Relevant bibliographies by topics / Sodium cooled fast reactors

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  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers
  6. Reports

Academic literature on the topic 'Sodium cooled fast reactors'

Author: Grafiati
Published: 4 June 2021
Last updated: 1 February 2022

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Journal articles on the topic "Sodium cooled fast reactors"

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Dawn, William C., and Scott Palmtag. "A MULTIPHYSICS SIMULATION SUITE FOR SODIUM COOLED FAST REACTORS." EPJ Web of Conferences 247 (2021): 06019. http://dx.doi.org/10.1051/epjconf/202124706019.

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A simulation suite has been developed to model reactor power distribution and multiphysics feedback effects in Sodium-cooled Fast Reactors (SFRs). This suite is based on the Finite Element Method (FEM) and employs a general, unstructured mesh to solve the Simplified P3 (SP3) neutron transport equations. In the FEM implementation, two-dimensional triangular elements and three-dimensional wedge elements are selected. Wedge elements are selected for their natural description of hexagonal geometry common to fast reactors. Thermal feedback effects within fast reactors are modeled within the simulation suite. A thermal hydraulic model is developed, modeling both axial heat convection and radial heat conduction within fuel assemblies. A thermal expansion model is included and is demonstrated to significantly affect reactivity. This simulation suite has been employed to model the Advanced Burner Reactor (ABR) benchmark, specifically the MET-1000. It has been demonstrated that these models sufficiently describe the multiphysics feedback phenomena and can be used to estimate multiphysics reactivity feedback coefficients.
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Mishra, Govind Kumar, M. Sakthivel, S. L. N. Swamy, and K. Madhusoodanan. "Instrumentation for Sodium-Cooled Fast Breeder Reactors." Nuclear Science and Engineering 174, no. 1 (May 2013): 96–102. http://dx.doi.org/10.13182/nse10-74.

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Konomura, Mamoru, and Masakazu Ichimiya. "Design challenges for sodium cooled fast reactors." Journal of Nuclear Materials 371, no. 1-3 (September 2007): 250–69. http://dx.doi.org/10.1016/j.jnucmat.2007.05.012.

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Bagdasarov, Yu E. "Unanticipated accidents in sodium-cooled fast reactors." Atomic Energy 108, no. 3 (June 25, 2010): 165–69. http://dx.doi.org/10.1007/s10512-010-9272-5.

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Horowitz, Emmanuel. "ICONE19-43335 FURTHER BASIC STUDIES NEEDED TO SPECIFY MATERIALS FOR SODIUM COOLED FAST REACTORS." Proceedings of the International Conference on Nuclear Engineering (ICONE) 2011.19 (2011): _ICONE1943. http://dx.doi.org/10.1299/jsmeicone.2011.19._icone1943_149.

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Zhang, Zhen Guo, and Jie Xie. "The Application of New Materials in New Nuclear Reactors." Advanced Materials Research 1030-1032 (September 2014): 197–200. http://dx.doi.org/10.4028/www.scientific.net/amr.1030-1032.197.

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This paper combines the technical features and operation condition of the new nuclear reactors, such as sodium cooled fast reactor (SFR), supercritical water cooled reactor (SCWR), very high temperature reactor (VHTR), molten salt reactor (MSR) and nuclear fusion reactor, and analyzes the performance requirements of materials, it introduces the current application and prospects of new materials in different new nuclear reactors.
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Tenchine, D., C. Fournier, and Y. Dolias. "Gas entrainment issues in sodium cooled fast reactors." Nuclear Engineering and Design 270 (April 2014): 302–11. http://dx.doi.org/10.1016/j.nucengdes.2014.02.002.

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Crawford, Douglas C., Douglas L. Porter, and Steven L. Hayes. "Fuels for sodium-cooled fast reactors: US perspective." Journal of Nuclear Materials 371, no. 1-3 (September 2007): 202–31. http://dx.doi.org/10.1016/j.jnucmat.2007.05.010.

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Le coz, Pierre, Jean-François Sauvage, and Jean-Pol Serpantie. "Sodium-Cooled Fast Reactors: the ASTRID Plant Project." Revue Générale Nucléaire, no. 5 (September 2011): 39–44. http://dx.doi.org/10.1051/rgn/20115039.

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Bagdasarov, Yu E., and A. A. Kamaev. "Successive development phases for sodium-cooled fast reactors." Atomic Energy 111, no. 6 (March 23, 2012): 389–97. http://dx.doi.org/10.1007/s10512-012-9509-6.

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Dissertations / Theses on the topic "Sodium cooled fast reactors"

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Wutzler, Whitney A. "Analysis of Accidents in Sodium-Cooled Fast Reactors." The Ohio State University, 2011. http://rave.ohiolink.edu/etdc/view?acc_num=osu1306791253.

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Gottfridsson, Filip. "Simulation of Reactor Transient and Design Criteria of Sodium-cooled Fast Reactors." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2010. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-148572.

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The need for energy is growing in the world and the market of nuclear power is now once more expanding. Some issues of the current light-water reactors can be solved by the next generation of nuclear power, Generation IV, where sodium-cooled reactors are one of the candidates. Phénix was a French prototype sodium-cooled reactor, which is seen as a success. Although it did encounter an earlier unexperienced phenomenon, A.U.R.N., in which a negative reactivity transient followed by an oscillating behavior forced an automatic emergency shutdown of the reactor. This phenomenon lead to a lot of downtime of the reactor and is still unsolved. However, the most probable cause of the transients is radial movements of the core, referred to as core-flowering. This study has investigated the available documentation of the A.U.R.N. events. A simplified model of core-flowering was also created in order to simulate how radial expansion affects the reactivity of a sodium-cooled core. Serpent, which is a Monte-Carlo based simulation code, was chosen as calculation tool. Furthermore, a model of the Phénix core was successfully created and partly validated. The model of the core has a k_eff = 1.00298 and a neutron flux of (8.43+-0.02)!10^15 neutrons/cm^2 at normal state. The result obtained from the simulations shows that an expansion of the core radius decreases the reactivity. A linear approximation of the result gave the relation: change in k_eff/core extension = - 60 pcm/mm. This value corresponds remarkably well to the around - 60 pcm/mm that was obtained from the dedicated core-flowering experiments in Phénix made by the CEA. Core-flowering can recreate similar signals to those registered during the A.U.R.N. events, though the absence of trace of core movements in Phénix speaks against this. However, if core-flowering is the sought answer, it can be avoided by design. The equipment that registered the A.U.R.N. events have proved to be insensitive to noise. Though, the high amplitude of the transients and their rapidness have made some researcher believe that the events are a combination of interference in the equipment of Phénix and a mechanical phenomenon. Regardless, the origin of A.U.R.N. seems to be bound to some specific parameter of Phénix due to the fact that the transients only have occurred in this reactor. A safety analysis made by an expert committee, appointed by CEA, showed that the A.U.R.N. events are not a threat to the safety of Phénix. However, the origin of these negative transients has to be found before any construction of a commercial size sodium-cooled fast reactor can begin. Thus, further research is needed.
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Umbel, Marissa. "Containment Source Terms for Sodium-Cooled Fast Reactor Accidents." The Ohio State University, 2011. http://rave.ohiolink.edu/etdc/view?acc_num=osu1306506275.

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Sumner, Tyler. "Effects of fuel type on the safety characteristics of a sodium cooled fast reactor." Diss., Georgia Institute of Technology, 2010. http://hdl.handle.net/1853/37217.

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A series of accident simulations were performed using INL's thermal hydraulics code RELAP5-3D to analyze steady-state and transient behavior of a sodium cooled fast reactor. The reactor chosen for this study was General Electric's S-PRISM, which is a 1,000 MWt pool-type sodium-cooled fast reactor, designed for either an Oxide or Metal fueled core. Once key core characteristics including power profiles, reactivity feedback coefficients and delayed neutron parameters were calculated, S-PRISM was redesigned for a Nitride fueled core to take advantage of the Nitride fuel's high thermal conductivity and melting temperature. Loss of flow, loss of heat sink, loss of power and inadvertent control rod withdrawal accidents were simulated for each core at beginning, middle and end of cycle to determine if one fuel type provides significant safety advantages over the others.
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Johnson, Brian C. (Brian Carl). "Application of the technology neutral framework to sodium cooled fast reactors." Thesis, Massachusetts Institute of Technology, 2010. http://hdl.handle.net/1721.1/76499.

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Thesis (Ph. D.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, 2010.
Cataloged from PDF version of thesis.
Includes bibliographical references (p. 115-116).
Sodium cooled fast reactors (SFRs) are considered as a novel example to exercise the Technology Neutral Framework (TNF) proposed in NUREG- 1860. One reason for considering SFRs is that they have historically had a licensing problem due to postulated core disruptive accidents. Two SFR designs are considered, and both meet the goals of the TNF that LWRs typically would not. Considering these goals have been met, a method for improving economics is proposed where systems of low risk-importance are identified as candidates for removal, simplification, or removal from safety grade. Seismic risk dominates these designs and is found to be a limiting factor when applying the TNF. The contributions of this thesis are the following: 1. Functional event trees are developed as a tool to allow different designs to be compared on an equal basis. Functional event trees are useful within the TNF as a method for the selection of Licensing Basis Events (LBEs) which take the place of traditional Design Basis Accidents. 2. A new importance measure, Limit Exceedance Factor (LEF), is introduced that measures the margin in system failure probability. It can be used directly with the TNF where standard importance measures cannot. It also reveals that some systems that appear to be of high risk-importance with standard importance measures may have significant margin. 3. The seismic risk dominates these designs. It is shown that even under optimistic assumptions, the goals of the TNF cannot be met by a typical reactor. The effect of seismic isolation to reduce the frequency of seismically initiated large releases is also analyzed and found to be insufficient to reach the goals of the TNF. A limit on initiating event frequency that is consistent with current practices is proposed.
by Brian C. Johnson.
Ph.D.
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Verma, Vasudha. "Development of a Neutron Flux Monitoring System for Sodium-cooled Fast Reactors." Doctoral thesis, Uppsala universitet, Tillämpad kärnfysik, 2017. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-319945.

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Safety and reliability are one of the key objectives for future Generation IV nuclear energy systems. The neutron flux monitoring system forms an integral part of the safety design of a nuclear reactor and must be able to detect any irregularities during all states of reactor operation. The work in this thesis mainly concerns the detection of in-core perturbations arising from unwanted movements of control rods with in-vessel neutron detectors in a sodium-cooled fast reactor. Feasibility study of self-powered neutron detectors (SPNDs) with platinum emitters as in-core power profile monitors for SFRs at full power is performed. The study shows that an SPND with a platinum emitter generates a prompt current signal induced by neutrons and gammas of the order of 600 nA/m, which is large enough to be measurable. Therefore, it is possible for the SPND to follow local power fluctuations at full power operation. Ex-core and in-core detector locations are investigated with two types of detectors, fission chambers and self-powered neutron detectors (SPNDs) respectively, to study the possibility of detection of the spatial changes in the power profile during two different transient conditions, i.e. inadvertent withdrawal of control rods (IRW) and one stuck rod during reactor shutdown (OSR). It is shown that it is possible to detect the two simulated transients with this set of ex-core and in-core detectors before any melting of the fuel takes place. The detector signal can tolerate a noise level up to 5% during an IRW and up to 1% during an OSR.
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Zhang, Youpeng. "Transmutation of Am in sodium fast reactors and accelerator driven systems." Doctoral thesis, KTH, Reaktorfysik, 2012. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-72427.

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In this thesis, the feasibility to use sodium cooled fast reactors loaded with MOX, metallic and nitride fuels for efficient transmutation of americium is investigated by performing transient analysis for cases with different americium contents in fuels, using safety parameters obtained with the SERPENT Monte Carlo code. It was then demonstrated that there is no solid limit for the Am introduction into oxide, metallic and nitride fuels that were loaded into sodium fast reactors. Instead, higher Am contents could be permitted if specific levels of power penalty were accepted. Transient analysis of a new Accelerator Driven System design with higher neutron source efficiency than the reference EFIT-400 design, was also performed. Based on simulation results, the suggested ADS design was proved to survive the full set of transients, preserving 130 K margin to cladding rupture during the most limiting transient. After comparing Am transmutation performances in SFRs and the suggested ADS, it can be concluded that: 1. Nitride fuel could provide the highest Am transmutation efficiency, when loaded into SFRs; 2. One SFR loaded with nitride fuel is sufficient to transmute Am inventory produced by more than 15 commercial LWRs within the same time period, which is three times higher than the supporting ratio reported for the suggested ADS; 3. The total fraction of ADS power in the power park is half of cases for critical reactors.
QC 20120201
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Nitta, Christopher C. "Applying risk informed methodologies to improve the economics of sodium-cooled fast reactors." Thesis, Massachusetts Institute of Technology, 2010. http://hdl.handle.net/1721.1/58086.

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Thesis (S.M.)--Massachusetts Institute of Technology, Dept. of Nuclear Science and Engineering, February 2010.
Cataloged from PDF version of thesis.
Includes bibliographical references (p. 94-97).
In order to support the increasing demand for clean sustainable electricity production and for nuclear waste management, the Sodium-Cooled Fast Reactor (SFR) is being developed. The main drawback has been its high capital and operating costs in comparison with traditional light water reactors. In order to compete, the SFR must be shown to be economically competitive. This study makes use of the proposed Technology Neutral Framework (TNF) being developed by the U.S. NRC. By applying this risk-based approach to safety, rather than the traditional approach of applying deterministic requirements, it will be shown that significant savings can be realized without compromising fundamental safety. A methodology was developed using the Technology Neutral Framework to judge design alternatives based on risk significance that provide acceptable safety within the framework at less cost. The key probabilistic metrics of Risk Achievement Worth and Limit Exceedence Factor will be used to assess whether a system or component plays an important safety function. If not the system, structure or component either can be eliminated, modified or its safety grade can be reduced resulting in cost savings. In addition, assessments were made to determine how to improve thermal efficiency by raising reactor exit temperature and by applying other design alternatives to reduce costs as evaluated on a safety, reliability and economic basis.
(cont.) This methodology was applied in a series of case studies demonstrating the value of the approach in design. The probabilistic risk assessment, the reference economic model and the Technology Neutral Framework tools required for this methodology are described. A reference economic model for a pool-type SFR was developed using the G4-ECONS model since it is an acceptable standard model for economic analysis. Since cost predictions for sodium cooled fast reactors are highly uncertain, the results of the economic analysis are used to estimate the relative improvement in cost as a function of the design alternatives proposed by the TNF methodology approach. This study used generic and comparative numbers for the ALMR and SPRISM reactors for cost of components of the SFR, to identify capital cost drivers for further study and cost reduction. For comparative purposes, the light water reactor (LWR) economic model in the G4-ECONS model was used and benchmarked to current LWR data. As a result of the case studies in which the methodology was applied, it was shown that the capital cost of the SFR could be reduced by almost 18% ($336 million) over the reference design and the levelized generating costs could be reduced by over 10% (almost 1 cent/kw-hr). These savings come largely from improvements in thermal efficiency, elimination of the energetic core disruptive accident as a design basis event and simplification of the reactor shutdown system based on risk analysis and safety significance. Should this methodology be applied to the entire plant design, it is expected that significant additional savings could be identified.
by Christopher C. Nitta.
S.M.
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Brunett, Acacia Joann. "A Methodology for Analyzing the Consequences of Accidents in Sodium-Cooled Fast Reactors." The Ohio State University, 2010. http://rave.ohiolink.edu/etdc/view?acc_num=osu1280956877.

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Svanström, Sebastian. "Load following with a passive reactor core using the SPARC design." Thesis, Uppsala universitet, Tillämpad kärnfysik, 2016. http://urn.kb.se/resolve?urn=urn:nbn:se:uu:diva-296803.

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This thesis is a follow up on "SPARC fast reactor design: Design of two passively metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control" by Tobias Lindström (2015). In this thesis the two reactors designed by Lindström in said thesis were evaluated. The goal was to determine the reactors ability to load follow as well as the burnup of the neutron absorber used in the passive control system. To be able to determine the dynamic behaviour of the reactors the reactivity feedbacks of the cores were modelled using Serpent, a Monte Carlo simulation software for 3D neutron transport calculations. These feedbacks were then implemented into a dynamic simulation of the core, primary and secondary circulation and steam generator. The secondary circulation and feedwater flow were used to regulate steam temperature and turbine power. The core was left at constant coolant flow and no control rods were used. The simulations showed that the reactor was able to load follow between 100 % and 40 % of rated power at a speed of 6 % per minute. It was also shown that the reactor could safely adjust its power between 100 % and 10 % of rated power suggesting that load following is possible below 40 % of rated power but at a lower speed. Finally the reactors were allowed compensate for the variations in a week of the Latvian wind power production in order to show one possible application of the reactor.
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Books on the topic "Sodium cooled fast reactors"

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Cheng, Songbai, and Ruicong Xu. Safety of Sodium-Cooled Fast Reactors. Singapore: Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-6116-7.

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V. M. (Vladimir Mikhaĭlovich) Poplavskiĭ. Bezopasnostʹ parogeneratorov natriĭ-voda. Moskva: Ėnergoatomizdat, 1990.

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Kikō, Genshiryoku Anzen Kiban. Natoriumu reikyakugata kōsokuro no nenryō anzen sekkei oyobi hyōka ni kansuru kihonteki kangaekata: Basic concept of fuel safety design and assessment for sodium-cooled fast reactor. Tōkyō-to Minato-ku: Genshiryoku Anzen Kiban Kikō, 2013.

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Arkhipov, V. M. Tekhnika raboty s natriem na AĖS. Moskva: Ėnergoatomizdat, 1986.

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Mi, Xu, ed. Na gong yi ji chu. Beijing Shi: Zhongguo yuan zi neng chu ban chuan mei you xian gong si, 2011.

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Oka, Yoshiaki. Super light water reactors and super fast reactors: Supercritical-pressure light water cooled reactor. New York: Springer, 2010.

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Glushkov, Evgeniĭ Serafimovich. Razrabotka bystrykh gazookhlazhdaemykh reaktorov v Rossii. Moskva: IzdAT, 2008.

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Glushkov, Evgeniĭ Serafimovich. Razrabotka bystrykh gazookhlazhdaemykh reaktorov v Rossii. Moskva: IzdAT, 2008.

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Willem Frederik Geert van Rooijen. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor. Amsterdam: IOS Press, 2006.

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Avariĭnye i perekhodnye prot͡sessy v bystrykh reaktorakh. Moskva: Ėnergoatomizdat, 1987.

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Book chapters on the topic "Sodium cooled fast reactors"

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Cheng, Songbai, and Ruicong Xu. "Molten-Pool Sloshing Motion." In Safety of Sodium-Cooled Fast Reactors, 51–134. Singapore: Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-6116-7_3.

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Cheng, Songbai, and Ruicong Xu. "Conclusion and Future Prospect." In Safety of Sodium-Cooled Fast Reactors, 301–6. Singapore: Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-6116-7_6.

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Cheng, Songbai, and Ruicong Xu. "Debris Bed Formation Behavior." In Safety of Sodium-Cooled Fast Reactors, 135–212. Singapore: Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-6116-7_4.

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Cheng, Songbai, and Ruicong Xu. "Introduction." In Safety of Sodium-Cooled Fast Reactors, 1–17. Singapore: Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-6116-7_1.

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Cheng, Songbai, and Ruicong Xu. "Molten-Pool Mobility." In Safety of Sodium-Cooled Fast Reactors, 19–50. Singapore: Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-6116-7_2.

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Cheng, Songbai, and Ruicong Xu. "Debris Bed Self-Leveling Behavior." In Safety of Sodium-Cooled Fast Reactors, 213–99. Singapore: Springer Singapore, 2021. http://dx.doi.org/10.1007/978-981-16-6116-7_5.

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Hill, Robert N., Christopher Grandy, and Hussein Khalil. "Generation-IV Sodium-Cooled Fast Reactors (SFR)." In Nuclear Energy Encyclopedia, 353–64. Hoboken, NJ, USA: John Wiley & Sons, Inc., 2011. http://dx.doi.org/10.1002/9781118043493.ch30.

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Raj, Baldev, P. R. Vasudeva Rao, P. Puthiyavinayagam, and K. Ananthasivan. "Advanced Ceramic Fuels for Sodium-Cooled Fast Reactors." In Handbook of Advanced Ceramics and Composites, 1–36. Cham: Springer International Publishing, 2019. http://dx.doi.org/10.1007/978-3-319-73255-8_21-1.

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Raj, Baldev, P. R. Vasudeva Rao, P. Puthiyavinayagam, and K. Ananthasivan. "Advanced Ceramic Fuels for Sodium-Cooled Fast Reactors." In Handbook of Advanced Ceramics and Composites, 667–702. Cham: Springer International Publishing, 2020. http://dx.doi.org/10.1007/978-3-030-16347-1_21.

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Yamano, Hidemasa, Hiroyuki Nishino, Yasushi Okano, Takahiro Yamamoto, and Takashi Takata. "Development of Risk Assessment Methodology Against External Hazards for Sodium-Cooled Fast Reactors." In Earthquakes, Tsunamis and Nuclear Risks, 111–21. Tokyo: Springer Japan, 2016. http://dx.doi.org/10.1007/978-4-431-55822-4_9.

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Conference papers on the topic "Sodium cooled fast reactors"

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Paumel, Kevin. "Subassembly identification by ultrasound in sodium cooled fast reactors." In 2018 IEEE International Instrumentation and Measurement Technology Conference (I2MTC ). IEEE, 2018. http://dx.doi.org/10.1109/i2mtc.2018.8409591.

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Song, Wei, Hongyi Yang, Chunming Zhang, and Jiaxu Zuo. "Discussion on the Application of Fire Probability Safety Assessment to Sodium Cooled Fast Reactor." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-15780.

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As sodium cooled fast reactor (SFR) uses liquid sodium as coolant, the risk of sodium fire is brought to reactor safety, which is different from conventional fire. Sodium is very chemically active, and violent chemical reactions can happen when sodium is exposed to air or water. Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is probably one of the main contributors to the total reactor risk. In this paper, the methodology of fast reactor sodium fire risk assessment is studied, and the principles and procedure of sodium fire probabilistic safety assessment (PSA) are given. The application of this technology in China Experimental Fast Reactor (CEFR) is explored, and several key problems which need more research in the future during the process of sodium fire probabilistic safety assessment are discussed.
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Asayama, Tai, Yugi Nagae, Takashi Wakai, Kazuyuki Tsukimori, and Masaki Morishita. "Structural Materials and Code Development for Japanese Sodium-Cooled Fast Reactors." In ASME 2014 Symposium on Elevated Temperature Application of Materials for Fossil, Nuclear, and Petrochemical Industries. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/etam2014-1037.

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This paper describes the latest status on the development of elevated temperature materials and structural codes for Japanese sodium-cooled fast reactors (SFRs). Based on the extensive research and development activities in the last decades in Japan, two materials, 316FR and Modified 9Cr-1Mo steels were recently incorporated into the 2012 Edition of Fast Reactor Design and Construction Code of the Japan Society of Mechanical Engineers (JSME). Structural design methodologies are continuously being improved towards the next major revision planed in 2016 Edition where methodologies for a 60-year design of Japanese demonstration fast reactor will be provided. Codes and guidelines for fitness-for-service, leak-before-break evaluation and reliability assessment are concurrently being developed utilizing the System Based Code concept aiming at establishing an integrated code system that encompasses a life cycle of SFRs. Paper published with permission.
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Nagaraj, C. P., M. Sivaramakrishna, K. Madhusoodanan, and P. Chellapandi. "Design of Neutron Flux Monitoring System for sodium cooled fast reactors." In 2013 3rd International Conference on Advancements in Nuclear Instrumentation, Measurement Methods and their Applications (ANIMMA). IEEE, 2013. http://dx.doi.org/10.1109/animma.2013.6727879.

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Morgan, Sarah, Sama Bilbao y Leon, Matthew Bucknor, Mark Anderson, Emilio Baglietto, James Schneider, Matthew Weathered, and Liangyu Xu. "Thermal Stratification Modeling for Sodium-Cooled Fast Reactors: A Status Update." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-82364.

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Thermal hydraulic behavior in the upper plenum of pool-type sodium-cooled fast reactors (SFRs) is a major concern, as many design challenges are concentrated in this region. As SFR designs aim for licensing and commercialization, it is important to accurately analyze and predict the thermal-hydraulic behavior in this region during accident scenarios, specifically thermal stratification. Thermal stratification models are currently a major source of uncertainty in most system codes for all types of power plants. Most system codes, including SAS4A/SASSYS-1, a system level code developed by Argonne National Laboratory (Argonne), use very coarse meshes that cannot capture the complexities of the stratification phenomena. While the commonly employed lumped-volume based models for thermal stratification are able to run in a matter of seconds, they result in approximate results and can only handle simple cases. Other 2-D and 3-D methods, such as computational fluid dynamics (CFD) models, can analyze simple configurations with higher fidelity, but come with a relatively large computational expense. Finding a modeling solution that is both accurate and computationally efficient has proven difficult. This paper provides details of a review and gap analysis of the various modeling approaches proposed to date and explores a path forward for future thermal stratification modeling efforts, with a focus on developing new models for the SAS4A/SASSYS-1 system code.
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Usov, Eduard V., Ivan G. Kudashov, Sergey A. Zhigach, Anton A. Butov, Nikolay A. Pribaturin, Sergey I. Lezhnin, Ruslan V. Chalyy, Sergey E. Yakush, and Uliya Vinogradova. "Coupled Code SOCRAT-BN Development for Safety Analysis of Sodium-Cooled Fast Reactors." In 2012 20th International Conference on Nuclear Engineering and the ASME 2012 Power Conference. American Society of Mechanical Engineers, 2012. http://dx.doi.org/10.1115/icone20-power2012-54316.

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SOCRAT-BN is a software package to simulate design and severe accidents of sodium-cooled fast reactors. The package consists of modules for calculating damage to the reactor’s core, thermohydraulic processes and neutron physics. The thermohydraulic module has been developed to calculate one- and two-phase flows in channels with different geometry and bundles. The module is based on a two-fluid model for equal pressures of phases. In this paper we present an explanation of the deciding constitutive models for equations used in the system. Validation of the module was performed on the experimental data for one- and two-fluid flows in complex geometry channels and on calculation of running a first loop of the reactor BN-600 in nominal mode.
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Mikityuk, K., A. Vasiliev, P. Fomichenko, T. Schepetina, S. Subbotin, and P. Alekseev. "RBEC-M Lead-Bismuth Cooled Fast Reactor: Optimization of Conceptual Decisions." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22329.

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A concept of the RBEC lead-bismuth fast reactor is a synthesis, on one hand, experience in development and operation of fast sodium power reactors and reactors with Pb-Bi coolant, and, on the other hand, of large R&D activities on development of the core concept for modified fast sodium reactor. The paper presents improved project RBEC-M, characterized by a number of innovative decisions, which allow to improve safety and cost parameters compared to the basic RBEC project. These innovative decisions include application of nitride fuel based on 15N, two-circuit scheme without main reactor pumps, gas lift system in the primary circuit, passive reactor auxiliary cooling system, etc.
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Cavaro, M., C. Payan, and J. P. Jeannot. "Towards the characterization of the bubble presence in liquid sodium of Sodium cooled Fast Reactors." In 2013 3rd International Conference on Advancements in Nuclear Instrumentation, Measurement Methods and their Applications (ANIMMA). IEEE, 2013. http://dx.doi.org/10.1109/animma.2013.6727947.

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Joseph, Jofred, Satish Kumar, Tanmay Vasal, and N. Theivarajan. "Brayton Cycle As an Alternate Power Conversion Option for Sodium Cooled Fast Reactor." In ASME 2019 Gas Turbine India Conference. American Society of Mechanical Engineers, 2019. http://dx.doi.org/10.1115/gtindia2019-2455.

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Abstract Enhancing the safety and economic competitiveness are major objectives in the development of advanced reactor designs with emphasis on the design of systems or components of the nuclear systems. Innovative power cycle development is another potential option to achieve these objectives. Sodium cooled fast reactor (SFR) is one among the six reactor design concepts identified by the Gen IV International Forum for development to meet the technology goals for new nuclear energy system. Similar to the power cycle used in conventional fossil fuel based thermal power plants, sodium-cooled fast reactors have adopted the Rankine cycle based power conversion system. However, the possibility of sodium water reaction is a major concern and it becomes necessary to adopt means of early detection of leaks and isolation of the affected SG module for mitigating any adverse impact of sodium water reaction. The high exothermic nature of the reaction calls for introducing an intermediate sodium heat transport loop, leading to high overall plant cost hindering commercialization of sodium fast reactors. The Indian Prototype Fast Breeder Reactor (PFBR) also uses Rankine cycle in the power generation system. The superheated steam temperature has been set at 490 degree Celsius based on optimisation studies and material limitations. Additional Fast Breeder reactors are planned in near future and further work is being done to develop more advanced sodium cooled fast reactors. The closed Brayton cycle is a promising alternative to conventional Rankine cycle. By selecting an inert gas or a gas with milder reaction with sodium, the vigorous sodium water reaction can be avoided and significant cost savings in the turbine island can be achieved as gas turbine power conversion systems are of much smaller size than comparable steam turbine systems due to their higher power density. In the study, various Brayton cycle designs on different working gases have been explored. Supercritical-CO2 (s-CO2), helium and nitrogen cycle designs are analyzed and compared in terms of cycle efficiency, component performance and physical size. The thermal efficiencies at the turbine inlet temperature of Indian PFBR have been compared for Rankine cycle and Brayton cycle based on different working fluids. Also binary mixtures of different gases are investigated to develop a more safe and efficient power generation system. Helium does not interact with sodium and other structural materials even at very high temperatures but its thermal performance is low when compared to other fluids. Nitrogen being an inert gas does not react with sodium and can serve to utilise existing turbomachinery because of the similarity with atmospheric air. The supercritical CO2 based cycle has shown best thermodynamic performance and efficiency when compared to other Brayton cycles for the turbine inlet temperature of Indian PFBR. CO2 also reacts with sodium but the reaction is mild compared to sodium water reaction. The cycle efficiency of the s-CO2 cycle can be further improved by adopting multiple reheating, inter cooling and recuperation.
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Kawasaki, Nobuchika, Tomoyoshi Watakabe, Takashi Wakai, Tomohiko Yamamoto, Tsuyoshi Fukasawa, and Shigeki Okamura. "Development of Seismic Isolation Systems for Sodium-Cooled Fast Reactors in Japan." In ASME 2016 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/pvp2016-63437.

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Sodium-cooled Fast Reactors (SFRs) have components with thinner walls as compared with light water reactors, although Japan is an earthquake-prone country. Thus, seismic isolation systems have been conventionally employed in SFR system design to reduce seismic forces on the systems in Japan. Implementation of seismic design in the reactor core and buckling design in the reactor vessel requires 8 Hz (or less) vertical frequency’s isolation system being applied. This paper introduces three isolation concepts to achieve the frequency. The isolation systems, which enable vertical 8 Hz natural frequency, comprise thicker laminated rubber bearings (TRBs). By combining coned disk springs with TRBs, vertical natural frequency is in a range from roughly 3 Hz to 5 Hz. Combining pneumatic springs to RBs and adding the rocking suppression system, vertical natural frequency becomes under 1 Hz. All isolation systems need horizontal damping like oil dampers. A vertical 8 Hz isolation system with TRBs and oil dampers is under development in Japan as a principal isolation concept. This system was selected because of its simple configuration and fewer issues to be resolved in the development. Since TRBs and oil dampers are basic isolation elements, they can be applied to other isolation systems. The response acceleration of 5 Hz vertical isolation is 50% of that of 8 Hz based on the analytical survey. A series of static tests of coned disk springs was carried out to confirm design equations. Based on these knowledge, 5 Hz vertical isolation system with TRBs and the coned disk springs can be designed. The response acceleration of 1 Hz vertical isolation is 10% of that of 8 Hz. A rocking suppression system was studied in the past, and the further simplification of this system is the largest challenge for this concept. These three isolation concepts are isolation candidates for SFRs in Japan. To obtain enough seismic margins for each plant site, these isolation systems need to be developed.
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Reports on the topic "Sodium cooled fast reactors"

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Chien, Hual-Te, James Sienicki, Thomas Elmer, and William Lawrence. In-Sodium Testing of a Prototype Thermoacoustic Power Sensor for Sodium-Cooled Fast Reactors. Office of Scientific and Technical Information (OSTI), September 2020. http://dx.doi.org/10.2172/1808318.

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Thomas, Justin W., Heather M. Connaway, James J. Grudzinski, Yu Tang, and Surya Narayana Prasad Vegendla. End-User Evaluation of SHARP for Analysis of Sodium-Cooled Fast Reactors. Office of Scientific and Technical Information (OSTI), September 2014. http://dx.doi.org/10.2172/1179453.

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Kamal, S., C. Grandy, M. Farmer, and A. Brunsvold. High strength and heat resistant chromium steels for sodium-cooled fast reactors. Office of Scientific and Technical Information (OSTI), December 2004. http://dx.doi.org/10.2172/836536.

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Belles, Randy, Willis Poore III, Nicholas R. Brown, George F. Flanagan, Mark Holbrook, Wayne Moe, and Tanju Sofu. Proposed Adaptation of the Standard Review Plan NUREG-0800, Chapter 4 (Reactor) for Sodium-Cooled Fast Reactors and Modular High-Temperature Gas-Cooled Reactors. Office of Scientific and Technical Information (OSTI), March 2017. http://dx.doi.org/10.2172/1492189.

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Belles, Randy, Willis P. Poore, III, Nicholas R. Brown, George F. Flanagan, Mark Holbrook, Wayne Moe, and Tanju Sofu. 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. Office of Scientific and Technical Information (OSTI), March 2017. http://dx.doi.org/10.2172/1361360.

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Jean Ragusa, Andrew Siegel, and Jean-Michel Ruggieri. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations. Office of Scientific and Technical Information (OSTI), September 2010. http://dx.doi.org/10.2172/992636.

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Youinou, Gilles J., Samuel E. Bays, and Giuseppe Palmiotti. Scoping Analysis of Sodium Cooled Fast Spectrum Test Reactor Cores. Office of Scientific and Technical Information (OSTI), January 2020. http://dx.doi.org/10.2172/1598332.

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Brown, Nicholas R., William David Pointer, Matt Sieger, George F. Flanagan, Wayne Moe, and Mark HolbrookINL. Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations. Office of Scientific and Technical Information (OSTI), April 2016. http://dx.doi.org/10.2172/1254091.

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Moe, Wayne, and George Honma. Implementation Plan for Qualification of Sodium-Cooled Fast Reactor Technology Information. Office of Scientific and Technical Information (OSTI), March 2016. http://dx.doi.org/10.2172/1260447.

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Cipiti, Benjamin. Sodium-Cooled Fast Reactor Proliferation Resistance and Physical Protection White Paper. Office of Scientific and Technical Information (OSTI), November 2020. http://dx.doi.org/10.2172/1710232.

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