhua [Institute of Functional Nano and Soft Materials (FUNSOM), Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou 215123 China; Zhao, Feipeng [Institute of Functional Nano and Soft Materials (FUNSOM), Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou 215123 China; Han, Na [Institute of Functional Nano and Soft Materials (FUNSOM), Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou 215123 China; Zhang, Peng [Department of Chemistry, Dalhousie University, Halifax NS B3H 4R2 Canada; Zhong, Jun [Institute of Functional Nano and Soft Materials (FUNSOM), Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou 215123 China; Sun, Xuhui [Institute of Functional Nano and Soft Materials (FUNSOM), Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou 215123 China; Li, Youyong [Institute of Functional Nano and Soft Materials (FUNSOM), Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou 215123 China; Li, Yanguang [Institute of Functional Nano and Soft Materials (FUNSOM), Jiangsu Key Laboratory for Carbon-Based Functional Materials and Devices, Soochow University, Suzhou 215123 China; Lu, Jun [Chemical Sciences and Engineering Division, Argonne National Laboratory, Lemont IL 60439 USA

    2016-11-17

    The search for earth-abundant and high-performance electrode materials for sodium-ion batteries represents an important challenge to current battery research. 2D transition metal dichalcogenides, particularly MoS2, have attracted increasing attention recently, but few of them so far have been able to meet expectations. In this study, it is demonstrated that another phase of molybdenum sulfide—amorphous chain-like MoS3—can be a better choice as the anode material of sodium-ion batteries. Highly compact MoS3 particles infiltrated with carbon nanotubes are prepared via the facile acid precipitation method in ethylene glycol. Compared to crystalline MoS2, the resultant amorphous MoS3 not only exhibits impressive gravimetric performance—featuring excellent specific capacity (≈615 mA h g−1), rate capability (235 mA h g−1 at 20 A g−1), and cycling stability but also shows exceptional volumetric capacity of ≈1000 mA h cm−3 and an areal capacity of >6.0 mA h cm−2 at very high areal loadings of active materials (up to 12 mg cm−2). The experimental results are supported by density functional theory simulations showing that the 1D chains of MoS3 can facilitate the adsorption and diffusion of Na+ ions. At last, it is demonstrated that the MoS3 anode can be paired with an Na3V2(PO4)3 cathode to afford full cells with great capacity and cycling performance.

  17. The collaborative project on European sodium fast reactor (CP ESFR project)

    International Nuclear Information System (INIS)

    Fiorini, Gian-Luigi

    2010-01-01

    The paper summarizes the key characteristics of the four years large Collaborative Project on European Sodium Fast Reactor (CP ESFR - 2009-2012); the CP ESFR follows the 6th FP project named 'Roadmap for a European Innovative SOdium cooled FAst Reactor - EISOFAR' further identifying, organizing and implementing a significant part of the needed R and D effort. The paper also gives insights concerning the so called 'working horses' cores and systems which are provided by CEA and AREVA and that will be used as a basis to test the performances and assess the pertinence of innovative solutions. The CP ESFR merges the contribution of 25 European partners (EU + CH); it will be performed under the aegis of the 7th Euratom FP under the Area - Advanced Nuclear Systems with a refund from the European Commission. It will be a key component of the European Sustainable Nuclear Energy Technology Platform (SNE TP) and its Strategic Research Agenda (SRA). The inputs for the project are the key research goals for fourth generation of European sodium cooled fast reactors which can be summarized as follows: an improved safety with in particular the achievement of a robust architecture vis-a-vis of abnormal situations and the robustness of the safety demonstrations; the guarantee of a financial risk similar to that of the other means of energy production; a flexible and robust management of nuclear materials and especially waste reduction through Minor Actinides burning

  18. European commission - 7th framework programme. The collaborative project on European sodium fast reactor (CP ESFR)

    International Nuclear Information System (INIS)

    Fiorini, G.L.

    2009-01-01

    The paper summarizes the key characteristics of the four years large Collaborative Project on European Sodium Fast Reactor (CP ESFR - 2009-2012); the CP ESFR follows the 6th FP project named 'Roadmap for a European Innovative SOdium cooled FAst Reactor - EISOFAR' further identifying, organizing and implementing a significant part of the needed R and D effort. The CP ESFR merges the contribution of 25 european partners; it will be realized under the aegis of the 7th FP under the Area - Advanced Nuclear Systems with a refund from the European Commission of 5.8 M euro (11.55 M euro total budget). It will be a key component of the European Sustainable Nuclear Energy Technology Platform (SNE TP) and its Strategic Research Agenda (SRA). The inputs for the project are the key research goals for fourth generation of European sodium cooled fast reactors which can be summarized as follow: an improved safety with in particular the achievement of a robust architecture vis a vis of abnormal situations and the robustness of the safety demonstrations; the guarantee of a financial risk comparable to that of the other means of energy production; a flexible and robust management of the nuclear materials and especially the waste reduction through the Minor Actinides burning. (author)

  19. Renewable Heating And Cooling

    Science.gov (United States)

    Renewable heating and cooling is a set of alternative resources and technologies that can be used in place of conventional heating and cooling technologies for common applications such as water heating, space heating, space cooling and process heat.

  20. An investigation of sodium iodide solubility in sodium-stainless steel systems

    International Nuclear Information System (INIS)

    Sagawa, Norihiko; Tashiro, Suguru

    1996-01-01

    Sodium iodide and major constituents of stainless steel in sodium are determined by using the steel capsules to obtain a better understanding on contribution of the constituents to the apparent iodide solubility in sodium. The capsule loaded with 20 g sodium and 0.1 - 0.3 g powder of sodium iodide is heated at its upper part in a furnace and cooled at its bottom on brass plates to establish a large temperature gradient along the capsule tube. After a given period of equilibration, the iodide and constituents are fixed in solidified sodium by quick quenching of the capsules. Sodium samples are taken from the sectioned capsule tube and submitted to sodium dissolution by vaporized water for determination of the iodine and to vacuum distillation for determination of the metal elements. Iron and nickel concentrations are observed to be lower in the samples at higher iodine concentrations. Chromium and manganese concentrations are seen to be insensitive to the iodine concentrations. The observations can be interpreted by a model that sodium oxide combines with metal iodide in sodium to form a complex compound and with consideration that the compound will fall and deposit onto the bottom of the capsule by thermal diffusion. (author)

  1. Restaurant food cooling practices.

    Science.gov (United States)

    Brown, Laura Green; Ripley, Danny; Blade, Henry; Reimann, Dave; Everstine, Karen; Nicholas, Dave; Egan, Jessica; Koktavy, Nicole; Quilliam, Daniela N

    2012-12-01

    Improper food cooling practices are a significant cause of foodborne illness, yet little is known about restaurant food cooling practices. This study was conducted to examine food cooling practices in restaurants. Specifically, the study assesses the frequency with which restaurants meet U.S. Food and Drug Administration (FDA) recommendations aimed at reducing pathogen proliferation during food cooling. Members of the Centers for Disease Control and Prevention's Environmental Health Specialists Network collected data on food cooling practices in 420 restaurants. The data collected indicate that many restaurants are not meeting FDA recommendations concerning cooling. Although most restaurant kitchen managers report that they have formal cooling processes (86%) and provide training to food workers on proper cooling (91%), many managers said that they do not have tested and verified cooling processes (39%), do not monitor time or temperature during cooling processes (41%), or do not calibrate thermometers used for monitoring temperatures (15%). Indeed, 86% of managers reported cooling processes that did not incorporate all FDA-recommended components. Additionally, restaurants do not always follow recommendations concerning specific cooling methods, such as refrigerating cooling food at shallow depths, ventilating cooling food, providing open-air space around the tops and sides of cooling food containers, and refraining from stacking cooling food containers on top of each other. Data from this study could be used by food safety programs and the restaurant industry to target training and intervention efforts concerning cooling practices. These efforts should focus on the most frequent poor cooling practices, as identified by this study.

  2. Fractional excretion of sodium

    Science.gov (United States)

    FE sodium; FENa ... a lab. There, they are examined for salt (sodium) and creatinine levels. Creatinine is a chemical waste ... Chernecky CC, Berger BJ. Excretion fraction of filtered sodium-blood and urine. In: Chernecky CC, Berger BJ, ...

  3. Sodium and Food Sources

    Science.gov (United States)

    ... Disease Cholesterol High Blood Pressure Million Hearts® WISEWOMAN Sodium and Food Sources Recommend on Facebook Tweet Share ... food [PDF-867K] and how to reduce sodium. Sodium Reduction Is Challenging Types of food matter: More ...

  4. Innovative technologies for Faraday shield cooling

    International Nuclear Information System (INIS)

    Rosenfeld, J.H.; Lindemuth, J.E.; North, M.T.; Goulding, R.H.

    1995-01-01

    Alternative advanced technologies are being evaluated for use in cooling the Faraday shields used for protection of ion cyclotron range of frequencies (ICR) antennae in Tokamaks. Two approaches currently under evaluation include heat pipe cooling and gas cooling. A Monel/water heat pipe cooled Faraday shield has been successfully demonstrated. Heat pipe cooling offers the advantage of reducing the amount of water discharged into the Tokamak in the event of a tube weld failure. The device was recently tested on an antenna at Oak Ridge National Laboratory. The heat pipe design uses inclined water heat pipes with warm water condensers located outside of the plasma chamber. This approach can passively remove absorbed heat fluxes in excess of 200 W/cm 2 ;. Helium-cooled Faraday shields are also being evaluated. This approach offers the advantage of no liquid discharge into the Tokamak in the event of a tube failure. Innovative internal cooling structures based on porous metal cooling are being used to develop a helium-cooled Faraday shield structure. This approach can dissipate the high heat fluxes typical of Faraday shield applications while minimizing the required helium blower power. Preliminary analysis shows that nominal helium flow and pressure drop can sufficiently cool a Faraday shield in typical applications. Plans are in progress to fabricate and test prototype hardware based on this approach

  5. Process fluid cooling system

    International Nuclear Information System (INIS)

    Farquhar, N.G.; Schwab, J.A.

    1977-01-01

    A system of heat exchangers is disclosed for cooling process fluids. The system is particularly applicable to cooling steam generator blowdown fluid in a nuclear plant prior to chemical purification of the fluid in which it minimizes the potential of boiling of the plant cooling water which cools the blowdown fluid

  6. Production of Depleted UO2Kernels for the Advanced Gas-Cooled Reactor Program for Use in TRISO Coating Development

    Energy Technology Data Exchange (ETDEWEB)

    Collins, J.L.

    2004-12-02

    The main objective of the Depleted UO{sub 2} Kernels Production Task at Oak Ridge National Laboratory (ORNL) was to conduct two small-scale production campaigns to produce 2 kg of UO{sub 2} kernels with diameters of 500 {+-} 20 {micro}m and 3.5 kg of UO{sub 2} kernels with diameters of 350 {+-} 10 {micro}m for the U.S. Department of Energy Advanced Fuel Cycle Initiative Program. The final acceptance requirements for the UO{sub 2} kernels are provided in the first section of this report. The kernels were prepared for use by the ORNL Metals and Ceramics Division in a development study to perfect the triisotropic (TRISO) coating process. It was important that the kernels be strong and near theoretical density, with excellent sphericity, minimal surface roughness, and no cracking. This report gives a detailed description of the production efforts and results as well as an in-depth description of the internal gelation process and its chemistry. It describes the laboratory-scale gel-forming apparatus, optimum broth formulation and operating conditions, preparation of the acid-deficient uranyl nitrate stock solution, the system used to provide uniform broth droplet formation and control, and the process of calcining and sintering UO{sub 3} {center_dot} 2H{sub 2}O microspheres to form dense UO{sub 2} kernels. The report also describes improvements and best past practices for uranium kernel formation via the internal gelation process, which utilizes hexamethylenetetramine and urea. Improvements were made in broth formulation and broth droplet formation and control that made it possible in many of the runs in the campaign to produce the desired 350 {+-} 10-{micro}m-diameter kernels, and to obtain very high yields.

  7. Design and Fabrication of Serpentine Tube Type Sodium to Air Heat Exchangers for PFBR SGDHR Circuits

    International Nuclear Information System (INIS)

    Pai, Aravinda; Mitra, Tarun Kumar; Loganathan, T.; Kumar, Prabhat

    2015-01-01

    Prototype Fast Breeder Reactor (PFBR) is a 500MWe pool type, sodium cooled nuclear reactor which is in advanced stage of construction by BHAVINI at Kalpakkam, India. The sodium to air heat exchanger (AHX) in the safety grade decay heat removal (SGDHR) loop transfers heat from the intermediate circuit sodium to atmospheric air by natural convection. As the operating temperature is high and AHX forms part of reactor SGDHR system, the boundaries of the AHX must possess a high degree of reliability against failure. This is achieved by comprehensive design, precise material selection, high standard quality control and quality assurance during manufacturing. The major material of construction of AHX is Modified 9Cr-1Mo steel (N&T). AHX has straight sodium inlet and outlet headers with 116 nos. of hot formed pullouts on its surface having OD38.1mmX2.6mm thickness which are connected to a finned tube bundle. Due to the many hot formed pullouts on the headers and complex geometry, the welding sequence, heat treatment and non-destructive examinations of the welds is an extremely difficult and challenging task. Various special tooling and fixtures were designed and developed for the manufacture of the AHXs. This paper provides details of the salient design aspects, challenges, innovations and success achieved during hot forming, welding and fabrication activities of the serpentine tube type AHXs for 500MWe Prototype Fast Breeder Reactor. (author)

  8. Hybrid radiator cooling system

    Science.gov (United States)

    France, David M.; Smith, David S.; Yu, Wenhua; Routbort, Jules L.

    2016-03-15

    A method and hybrid radiator-cooling apparatus for implementing enhanced radiator-cooling are provided. The hybrid radiator-cooling apparatus includes an air-side finned surface for air cooling; an elongated vertically extending surface extending outwardly from the air-side finned surface on a downstream air-side of the hybrid radiator; and a water supply for selectively providing evaporative cooling with water flow by gravity on the elongated vertically extending surface.

  9. 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; Comparacion y validacion de los resultados del codigo AZNHEX v.1.0 con el codigo MCNP simulando el nucleo de un reactor rapido refrigerado con sodio

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L.; Bastida O, G. E. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Esquivel E, J., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    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)

  10. Experimental study on oxidation and combustion characteristics of sodium droplets

    International Nuclear Information System (INIS)

    Zhang Zhigang; Sun Shubin; Liu Chongchong; Tang Yexin

    2015-01-01

    In the operation of the sodium-cooled fast reactor, the accident caused by the leakage and combustion of liquid sodium is common and frequent. In this paper, the oxidation and combustion characteristics of sodium droplets were studied by carrying out the experiments of the oxidation and combustion under different conditions of initial temperatures (140-370℃) of the sodium droplets and oxygen concentrations (4%-21%). The oxidation and combustion behaviors were visualized by a set of combustion apparatus of sodium droplet and a high speed camera. The experiment results show that the columnar oxides grow longer as the initial temperature of sodium droplet and oxygen concentration become lower. Under the same oxygen concentration condition, the sodium droplet with the higher initial temperature is easier to ignite and burn. When the initial temperature of sodium droplet is below 200℃, it is very difficult to ignite. If there is a turbulence damaging the oxide layer on the surface, the sodium droplet will also burn gradually. When the initial temperature ranges from 140℃ to 370℃ and the oxygen fraction is equal to or higher than 12%, the sodium droplet could burn completely and the maximum combustion temperature could roughly reach 600-800℃. When the oxygen concentration is below 12%, the sodium droplet could not burn completely and the highest combustion temperature is below 600℃. The results are helpful to the research on the columnar flow and spray sodium fire. (authors)

  11. Variable Conductance Heat Pipe Cooling of Stirling Convertor and General Purpose Heat Source

    Science.gov (United States)

    Tarau, Calin; Schwendeman, Carl; Anderson, William G.; Cornell, Peggy A.; Schifer, Nicholas A.

    2013-01-01

    In a Stirling Radioisotope Power System (RPS), heat must be continuously removed from the General Purpose Heat Source (GPHS) modules to maintain the modules and surrounding insulation at acceptable temperatures. The Stirling convertor normally provides this cooling. If the Stirling convertor stops in the current system, the insulation is designed to spoil, preventing damage to the GPHS at the cost of an early termination of the mission. An alkali-metal Variable Conductance Heat Pipe (VCHP) can be used to passively allow multiple stops and restarts of the Stirling convertor. In a previous NASA SBIR Program, Advanced Cooling Technologies, Inc. (ACT) developed a series of sodium VCHPs as backup cooling systems for Stirling RPS. The operation of these VCHPs was demonstrated using Stirling heater head simulators and GPHS simulators. In the most recent effort, a sodium VCHP with a stainless steel envelope was designed, fabricated and tested at NASA Glenn Research Center (GRC) with a Stirling convertor for two concepts; one for the Advanced Stirling Radioisotope Generator (ASRG) back up cooling system and one for the Long-lived Venus Lander thermal management system. The VCHP is designed to activate and remove heat from the stopped convertor at a 19 degC temperature increase from the nominal vapor temperature. The 19 degC temperature increase from nominal is low enough to avoid risking standard ASRG operation and spoiling of the Multi-Layer Insulation (MLI). In addition, the same backup cooling system can be applied to the Stirling convertor used for the refrigeration system of the Long-lived Venus Lander. The VCHP will allow the refrigeration system to: 1) rest during transit at a lower temperature than nominal; 2) pre-cool the modules to an even lower temperature before the entry in Venus atmosphere; 3) work at nominal temperature on Venus surface; 4) briefly stop multiple times on the Venus surface to allow scientific measurements. This paper presents the experimental

  12. Liquid sodium pool fires

    International Nuclear Information System (INIS)

    Casselman, C.

    1979-01-01

    Experimental sodium pool combustion results have led to a definition of the combustion kinetics, and have revealed the hazards of sodium-concrete contact reactions and the possible ignition of organic matter (paint) by hydration of sodium peroxide aerosols. Analysis of these test results shows that the controlling mechanism is sodium evaporation diffusion. (author)

  13. Operational experience on sodium deposits in KNK reactor and RSB test facility

    International Nuclear Information System (INIS)

    Jansing, W.; Kirchner, G.; Menck, J.

    1977-01-01

    A specific problem of sodium-cooled reactor plants is the formation of sodium aerosols which deposit at cold sections of the plant. Formation and behaviour of sodium aerosols depend on various factors. These may show extreme different effects under conditions which first seem to be identical. Thus, it is very difficult to set up general valid rules on sodium aerosols. By operational experience gained in different plants under divers operating conditions, knowledge is drawn which corresponds well with theoretical considerations. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

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

  15. Recent Development in Turbine Blade Film Cooling

    Directory of Open Access Journals (Sweden)

    Je-Chin Han

    2001-01-01

    Full Text Available Gas turbines are extensively used for aircraft propulsion, land-based power generation, and industrial applications. Thermal efficiency and power output of gas turbines increase with increasing turbine rotor inlet temperature (RIT. The current RIT level in advanced gas turbines is far above the .melting point of the blade material. Therefore, along with high temperature material development, a sophisticated cooling scheme must be developed for continuous safe operation of gas turbines with high performance. Gas turbine blades are cooled internally and externally. This paper focuses on external blade cooling or so-called film cooling. In film cooling, relatively cool air is injected from the inside of the blade to the outside surface which forms a protective layer between the blade surface and hot gas streams. Performance of film cooling primarily depends on the coolant to mainstream pressure ratio, temperature ratio, and film hole location and geometry under representative engine flow conditions. In the past number of years there has been considerable progress in turbine film cooling research and this paper is limited to review a few selected publications to reflect recent development in turbine blade film cooling.

  16. Restaurant Food Cooling Practices†

    Science.gov (United States)

    BROWN, LAURA GREEN; RIPLEY, DANNY; BLADE, HENRY; REIMANN, DAVE; EVERSTINE, KAREN; NICHOLAS, DAVE; EGAN, JESSICA; KOKTAVY, NICOLE; QUILLIAM, DANIELA N.

    2017-01-01

    Improper food cooling practices are a significant cause of foodborne illness, yet little is known about restaurant food cooling practices. This study was conducted to examine food cooling practices in restaurants. Specifically, the study assesses the frequency with which restaurants meet U.S. Food and Drug Administration (FDA) recommendations aimed at reducing pathogen proliferation during food cooling. Members of the Centers for Disease Control and Prevention’s Environmental Health Specialists Network collected data on food cooling practices in 420 restaurants. The data collected indicate that many restaurants are not meeting FDA recommendations concerning cooling. Although most restaurant kitchen managers report that they have formal cooling processes (86%) and provide training to food workers on proper cooling (91%), many managers said that they do not have tested and verified cooling processes (39%), do not monitor time or temperature during cooling processes (41%), or do not calibrate thermometers used for monitoring temperatures (15%). Indeed, 86% of managers reported cooling processes that did not incorporate all FDA-recommended components. Additionally, restaurants do not always follow recommendations concerning specific cooling methods, such as refrigerating cooling food at shallow depths, ventilating cooling food, providing open-air space around the tops and sides of cooling food containers, and refraining from stacking cooling food containers on top of each other. Data from this study could be used by food safety programs and the restaurant industry to target training and intervention efforts concerning cooling practices. These efforts should focus on the most frequent poor cooling practices, as identified by this study. PMID:23212014

  17. Sodium coolant of fast reactors: Experience and problems

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Volchkov, L.G.; Drobyshev, A.V.; Nikulin, M.P.; Kochetkov, L.A.; Alexeev, V.V.

    1997-01-01

    In present report the following subjects are considered: state of the coolant and sodium systems under normal operating condition as well as under decommissioning, disclosing of sodium circuits and liquidation of its consequences, cleaning from sodium and decontamination under repairing works of equipment and circuits. Cleaning of coolant and sodium systems under normal operating conditions and under accident contamination. Cleaning of the equipment under repairing works and during decommissioning from sodium and products of its interaction with water and air. Treatment of sodium waste, taking into account a possibility of sodium fires. It is shown that the state of coolant, cover gas, surfaces of constructive materials which are in contact with them, cleaning systems, formed during installation operation require development of specific technologies. Developed technologies ensured safety operation of sodium cooled installations as in normal operating conditions so in abnormal situations. R and D activities in this field and experience gained provided a solid base for coping with problems arising during decommissioning. Prospective research problems are emphasized where the future efforts should be concentrated in order to improve characteristics of sodium cooled reactors and to make their decommissioning optimal and safe. (author)

  18. An equation of state for sodium

    International Nuclear Information System (INIS)

    Browning, P.

    1981-03-01

    The equation of state (EOS) for sodium which has been employed in assessments of hypothetical accidents in liquid metal cooled fast breeder nuclear reactors has been in use for some years in the British programme. During this time some important experimental reference data, upon which the EOS is based, have been revised. The purpose of this report is primarily to update the sodium EOS by incorporating these revised data. In addition, a number of improvements have been made in the calculational technique used in deriving properties in the single phase. These refinements, which have indicated numerical errors in the earlier EOS output, have improved the precision of the reported data. (author)

  19. Muon cooling channels

    CERN Document Server

    Eberhard-K-Kei

    2003-01-01

    A procedure uses the equations that govern ionization cooling, and leads to the most important parameters of a muon cooling channel that achieves assumed performance parameters. First, purely transverse cooling is considered, followed by both transverse and longitudinal cooling in quadrupole and solenoid channels. Similarities and differences in the results are discussed in detail, and a common notation is developed. Procedure and notation are applied to a few published cooling channels. The parameters of the cooling channels are derived step by step, starting from assumed values of the initial, final and equilibrium emittances, both transverse and longitudinal, the length of the cooling channel, and the material properties of the absorber. The results obtained include cooling lengths and partition numbers, amplitude functions and limits on the dispersion at the absorber, length, aperture and spacing of the absorber, parameters of the RF system that achieve the longitudinal amplitude function and bucket area ...

  20. Solid State Cooling with Advanced Oxide Materials

    Science.gov (United States)

    2014-06-03

    Engineering University of Illinois, Urbana-Champaign Program Overview The focus of this program was to probe electro-( magneto -)caloric materials for...piezoelectric coefficient, the elastic constant, and the coefficient of thermal expansion. We propose that the piezoelectric effect causes a lattice expansion...structure to the epitaxial strain, the polarization and switching characteristics are found to vary with substrate. The elastic constraint from the

  1. Advanced Turbine Blade Cooling Techniques, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Gas turbine engine technology is constantly challenged to operate at higher combustor outlet temperatures. In a modern gas turbine engine, these temperatures can...

  2. Peltier cooling in molecular junctions

    Science.gov (United States)

    Cui, Longji; Miao, Ruijiao; Wang, Kun; Thompson, Dakotah; Zotti, Linda Angela; Cuevas, Juan Carlos; Meyhofer, Edgar; Reddy, Pramod

    2018-02-01

    The study of thermoelectricity in molecular junctions is of fundamental interest for the development of various technologies including cooling (refrigeration) and heat-to-electricity conversion1-4. Recent experimental progress in probing the thermopower (Seebeck effect) of molecular junctions5-9 has enabled studies of the relationship between thermoelectricity and molecular structure10,11. However, observations of Peltier cooling in molecular junctions—a critical step for establishing molecular-based refrigeration—have remained inaccessible. Here, we report direct experimental observations of Peltier cooling in molecular junctions. By integrating conducting-probe atomic force microscopy12,13 with custom-fabricated picowatt-resolution calorimetric microdevices, we created an experimental platform that enables the unified characterization of electrical, thermoelectric and energy dissipation characteristics of molecular junctions. Using this platform, we studied gold junctions with prototypical molecules (Au-biphenyl-4,4'-dithiol-Au, Au-terphenyl-4,4''-dithiol-Au and Au-4,4'-bipyridine-Au) and revealed the relationship between heating or cooling and charge transmission characteristics. Our experimental conclusions are supported by self-energy-corrected density functional theory calculations. We expect these advances to stimulate studies of both thermal and thermoelectric transport in molecular junctions where the possibility of extraordinarily efficient energy conversion has been theoretically predicted2-4,14.

  3. Cooling water distribution system

    Science.gov (United States)

    Orr, Richard

    1994-01-01

    A passive containment cooling system for a nuclear reactor containment vessel. Disclosed is a cooling water distribution system for introducing cooling water by gravity uniformly over the outer surface of a steel containment vessel using an interconnected series of radial guide elements, a plurality of circumferential collector elements and collector boxes to collect and feed the cooling water into distribution channels extending along the curved surface of the steel containment vessel. The cooling water is uniformly distributed over the curved surface by a plurality of weirs in the distribution channels.

  4. Cooling tower calculations

    International Nuclear Information System (INIS)

    Simonkova, J.

    1988-01-01

    The problems are summed up of the dynamic calculation of cooling towers with forced and natural air draft. The quantities and relations are given characterizing the simultaneous exchange of momentum, heat and mass in evaporative water cooling by atmospheric air in the packings of cooling towers. The method of solution is clarified in the calculation of evaporation criteria and thermal characteristics of countercurrent and cross current cooling systems. The procedure is demonstrated of the calculation of cooling towers, and correction curves and the effect assessed of the operating mode at constant air number or constant outlet air volume flow on their course in ventilator cooling towers. In cooling towers with the natural air draft the flow unevenness is assessed of water and air relative to its effect on the resulting cooling efficiency of the towers. The calculation is demonstrated of thermal and resistance response curves and cooling curves of hydraulically unevenly loaded towers owing to the water flow rate parameter graded radially by 20% along the cross-section of the packing. Flow rate unevenness of air due to wind impact on the outlet air flow from the tower significantly affects the temperatures of cooled water in natural air draft cooling towers of a design with lower demands on aerodynamics, as early as at wind velocity of 2 m.s -1 as was demonstrated on a concrete example. (author). 11 figs., 10 refs

  5. Diclofenac sodium overdose

    Science.gov (United States)

    ... page: //medlineplus.gov/ency/article/002630.htm Diclofenac sodium overdose To use the sharing features on this page, please enable JavaScript. Diclofenac sodium is a prescription medicine used to relieve pain ...

  6. Sodium Ferric Gluconate Injection

    Science.gov (United States)

    Sodium ferric gluconate injection is used to treat iron-deficiency anemia (a lower than normal number of ... are also receiving the medication epoetin (Epogen, Procrit). Sodium ferric gluconate injection is in a class of ...

  7. Naproxen sodium overdose

    Science.gov (United States)

    ... page: //medlineplus.gov/ency/article/002507.htm Naproxen sodium overdose To use the sharing features on this page, please enable JavaScript. Naproxen sodium is a nonsteroidal anti-inflammatory drug (NSAID) used ...

  8. Sodium carbonate poisoning

    Science.gov (United States)

    Sodium carbonate (known as washing soda or soda ash) is a chemical found in many household and ... products. This article focuses on poisoning due to sodium carbonate. This article is for information only. Do ...

  9. Sodium hydroxide poisoning

    Science.gov (United States)

    Sodium hydroxide is a very strong chemical. It is also known as lye and caustic soda. This ... poisoning from touching, breathing in (inhaling), or swallowing sodium hydroxide. This article is for information only. Do ...

  10. Docusate Sodium and Pregnancy

    Science.gov (United States)

    ... live chat Live Help Fact Sheets Share Docusate Sodium Friday, 01 April 2016 In every pregnancy, a ... This sheet talks about whether exposure to docusate sodium may increase the risk for birth defects over ...

  11. Sodium pumping: pump problems

    International Nuclear Information System (INIS)

    Guer, M.; Guiton, P.

    Information on sodium pumps for LMFBR type reactors is presented concerning ring pump design, pool reactor pump design, secondary pumps, sodium bearings, swivel joints of the oscillating annulus, and thermal shock loads

  12. ELTA: Citatrademark: Sodium measurement

    International Nuclear Information System (INIS)

    Mauvais, O.

    2002-01-01

    ELTA is pleased to present its last model of Sodium analyzers: CITA 2340: Automatically controlled sodium meter, integrating more automation and performances results respecting costs and wastes reduction. (authors)

  13. Biological behavior of mixed sodium and plutonium oxide aerosols

    International Nuclear Information System (INIS)

    Metivier, H.

    1976-01-01

    New risks from sodium cooled fast breeders are due to solubilization of plutonium dioxide by sodium oxides. The resulting chemical forms of higher valency stage are more transportable than PuO 2 . Bone burden is about 100 times as high as observed with PuO 2 . Diffusion is fast, therapy must be started within 6 h. DTPA is still effective, however chelation efficiency is lower than in the case of Pu IV-DTPA chelation [fr

  14. Study of Polymorphism of Borovanadate Glass of Sodium by Raman ...

    African Journals Online (AJOL)

    Study of Polymorphism of Borovanadate Glass of Sodium by Raman Spectroscopy Low Frequencies. MK Rabia, M Mayoufi, L Grosvalet, B Champagnon. Abstract. Sodium tetraborate (100 – x)(Na2B4O7.10H2O)– xV2O5, (x = 0 to 20 mole %) has been elaborated by splat cooling technique. Raman Measurements on the ...

  15. Liquid sodium technology research

    International Nuclear Information System (INIS)

    Kim, W.C.; Lee, Y.W.; Nam, H.Y.; Chun, S.Y.; Kim, J.; Won, S.Y.

    1982-01-01

    This report describes the technology of impurity control and measurement of liquid sodium, problems associated with material degradation and change of heat transfer characteristics in liquid sodium, and the conceptual design of multipurpose sodium test loop. Discussion and the subsequent analysis are also made with regard to the test results for the sodium-H 2 0 reaction and its effects on the system. (author)

  16. Review of Sodium and Plutonium related Technical Standards in Trans-Uranium Fuel Fabrication Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Misuk; Jeon, Jong Seon; Kang, Hyun Sik; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

    2016-10-15

    In this paper, we would introduce and review technical standards related to sodium fire and plutonium criticality safety. This paper may be helpful to identify considerations in the development of equipment, standards, and etc., to meet the safety requirements in the design, construction and operating of TFFF, KAPF and SFR. The feasibility and conceptual designs are being examined on related facilities, for example, TRU Fuel Fabrication Facilities (TFFF), Korea Advanced Pyro-process Facility (KAPF), and Sodium Cooled Fast Reactor (SFR), in Korea. However, the safety concerns of these facilities have been controversial in part because of the Sodium fire accident and Plutonium related radiation safety caused by transport and handling accident. Thus, many researches have been performed to ensure safety and various documents including safety requirements have been developed. In separating and reducing the long-lived radioactive transuranic(TRU) in the spent nuclear fuel, reusing as the potential energy of uranium fuel resources and reducing the high level wastes, TFFF would be receiving the attention of many people. Thus, people would wonder whether compliance with technical standards that ensures safety. For new facility design, one of the important tasks is to review of technical standards, especially for sodium and Plutonium because of water related highly reactive characteristics and criticality hazard respectively. We have introduced and reviewed two important technical standards for TFFF, which are sodium fire and plutonium criticality safety, in this paper. This paper would provide a brief guidance, about how to start and what is important, to people who are responsible for the initial design to operation of TFFF.

  17. U.S. Department of Energy & Nuclear Regulatory Commission Advanced Fuel Cycle Research & Development Seminar Series FY 2007 & 2008

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

    2008-08-01

    In fiscal year 2007, the Advanced Burner Reactor project initiated an educational seminar series for the Department of Energy (DOE) and Nuclear Regulatory Commission (NRC) personnel on various aspects of fast reactor fuel cycle closure technologies. This important work was initiated to inform DOE and NRC personnel on initial details of sodium-cooled fast reactor, separations, waste form, and safeguard technologies being considered for the Advanced Fuel Cycle Research and Development program, and to learn the important lesson from the licensing process for the Clinch River Breeder Reactor Plant that educating the NRC staff early in the regulatory process is very important and critical to a project success.

  18. Development of a sodium ionization detector for sodium-to-gas leaks

    International Nuclear Information System (INIS)

    Swaminathan, K.; Elumalai, G.

    1984-01-01

    A sensitive sodium-to-gas leak detector has been indigenously developed for use in liquid metal cooled fast breeder reactor. The detector relies on the relative ease with which sodium vapour or its aerosols including its oxides and hydroxides can be thermally ionized compared with other possible constituents such as nitrogen, oxygen, water vapour etc. in a carrier gas and is therefore called sodium ionization detector (SID). The ionization current is a measure of sodium concentration in the carrier gas sampled through the detector. Different sensor designs using platinum and rhodium as filament materials in varying sizes were constructed and their responses to different sodium aerosol concentrations in the carrier gas were investigated. Nitrogen was used as the carrier gas. Both the background current and speed of response were found to depend on the diameter of the filament. There was also a particular collector voltage which yielded maximum sensitivity of the detector. The sensor was therefore optimised considering influence of above factors and a detector has been built which demonstrates a sensitivity better than 0.3 nanogram of sodium per cubic centimetre of carrier gas for a signal to background ratio of 1:1. Its usefulness in detecting sodium fires in experimental area was also demonstrated. Currently efforts are under way to improve the life time of the filament used in the above detector. (author)

  19. Corrosion behaviors of ceramics against liquid sodium. Sodium corrosion characteristics of sintering additives

    International Nuclear Information System (INIS)

    Tachi, Yoshiaki; Kano, Shigeki; Hirakawa, Yasushi; Yoshida, Eiichi

    1998-01-01

    It has been progressed as the Frontier Materials Research to research and develop ceramics to apply for several components of fast breeder reactor using liquid sodium as coolant instead of metallic materials. Grain boundary of ceramics has peculiar properties compared with matrix because most of ceramics are produced by hardening and firing their raw powders. Some previous researchers indicated that ceramics were mainly corroded at grain boundaries by liquid sodium, and ceramics could not be used under corrosive environment. Thus, it is the most important for the usage of ceramics in liquid sodium to improve corrosion resistance of grain boundaries. In order to develop the advanced ceramics having good sodium corrosion resistance among fine ceramics, which have recently been progressed in quality and characteristics remarkably, sodium corrosion behaviors of typical sintering additives such as MgO, Y 2 O 3 and AlN etc. have been examined and evaluated. As a result, the followings have been clarified and some useful knowledge about developing advanced ceramics having good corrosion resistance against liquid sodium has been obtained. (1) Sodium corrosion behavior of MgO depended on Si content. Samples containing large amount of Si were corroded severely by liquid sodium, whereas others with low Si contents showed good corrosion resistance. (2) Both Y 2 O 3 and AlN, which contained little Si, showed good sodium corrosion resistance. (3) MgO, Y 2 O 3 and AlN are thought to be corroded by liquid sodium, if they contain some SiO 2 . Therefore, in order to improve sodium corrosion resistance, it is very important for these ceramics to prevent the contamination of matrix with SiO 2 through purity control of their raw powders. (author)

  20. Example Work Domain Analysis for a Reference Sodium Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hugo, Jacques [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-01-01

    The nuclear industry is currently designing and building a new generation of reactors that will include different structural, functional, and environmental aspects, all of which are likely to have a significant impact on the way these plants are operated. In order to meet economic and safety objectives, these new reactors will all use advanced technologies to some extent, including new materials and advanced digital instrumentation and control systems. New technologies will affect not only operational strategies, but will also require a new approach to how functions are allocated to humans or machines to ensure optimal performance. Uncertainty about the effect of large scale changes in plant design will remain until sound technical bases are developed for new operational concepts and strategies. Up-to-date models and guidance are required for the development of operational concepts for complex socio-technical systems. This report describes how the classical Work Domain Analysis method was adapted to develop operational concept frameworks for new plants. This adaptation of the method is better able to deal with the uncertainty and incomplete information typical of first-of-a-kind designs. Practical examples are provided of the systematic application of the method in the operational analysis of sodium-cooled reactors. Insights from this application and its utility are reviewed and arguments for the formal adoption of Work Domain Analysis as a value-added part of the Systems Engineering process are presented.

  1. Degradation behavior of limestone concrete under limited time sodium exposure

    International Nuclear Information System (INIS)

    Das, S.K.; Sharma, A.K.; Ramesh, S.S.; Parida, F.C.; Kasinathan, N.; Chellapandi, P.

    2009-01-01

    Adequate safety measures are taken during design, fabrication, construction and operation of liquid sodium cooled fast breeder reactor (FBR). However, possibility of sodium leak from secondary heat transport circuits of FBR has not been completely ruled out. In the areas housing sodium pipelines such as Steam Generator Building (SGB), spilled liquid sodium not only reacts with air causing fire but also interacts with structural concrete resulting in its degradation. The structural concrete can be protected from sodium attack using sodium resistant sacrificial concrete layer or steel/refractory liners. Moreover, design and construction of sloping floor with sodium collection pit helps in minimizing the mass of sodium accumulated on the floor and exposure period. Sacrificial concrete layer on the structural concrete should meet key factors like economy, castability, easy removal of affected concrete in the event of a sodium fire and disposability of debris apart from its good resistance against hot burning sodium. Present study is directed towards testing of limestone concrete blocks (made out of 13% ordinary portland cement, 8% water, 48% coarse limestone and 31 % fine limestone aggregates)

  2. Sodium sieving in children.

    NARCIS (Netherlands)

    Rusthoven, E.; Krediet, R.T.; Willems, J.L.; Monnens, L.A.H.; Schroder, C.H.

    2005-01-01

    Sodium sieving is a consequence of dissociation between the amount of water and sodium transported over the peritoneal membrane. This dissociation occurs in the presence of aquaporin-mediated water transport. Sieving of sodium can be used as a rough measure for aquaporin-mediated water transport.

  3. Sodium sieving in children

    NARCIS (Netherlands)

    Rusthoven, Esther; Krediet, Raymond T.; Willems, Hans L.; Monnens, Leo A.; Schröder, Cornelis H.

    2005-01-01

    Sodium sieving is a consequence of dissociation between the amount of water and sodium transported over the peritoneal membrane. This dissociation occurs in the presence of aquaporin-mediated water transport. Sieving of sodium can be used as a rough measure for aquaporin-mediated water transport.

  4. Acoustic signal processing for the detection of sodium boiling or sodium-water reaction in LMFRs. Final report of a co-ordinated research programme 1990-1995

    International Nuclear Information System (INIS)

    1997-05-01

    This report is a summary of the work performed under a co-ordinated research programme entitled Acoustic Signal Processing for the Detection of Sodium Boiling or Sodium-Water Reaction in Liquid Metal Cooled Fast Reactors. The programme was organized by the IAEA and carried out from 1990 to 1995. It was the continuation of an earlier research co-ordination programme entitled Signal Processing Techniques for Sodium Boiling Noise Detection, which was carried out from 1984 to 1989. Refs, figs, tabs

  5. Safety instrumentation for the sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Hall, R.S.

    1975-01-01

    The particular safety problems of the fast reactor and the role of instrumented protection in relation to the overall safety design of the reactor are discussed. The importance of the accident sequence arising from a fault within one subassembly is indicated, and the physical phenomena involved are discussed with regard to the generation of detectable signals. Several possible techniques for detecting subassembly accidents are described, including those with detectors situated at the outlet of each subassembly and also those involving whole-core parameters. Reference is made to the way in which types of instruments would have to be combined to give a high degree of protection to the system, the actual protection required being dependent on the overall safety intentions. Attention is drawn to the problems of minimizing the spurious trip rate for a well-instrumented reactor, which lead to stringent requirements on instrument reliability and/or replaceability. The possible role of the computer in handling the multiplicity of complex signals is mentioned, together with the problems that have to be solved before this can be done. It is concluded that satisfactory instrument protection is available for whole-core faults, but with regard to subassembly fault detection the situation is less clear. Although some information is available for guidance on the instruments and their specifications, the justification and achievability of the latter are dependent on development work that is still proceeding. It may well be that uncertainties concerning the effects of the reactor environment will require that some of this work take the form of in-reactor experiments. (auth)

  6. Sodium safety manual

    International Nuclear Information System (INIS)

    Hayes, D.J.; Gardiner, R.L.

    1980-09-01

    The sodium safety manual is based upon more than a decade of experience with liquid sodium at Berkeley Nuclear Laboratories (BNL). It draws particularly from the expertise and experience developed in the course of research work into sodium fires and sodium water reactions. It draws also on information obtained from the UKAEA and other sodium users. Many of the broad principles will apply to other Establishments but much of the detail is specific to BNL and as a consequence its application at other sites may well be limited. Accidents with sodium are at best unpleasant and at worst lethal in an extremely painful way. The object of this manual is to help prevent sodium accidents. It is not intended to give detailed advice on specific precautions for particular situations, but rather to set out the overall strategy which will ensure that sodium activities will be pursued safely. More detail is generally conveyed to staff by the use of local instructions known as Sodium Working Procedures (SWP's) which are not reproduced in this manual although a list of current SWP's is included. Much attention is properly given to the safe design and operation of larger facilities; nevertheless evidence suggests that sodium accidents most frequently occur in small-scale work particularly in operations associated with sodium cleaning and special care is needed in all such cases. (U.K.)

  7. Report of sodium cavitation

    International Nuclear Information System (INIS)

    Murai, Hitoshi; Shima, Akira; Oba, Toshisaburo; Kobayashi, Ryoji; Hashimoto, Hiroyuki

    1975-01-01

    The damage of components for LMFBRs due to sodium cavitation is serious problem. This report summarizes the following items, (1) mechanism of the incipience of sodium cavitation, (2) damage due to sodium cavitation, (3) detection method for sodium cavitation, and (4) estimation method for sodium cavitation by the comparison with water cavitation. Materials were collected from the reports on liquid metal cavitation, sodium cavitation and water cavitation published from 1965 to now. The mechanism of the incipience of sodium cavitation cavitation parameters (mean location, distributed amount or occurrence aspect and stability), experiment of causing cavitation with Venturi tube, and growth of bubbles within superheated sodium. The sodium cavitation damage was caused by magnetostriction vibration method and with Venturi tube. The state of damage was investigated with the cavitation performance of a sodium pump, and the damage was examined in view of the safety of LMFBR plants. Sodium cavitation was detected with acoustic method, radiation method, and electric method. The effect of physical property of liquid on incipient cavitation was studied. These are thermodynamic effect based on quasistatic thermal equilibrium condition and the effect of the physical property of liquid based on bubble dynamics. (Iwase, T.)

  8. Confirmation of shutdown cooling effects

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Kotaro, E-mail: ksato@nelted.co.jp; Tabuchi, Masato; Sugimura, Naoki; Tatsumi, Masahiro [Nuclear Engineering, Limited, 1-3-7 Tosabori Nishi-ku, Osaka-shi, Osaka 550-0001 (Japan)

    2015-12-31

    After the Fukushima accidents, all nuclear power plants in Japan have gradually stopped their operations and have long periods of shutdown. During those periods, reactivity of fuels continues to change significantly especially for high-burnup UO{sub 2} fuels and MOX fuels due to radioactive decays. It is necessary to consider these isotopic changes precisely, to predict neutronics characteristics accurately. In this paper, shutdown cooling (SDC) effects of UO{sub 2} and MOX fuels that have unusual operation histories are confirmed by the advanced lattice code, AEGIS. The calculation results show that the effects need to be considered even after nuclear power plants come back to normal operation.

  9. Confirmation of shutdown cooling effects

    Science.gov (United States)

    Sato, Kotaro; Tabuchi, Masato; Sugimura, Naoki; Tatsumi, Masahiro

    2015-12-01

    After the Fukushima accidents, all nuclear power plants in Japan have gradually stopped their operations and have long periods of shutdown. During those periods, reactivity of fuels continues to change significantly especially for high-burnup UO2 fuels and MOX fuels due to radioactive decays. It is necessary to consider these isotopic changes precisely, to predict neutronics characteristics accurately. In this paper, shutdown cooling (SDC) effects of UO2 and MOX fuels that have unusual operation histories are confirmed by the advanced lattice code, AEGIS. The calculation results show that the effects need to be considered even after nuclear power plants come back to normal operation.

  10. Corrosion inhibitors for solar-heating and cooling

    Science.gov (United States)

    Humphries, T. S.

    1979-01-01

    Report describes results of tests conducted to evaluate abilities of 12 candidate corrosion inhibitors to protect aluminum, steel, copper, or stainless steel at typical conditions encountered in solar heating and cooling systems. Inhibitors are based on sodium salts including nitrates, borates, silicates, and phosphates.

  11. Development of Korea advanced liquid metal reactor

    International Nuclear Information System (INIS)

    Park, C.K.

    1998-01-01

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

  12. Evaporative cooling of the dipolar hydroxyl radical.

    Science.gov (United States)

    Stuhl, Benjamin K; Hummon, Matthew T; Yeo, Mark; Quéméner, Goulven; Bohn, John L; Ye, Jun

    2012-12-20

    Atomic physics was revolutionized by the development of forced evaporative cooling, which led directly to the observation of Bose-Einstein condensation, quantum-degenerate Fermi gases and ultracold optical lattice simulations of condensed-matter phenomena. More recently, substantial progress has been made in the production of cold molecular gases. Their permanent electric dipole moment is expected to generate systems with varied and controllable phases, dynamics and chemistry. However, although advances have been made in both direct cooling and cold-association techniques, evaporative cooling has not been achieved so far. This is due to unfavourable ratios of elastic to inelastic scattering and impractically slow thermalization rates in the available trapped species. Here we report the observation of microwave-forced evaporative cooling of neutral hydroxyl (OH(•)) molecules loaded from a Stark-decelerated beam into an extremely high-gradient magnetic quadrupole trap. We demonstrate cooling by at least one order of magnitude in temperature, and a corresponding increase in phase-space density by three orders of magnitude, limited only by the low-temperature sensitivity of our spectroscopic thermometry technique. With evaporative cooling and a sufficiently large initial population, much colder temperatures are possible; even a quantum-degenerate gas of this dipolar radical (or anything else it can sympathetically cool) may be within reach.

  13. Phase Transformations During Cooling of Automotive Steels

    Science.gov (United States)

    Padgett, Matthew C.

    This thesis explores the effect of cooling rate on the microstructure and phases in advanced high strength steels (AHSS). In the manufacturing of automobiles, the primary joining mechanism for steel is resistance spot welding (RSW), a process that produces a high heat input and rapid cooling in the welded metal. The effect of RSW on the microstructure of these material systems is critical to understanding their mechanical properties. A dual phase steel, DP-600, and a transformation induced plasticity bainitic-ferritic steel, TBF-1180, were studied to assess the changes to their microstructure that take place in controlled cooling environments and in uncontrolled cooling environments, i.e. resistance spot welding. Continuous cooling transformation (CCT) diagrams were developed using strip specimens of DP-600 and TBF-1180 to determine the phase transformations that occur as a function of cooling rate. The resulting phases were determined using a thermal-mechanical simulator and dilatometry, combined with light optical microscopy and hardness measurements. The resulting phases were compared with RSW specimens where cooling rate was controlled by varying the welding time for two-plate welds. Comparisons were drawn between experimental welds of DP-600 and simulations performed using a commercial welding software. The type and quantity of phases present after RSW were examined using a variety of techniques, including light optical microscopy using several etchants, hardness measurements, and x-ray diffraction (XRD).

  14. The cooling of particle beams

    International Nuclear Information System (INIS)

    Sessler, A.M.

    1994-10-01

    A review is given of the various methods which can be employed for cooling particle beams. These methods include radiation damping, stimulated radiation damping, ionization cooling, stochastic cooling, electron cooling, laser cooling, and laser cooling with beam coupling. Laser Cooling has provided beams of the lowest temperatures, namely 1 mK, but only for ions and only for the longitudinal temperature. Recent theoretical work has suggested how laser cooling, with the coupling of beam motion, can be used to reduce the ion beam temperature in all three directions. The majority of this paper is devoted to describing laser cooling and laser cooling with beam coupling

  15. Implications of recent research on microstructure modifications, through heat-related processing and trait alteration to bio-functions, molecular thermal stability and mobility, metabolic characteristics and nutrition in cool-climate cereal grains and other types of seeds with advanced molecular techniques.

    Science.gov (United States)

    Ying, Yuguang; Zhang, Huihua; Yu, Peiqiang

    2018-02-16

    The cutting-edge synchrotron radiation based and globar-sourced vibrational infrared microspectroscopy have recently been developed. These novel techniques are able to reveal structure features at cellular and molecular levels with the tested tissues being intact. However, to date, the advanced techniques are unfamiliar or unknown to food and feed scientists and have not been used to study the molecular structure changes in cool-climate cereal grain seeds and other types of bio-oil and bioenergy seeds. This article aims to provide some recent research in cool-climate cereal grains and other types of seeds on molecular structures and metabolic characteristics of carbohydrate and protein, and implication of microstructure modification through heat-related processing and trait alteration to bio-functions, molecular thermal stability and mobility, and nutrition with advanced molecular techniques- synchrotron radiation based and globar-sourced vibrational infrared microspectroscopy in the areas of (1) Inherent microstructure of cereal grain seeds; (2) The nutritional values of cereal grains; (3) Impact and modification of heat-related processing to cereal grain; (4) Conventional nutrition evaluation methodology; (5) Synchrotron radiation-based and globar-sourced vibrational (micro)-spectroscopy for molecular structure study and molecular thermal stability and mobility, and (6) Recent molecular spectroscopic technique applications in research on raw, traits altered and processed cool-climate cereal grains and other types of seeds. The information described in this article gives better insights of research progress and update in cool-climate cereal grains and other seeds with advanced molecular techniques.

  16. Present status of the RCNP circulation ring comments on magnetized electron cooling

    International Nuclear Information System (INIS)

    Ando, A.

    1989-01-01

    After the brief introduction of the RCNP cooling synchrotron (MSR - Multipurpose Storage Ring), the characteristics of the circulation ring (CR), which is now under construction, are summarized. Some efforts toward the MSR, the next stage of the CR, are also described. In the MSR the completely new scheme will be used for beam cooling. The possibility of stimmulated magnetized electron cooling (say, advanced electron cooling), which lies between the ordinary electron cooling and the new scheme, is discussed. (author)

  17. Initial Cooling Experiment (ICE)

    CERN Multimedia

    Photographic Service; CERN PhotoLab

    1978-01-01

    In 1977, in a record-time of 9 months, the magnets of the g-2 experiment were modified and used to build a proton/antiproton storage ring: the "Initial Cooling Experiment" (ICE). It served for the verification of the cooling methods to be used for the "Antiproton Project". Stochastic cooling was proven the same year, electron cooling followed later. Also, with ICE the experimental lower limit for the antiproton lifetime was raised by 9 orders of magnitude: from 2 microseconds to 32 hours. For its previous life as g-2 storage ring, see 7405430. More on ICE: 7711282, 7809081, 7908242.

  18. Turbine airfoil cooling system with cooling systems using high and low pressure cooling fluids

    Energy Technology Data Exchange (ETDEWEB)

    Marsh, Jan H.; Messmann, Stephen John; Scribner, Carmen Andrew

    2017-10-25

    A turbine airfoil cooling system including a low pressure cooling system and a high pressure cooling system for a turbine airfoil of a gas turbine engine is disclosed. In at least one embodiment, the low pressure cooling system may be an ambient air cooling system, and the high pressure cooling system may be a compressor bleed air cooling system. In at least one embodiment, the compressor bleed air cooling system in communication with a high pressure subsystem that may be a snubber cooling system positioned within a snubber. A delivery system including a movable air supply tube may be used to separate the low and high pressure cooling subsystems. The delivery system may enable high pressure cooling air to be passed to the snubber cooling system separate from low pressure cooling fluid supplied by the low pressure cooling system to other portions of the turbine airfoil cooling system.

  19. Red blood cell sodium transport in patients with cirrhosis

    DEFF Research Database (Denmark)

    Henriksen, Ulrik Lütken; Kiszka-Kanowitz, Marianne; Bendtsen, Flemming

    2016-01-01

    Patients with advanced cirrhosis have abnormal sodium homoeostasis. The study was undertaken to quantify the sodium transport across the plasma membrane of red blood cells (RBC) in patients with cirrhosis. RBC efflux and influx of sodium were studied in vitro with tracer (22) Na(+) according...... to linear kinetics in 24 patients with cirrhosis and 14 healthy controls. The sodium efflux was modified by ouabain (O), furosemide (F) and a combination of O and F (O + F). RBC sodium was significantly decreased (4·6 versus control 6·3 mmol l(-1) , Psodium (r = 0·57, P......sodium efflux was higher in patients with cirrhosis (+46%, Psodium buffers showed that the F-insensitive sodium efflux was twice as high in cirrhosis as in controls (P = 0...

  20. Clinch River breeder reactor sodium fire protection system design and development

    International Nuclear Information System (INIS)

    Foster, K.W.; Boasso, C.J.; Kaushal, N.N.

    1984-01-01

    To assure the protection of the public and plant equipment, improbable accidents were hypothesized to form the basis for the design of safety systems. One such accident is the postulated failure of the Intermediate Heat Transfer System (IHTS) piping within the Steam Generator Building (SGB), resulting in a large-scale sodium fire. This paper discusses the design and development of plant features to reduce the consequences of the accident to acceptable levels. Additional design solutions were made to mitigate the sodium spray contribution to the accident scenario. Sodium spill tests demonstrated that large sodium leaks can be safely controlled in a sodium-cooled nuclear power plant

  1. Computer analysis of sodium cold trap design and performance

    International Nuclear Information System (INIS)

    McPheeters, C.C.; Raue, D.J.

    1983-11-01

    Normal steam-side corrosion of steam-generator tubes in Liquid Metal Fast Breeder Reactors (LMFBRs) results in liberation of hydrogen, and most of this hydrogen diffuses through the tubes into the heat-transfer sodium and must be removed by the purification system. Cold traps are normally used to purify sodium, and they operate by cooling the sodium to temperatures near the melting point, where soluble impurities including hydrogen and oxygen precipitate as NaH and Na 2 O, respectively. A computer model was developed to simulate the processes that occur in sodium cold traps. The Model for Analyzing Sodium Cold Traps (MASCOT) simulates any desired configuration of mesh arrangements and dimensions and calculates pressure drops and flow distributions, temperature profiles, impurity concentration profiles, and impurity mass distributions

  2. Theoretical Study of Sodium-Water Surface Reaction Mechanism

    Science.gov (United States)

    Kikuchi, Shin; Kurihara, Akikazu; Ohshima, Hiroyuki; Hashimoto, Kenro

    Computational study of the sodium-water reaction at the gas (water) - liquid (sodium) interface has been carried out using the ab initio (first-principle) method. A possible reaction channel has been identified for the stepwise OH bond dissociations of a single water molecule. The energetics including the binding energy of a water molecule on the sodium surface, the activation energies of the bond cleavages, and the reaction energies, have been evaluated, and the rate constants of the first and second OH bond-breakings have been compared. It was found that the estimated rate constant of the former was much larger than the latter. The results are the basis for constructing the chemical reaction model used in a multi-dimensional sodium-water reaction code, SERAPHIM, being developed by Japan Atomic Energy Agency (JAEA) toward the safety assessment of the steam generator (SG) in a sodium-cooled fast reactor (SFR).

  3. Theoretical study of sodium-water surface reaction mechanism

    International Nuclear Information System (INIS)

    Kikuchi, Shin; Kurihara, Akikazu; Ohshima, Hiroyuki; Hashimoto, Kenro

    2012-01-01

    Computational study of the sodium-water reaction at the gas (water) - liquid (sodium) interface has been carried out using the ab initio (first-principle) method. A possible reaction channel has been identified for the stepwise OH bond dissociations of a single water molecule. The energetics including the binding energy of a water molecule on the sodium surface, the activation energies of the bond cleavages, and the reaction energies, have been evaluated, and the rate constants of the first and second OH bond-breakings have been compared. It was found that the estimated rate constant of the former was much larger than the latter. The results are the basis for constructing the chemical reaction model used in a multi-dimensional sodium-water reaction code, SERAPHIM, being developed by Japan Atomic Energy Agency (JAEA) toward the safety assessment of the steam generator (SG) in a sodium-cooled fast reactor (SFR). (author)

  4. Dialysate sodium and sodium gradient in maintenance hemodialysis: a neglected sodium restriction approach?

    OpenAIRE

    Munoz Mendoza, Jair; Sun, Sumi; Chertow, Glenn M.; Moran, John; Doss, Sheila; Schiller, Brigitte

    2011-01-01

    Background. A higher sodium gradient (dialysate sodium minus pre-dialysis plasma sodium) during hemodialysis (HD) has been associated with sodium loading; however, its role is not well studied. We hypothesized that a sodium dialysate prescription resulting in a higher sodium gradient is associated with increases in interdialytic weight gain (IDWG), blood pressure (BP) and thirst.

  5. Polymers as advanced materials for desiccant applications, 1988

    Energy Technology Data Exchange (ETDEWEB)

    Czanderna, A.W.; Neidlinger, H.H.

    1990-09-01

    This report documents work to identify a next-generation, low-cost material with which solar energy or heat from another low-cost energy source can be used for regenerating the water vapor sorption activity of the desiccant. The objective of the work is to determine how the desired sorption performance of advanced desiccant materials can be predicted by understanding the role of the material modifications and material surfaces. The work concentrates on solid materials to be used for desiccant cooling systems and which process water vapor in an atmosphere to produce cooling. The work involved preparing modifications of polystyrene sulfonic acid sodium salt, synthesizing a hydrogel, and evaluating the sorption performances of these and similar commercially available polymeric materials; all materials were studied for their potential application in solid commercial desiccant cooling systems. Background information is also provided on desiccant cooling systems and the role of a desiccant material within such a system, and it includes the use of polymers as desiccant materials. 31 refs., 16 figs., 5 tabs.

  6. Decommissioning of experimental breeder reactor - II. Complex, post sodium draining

    International Nuclear Information System (INIS)

    Michelbacher, J.A.; Henslee, S. P.; Knight, C.J.; Sherman, S.R.

    2005-01-01

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

  7. The final cool down

    CERN Multimedia

    Thursday 29th May, the cool-down of the final sector (sector 4-5) of LHC has begun, one week after the start of the cool-down of sector 1-2. It will take five weeks for the sectors to be cooled from room temperature to 5 K and a further two weeks to complete the cool down to 1.9 K and the commissioning of cryogenic instrumentation, as well as to fine tune the cryogenic plants and the cooling loops of cryostats.Nearly a year and half has passed since sector 7-8 was cooled for the first time in January 2007. For Laurent Tavian, AT/CRG Group Leader, reaching the final phase of the cool down is an important milestone, confirming the basic design of the cryogenic system and the ability to operate complete sectors. “All the sectors have to operate at the same time otherwise we cannot inject the beam into the machine. The stability and reliability of the cryogenic system and its utilities are now very important. That will be the new challenge for the coming months,” he explains. The status of the cool down of ...

  8. Solar absorption cooling

    NARCIS (Netherlands)

    Kim, D.S.

    2007-01-01

    As the world concerns more and more on global climate changes and depleting energy resources, solar cooling technology receives increasing interests from the public as an environment-friendly and sustainable alternative. However, making a competitive solar cooling machine for the market still

  9. Cooling of electronic equipment

    DEFF Research Database (Denmark)

    A. Kristensen, Anders Schmidt

    2003-01-01

    Cooling of electronic equipment is studied. The design size of electronic equipment decrease causing the thermal density to increase. This affect the cooling which can cause for example failures of critical components due to overheating or thermal induced stresses. Initially a pin fin heat sink...

  10. Coherent electron cooling

    Energy Technology Data Exchange (ETDEWEB)

    Litvinenko,V.

    2009-05-04

    Cooling intense high-energy hadron beams remains a major challenge in modern accelerator physics. Synchrotron radiation is still too feeble, while the efficiency of two other cooling methods, stochastic and electron, falls rapidly either at high bunch intensities (i.e. stochastic of protons) or at high energies (e-cooling). In this talk a specific scheme of a unique cooling technique, Coherent Electron Cooling, will be discussed. The idea of coherent electron cooling using electron beam instabilities was suggested by Derbenev in the early 1980s, but the scheme presented in this talk, with cooling times under an hour for 7 TeV protons in the LHC, would be possible only with present-day accelerator technology. This talk will discuss the principles and the main limitations of the Coherent Electron Cooling process. The talk will describe the main system components, based on a high-gain free electron laser driven by an energy recovery linac, and will present some numerical examples for ions and protons in RHIC and the LHC and for electron-hadron options for these colliders. BNL plans a demonstration of the idea in the near future.

  11. Measure Guideline: Ventilation Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Springer, D.; Dakin, B.; German, A.

    2012-04-01

    The purpose of this measure guideline on ventilation cooling is to provide information on a cost-effective solution for reducing cooling system energy and demand in homes located in hot-dry and cold-dry climates. This guideline provides a prescriptive approach that outlines qualification criteria, selection considerations, and design and installation procedures.

  12. A review of thermoelectric cooling: Materials, modeling and applications

    International Nuclear Information System (INIS)

    Zhao, Dongliang; Tan, Gang

    2014-01-01

    This study reviews the recent advances of thermoelectric materials, modeling approaches, and applications. Thermoelectric cooling systems have advantages over conventional cooling devices, including compact in size, light in weight, high reliability, no mechanical moving parts, no working fluid, being powered by direct current, and easily switching between cooling and heating modes. In this study, historical development of thermoelectric cooling has been briefly introduced first. Next, the development of thermoelectric materials has been given and the achievements in past decade have been summarized. To improve thermoelectric cooling system's performance, the modeling techniques have been described for both the thermoelement modeling and thermoelectric cooler (TEC) modeling including standard simplified energy equilibrium model, one-dimensional and three-dimensional models, and numerical compact model. Finally, the thermoelectric cooling applications have been reviewed in aspects of domestic refrigeration, electronic cooling, scientific application, and automobile air conditioning and seat temperature control, with summaries for the commercially available thermoelectric modules and thermoelectric refrigerators. It is expected that this study will be beneficial to thermoelectric cooling system design, simulation, and analysis. - Highlights: •Thermoelectric cooling has great prospects with thermoelectric material's advances. •Modeling techniques for both thermoelement and TEC have been reviewed. •Principle thermoelectric cooling applications have been reviewed and summarized

  13. Thermionic cooling in semiconductor multilayers

    International Nuclear Information System (INIS)

    Lee, S.; Lewis, R.A.; Lough, B.; Zhang, C.

    2000-01-01

    Full text: A solid-state refrigerator in which electrons transport heat has advantages over the conventional vapour-cycle, compressor-based domestic refrigerator since it has no moving parts, it is low-maintenance, silent, vibration-free and does not require the use of refrigerant gases. The usual approach to making an all-electrical refrigerator is by thermoelectric refrigeration. After a period of intense research in the 1950s and 60s it was realised that the efficiency of thermoelectric devices was less than, and unlikely to exceed, that of conventional compressor units. While thermoelectric cooling has found specialised applications in cases where reliability, compactness and weight are important considerations, it does not appear that thermo-electrics will ever successfully compete in the domestic market, in spite of recent advances in the design and fabrication of thermoelectric materials. A new approach to an all-electric refrigerator is to employ thermionic emission over potential barriers. A key difference between a thermoelectric device and a thermionic device is that in the former the electrons are scattered in their motion and in the latter they are not. Thus thermionic cooling, in principle, can be much more efficient than thermoelectric cooling. A radical new realisation of the thermionic refrigerator was suggested recently in which a multilayer semiconductor structure would be used. We discuss the optimisation of such a multilayer semiconductor cooling system by considering (1) electron-phonon interactions in the barriers and electrodes; (2) the detailed treatment of thermal conductivity; (3) an exact numerical solution of the heat and energy currents (in contrast to the previous approximate analytic solutions); (4) the effect of varying layer thickness across the device; and (5) the effect of varying current density across the device

  14. INITIAL COOLING EXPERIMENT (ICE)

    CERN Multimedia

    CERN PhotoLab

    1979-01-01

    ICE was built in 1977, using the modified bending magnets of the g-2 muon storage ring (see 7405430). Its purpose was to verify the validity of stochastic and electron cooling for the antiproton project. Stochastic cooling proved a resounding success early in 1978 and the antiproton project could go ahead, now entirely based on stochastic cooling. Electron cooling was experimented with in 1979. The 26 kV equipment is housed in the cage to the left of the picture, adjacent to the "e-cooler" located in a straight section of the ring. With some modifications, the cooler was later transplanted into LEAR (Low Energy Antiproton Ring) and then, with further modifications, into the AD (Antiproton Decelerator), where it cools antiprotons to this day (2006). See also: 7711282, 7802099, 7809081.

  15. Initial Cooling Experiment (ICE)

    CERN Multimedia

    CERN PhotoLab

    1978-01-01

    ICE was built in 1977, in a record time of 9 months, using the modified bending magnets of the g-2 muon storage ring. Its purpose was to verify the validity of stochastic and electron cooling for the antiproton project, to be launched in 1978. Already early in 1978, stochastic cooling proved a resounding success, such that the antiproton (p-pbar)project was entirely based on it. Tests of electron cooling followed later: protons of 46 MeV kinetic energy were cooled with an electron beam of 26 kV and 1.3 A. The cage seen prominently in the foreground houses the HV equipment, adjacent to the "cooler" installed in a straight section of the ring. With some modifications, the cooler was later transplanted into LEAR (Low Energy Antiproton Ring) and then, with further modifications, into the AD (Antiproton Decelerator), where it cools antiprotons to this day (2006). See also: 7711282, 7802099, 7908242.

  16. Oscillator strength of instantaneous diatomic sodium molecules

    Energy Technology Data Exchange (ETDEWEB)

    Perny, G.

    1991-10-01

    We introduce definitely the notion of instantaneous molecules, real or fictitious, in spite of its utilization in certain fields of the supraconductivity, genetics and advanced medical research. Calculation of the oscillator strength of instantaneous sodium diatomic molecules gives (f{sub mol})sup(Na{sub 2(i)})=6,86. This method is transposable at lithium and other Ia elements. (orig.).

  17. Oscillator strength of instantaneous diatomic sodium molecules

    Science.gov (United States)

    Perny, G.

    1991-12-01

    We introduce definitely the notion of instantaneous molecules, real or fictitious, in spite of its utilization in certain fields of the supraconductivity, genetics and advanced medical research. Calculation of the oscillator strength of instantaneous sodium diatomic molecules gives [ f mol]Na 2( i)=6,86. This method is transposable at lithium and other Ia elements.

  18. Sodium fire protection

    International Nuclear Information System (INIS)

    Raju, C.; Kale, R.D.

    1979-01-01

    Results of experiments carried out with sodium fires to develop extinguishment techniques are presented. Characteristics, ignition temperature, heat evolution and other aspects of sodium fires are described. Out of the powders tested for extinguishment of 10 Kg sodium fires, sodium bi-carbonate based dry chemical powder has been found to be the best extinguisher followed by large sized vermiculite and then calcium carbonate powders distributed by spray nozzles. Powders, however, do not extinguish large fires effectively due to sodium-concrete reaction. To control large scale fires in a LMFBR, collection trays with protective cover have been found to cause oxygen starvation better than flooding with inert gas. This system has an added advantage in that there is no damage to the sodium facilities as has been in the case of powders which often contain chlorine compounds and cause stress corrosion cracking. (M.G.B.)

  19. LOX/Methane Regeneratively-Cooled Rocket Engine Development

    Data.gov (United States)

    National Aeronautics and Space Administration — The purpose of this project is to advance the technologies required to build a subcritical regeneratively cooled liquid oxygen/methane rocket combustion chamber for...

  20. Construction, assembling and operation of an equipment for sodium purity

    International Nuclear Information System (INIS)

    Becquart, E.T.; Botbol, J.; Echenique, P.N.; Fruchtenicht, F.W.; Gil, D.A.; Perillo, P.; Vardich, R.N.; Vigo, D.E.

    1993-01-01

    The purpose of this work is the production of high purity metallic sodium for bench-scale, research studies. A stainless steel equipment was built and assembled, including high vacuum, heating and cooling systems. It was satisfactorily operated in two successive steps, filtration and vacuum distillation, with a good yield. (Author). 5 refs., 5 figs