Bibliographies thématiques / NUCLEAR REACTOR SYSTEMS

Littérature scientifique sur le sujet « NUCLEAR REACTOR SYSTEMS »

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Publié le 11 septembre 2023

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Articles de revues sur le sujet "NUCLEAR REACTOR SYSTEMS"

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Petti, D., D. Crawford et N. Chauvin. « Fuels for Advanced Nuclear Energy Systems ». MRS Bulletin 34, no 1 (janvier 2009) : 40–45. http://dx.doi.org/10.1557/mrs2009.11.

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AbstractFuels for advanced nuclear reactors differ from conventional light water reactor fuels and also vary widely because of the specific architectures and intended missions of the reactor systems proposed to deploy them. Functional requirements of all fuel designs for advanced nuclear energy systems include (1) retention of fission products and fuel nuclides, (2) dimensional stability, and (3) maintenance of a geometry that can be cooled. In all cases, anticipated fuel performance is the limiting factor in reactor system design, and cumulative effects of increased utilization and increased exposure to inservice environments degrade fuel performance. In this article, the current status of each fuel system is reviewed, and technical challenges confronting the implementation of each fuel in the context of the entire advanced reactor fuel cycle (fabrication, reactor performance, recycle) are discussed.
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Caciuffo, R., C. Fazio et C. Guet. « Generation-IV nuclear reactor systems ». EPJ Web of Conferences 246 (2020) : 00011. http://dx.doi.org/10.1051/epjconf/202024600011.

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In this paper, we provide a concise description of the six nuclear reactor concepts that are under development in the framework of the Generation-IV International Forum. After a brief introduction on the world energy needs, its plausible evolution during the next fifty years, and the constraints imposed by the necessity to address the climate challenges we are facing today, we will present the main features of the innovative nuclear energy systems that hold the promise to produce almost-zero-carbon-emission electricity, heat for chemistry and industrial manufacturing, hydrogen to be used as energy vector, and affordable freshwater. Potential advantages over currently available nuclear systems in terms of increased safety, reduced proliferation risks, economical affordability, sustainability of the fuel cycle, and management of the waste inventory will be critically discussed against the technical challenges that remain to be overcome.
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Large, J. H. « Decommissioning of Nuclear Reactor Systems ». Proceedings of the Institution of Mechanical Engineers, Part A : Journal of Power and Energy 206, no 4 (novembre 1992) : 273–80. http://dx.doi.org/10.1243/pime_proc_1992_206_044_02.

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The decision-making process involving the decommissioning of the British graphite-moderated, gas-cooled Magnox power stations is complex. There are timing, engineering, waste disposal, cost and lost generation capacity factors and the ultimate uptake of radiation dose to consider and, bearing on all of these, the overall decision of when and how to proceed with decommissioning may be heavily weighed by political and public tolerance dimensions. These factors and dimensions are briefly reviewed with reference to the ageing Magnox nuclear power stations, of which Berkeley and Hunterston A are now closed down and undergoing the first stages of decommissioning and Trawsfynydd, although still considered as available capacity, has had both reactors closed down since February 1991 and is awaiting substantiation and acceptance of a revised reactor pressure vessel safety case. Although the other first-generation Magnox power stations at Hinkley Point, Bradwell, Dungeness and Sizewell are operational, it is most doubtful that these stations will be. able to eke out a generating function for much longer. It is concluded that the British nuclear industry has adopted a policy of deferred decommissioning, that is delaying the process of complete dismantlement of the radioactive components and assemblies for at least one hundred years following close-down of the plant. In following this option the nuclear industry has expressed considerable confidence that the decommissioning technology required will he developed with passing time, that acceptable radioactive waste disposal methods and facilities will be available and that the eventual costs of decommissioning will not escalate without restraint.
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Hiscox, Briana, Benjamin Betzler, Vladimir Sobes et William J. Marshall. « NEUTRONIC BENCHMARKING OF SMALL GAS-COOLED SYSTEMS ». EPJ Web of Conferences 247 (2021) : 10033. http://dx.doi.org/10.1051/epjconf/202124710033.

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To demonstrate that nuclear reactors can be built faster and more economically than they have been in the past, the US Department of Energy Office of Nuclear Energy is sponsoring the development of a small nuclear reactor called the Transformational Challenge Reactor (TCR) [1–2]. An important part of the design and licencing process of a new reactor is validation of the software used to analyze the reactor using established reactor physics benchmarks. This paper discusses validation of the neutronics software used to model four preliminary designs of the TCR core [2]. Because the TCR core design uses innovative technology and methods, comparable established benchmarks are limited or do not exist. For this effort, established benchmarks from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) [3] were considered to be suitable for this design based on analysis using the SCALE/TSUNAMI-computed similarity indices to determine the amount of shared uncertainty between the design and each selected benchmark experiment. This paper addresses the challenges faced in benchmarking a unique reactor for licensing and construction, a task that will become more common as a new generation of innovative nuclear reactors are designed and built.
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Saha, Dilip, et John Cleveland. « Natural Circulation in Nuclear Reactor Systems ». Science and Technology of Nuclear Installations 2008 (2008) : 1. http://dx.doi.org/10.1155/2008/932319.

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Moore, Julian. « Upgrading of nuclear reactor display systems ». Electronics and Power 32, no 10 (1986) : 735. http://dx.doi.org/10.1049/ep.1986.0429.

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Bonal, Jean-Pierre, Akira Kohyama, Jaap van der Laan et Lance L. Snead. « Graphite, Ceramics, and Ceramic Composites for High-Temperature Nuclear Power Systems ». MRS Bulletin 34, no 1 (janvier 2009) : 28–34. http://dx.doi.org/10.1557/mrs2009.9.

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AbstractThe age of nuclear power originated with the gas-cooled, graphite-moderated reactor in the 1940s. Although this reactor design had intrinsic safety features and enjoyed initial widespread use, gas-cooled reactor technology was supplanted by higher power density water-cooled systems in the 1960s. However, the next-generation reactors seek enhanced power conversion efficiency and the ability to produce hydrogen, best accomplished with high-temperature gas-cooled systems. Thus, international interest in gas-cooled reactor systems is reemerging. Although the materials systems of these reactors are fairly simple, the reactor environment, particularly its high temperatures and intense irradiation, present extreme challenges in terms of material selection and survivability. This article provides a brief review of materials issues and recent progress related to graphite and ceramic materials for application in gas-cooled nuclear reactor environments. Of particular interest are the drastic, irradiation-induced microstructural evolution and thermophysical property changes occurring as a result of energetic neutron irradiation, which significantly impact the performance and lifetime of much of the reactor core. For “nuclear” graphite, the performance and lifetime not only are closely related to the irradiation environment but also are dramatically affected by the specifics of the particular graphite: manufacturing process, graphitization temperature, composition (amount of coke, filler, etc., depending on where it was mined), and so on. Moreover, the extreme environmental challenges set down by this next generation of fission nuclear plants have driven the development and application of ceramic composites for critical components, pushing beyond upper temperature limits set by metallic alloys used in previous generations of nuclear reactors. The composite material systems of particular interest are continuous carbon-fiber composites and newly developed radiation-resistant silicon carbide fiber composites.
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Kondylakis, J. S. « Theoretically and under very special applied conditions a nuclear fission reactor may explode as nuclear bomb ». HNPS Proceedings 18 (23 novembre 2019) : 121. http://dx.doi.org/10.12681/hnps.2558.

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This article/presentation describes a theoretical and applied research in nuclear fission reactor systems. It concerns with theoretical approaches and in very special applied cases consideration where a common nuclear fission reactor system may be considered to explode as nuclear bomb. This research gives critical impacts to the design, operation, management and philosophy of nuclear fission reactors systems. It also includes a sensitivity analysis of a particular applied problem concerning the core melting of a nuclear reactor and its deposit to the bottom of reactor vessel. Specifically, in a typical nuclear fission power reactor system of about 1000 MWe, the nuclear core material (corium) in certain cases can be melted and it may deposited in the bottom of nuclear reactor vessel. So, the nuclear criticality conditions are evaluated for a particular example case(s). Assuming an example composition of melted corium of 98 tones of U238 , 1 tone of U235 , 1 tone Pu239 and 25 tones Fe56 (supporting material) in a 5 m diameter of a finite cylindrical nuclear reactor vessel it is found that it may result in nuclear criticality above the unit. This condition corresponds to Supercritical Fast Nuclear Fission Reactor case, which may under certain very special applied conditions to nuclear explode as nuclear bomb.
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Sutopo, Catur Febriyanto, et Arifin M. Susanto. « Kajian pembentukan peraturan mengenai sistem pendingin reaktor dan sistem terkait untuk reaktor berpendingin gas ». Jurnal Pengawasan Tenaga Nuklir 1, no 2 (15 décembre 2021) : 11–19. http://dx.doi.org/10.53862/jupeten.v1i2.014.

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IN 2021, BAPETEN, AS THE REGULATORY BODY, IS ESTABLISHING A BAPETEN REGULATION REGARDING THE REACTOR COOLANT SYSTEM AND RELATED SYSTEMS, WHICH CURRENTLY ARE NOT YET AVAILABLE. Therefore, it is crucial to establish the BAPETEN Regulation. Based on the reasons, before setting the BAPETEN Regulation, it is necessary to conduct a study that is expected to provide a more comprehensive description and provide recommendations on what things need to be regulated in the BAPETEN Regulation, especially for gas-cooled reactors. The method used in this study is a literature study from various relevant references. The result of this study is that it is essential to require a capacity of the ultimate heat sink, including the spent nuclear fuel storage pool and a minimum period of the ability of the top heat sink in the accident analysis if the decay heat in the storage pool and the residual heat in the reactor core fail simultaneously. On the other hand, it is also necessary to require a margin of uncertainty to evaluate a situation and take corrective action. Likewise, independent and redundant access to the ultimate heat sink is needed to increase reliability. As for gas-cooled reactors, it is required to adapt the terms used. In addition, it is necessary to determine the appropriate definition because some of the terms used in water-cooled reactors have the same terms as gas-cooled reactors but have different functions. Keywords: Regulatory assessment, coolant system, related systems, gas-cooled reactors
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Obaidurrahman, Khalilurrahman, et Om Singh. « A comparative study of kinetics of nuclear reactors ». Nuclear Technology and Radiation Protection 24, no 3 (2009) : 167–76. http://dx.doi.org/10.2298/ntrp0903167o.

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The paper deals with the study of reactivity initiated transients to investigate major differences in the kinetics behavior of various reactor systems under different operating conditions. The article also states guidelines to determine the safety limits on reactivity insertion rates. Three systems, light water reactors (pressurized water reactors), heavy water reactors (pressurized heavy water reactors), and fast breeder reactors are considered for the sake of analysis. The upper safe limits for reactivity insertion rate in these reactor systems are determined. The analyses of transients are performed by a point kinetics computer code, PKOK. A simple but accurate method for accounting total reactivity feedback in kinetics calculations is suggested and used. Parameters governing the kinetics behavior of the core are studied under different core states. A few guidelines are discussed to project the possible kinetics trends in the next generation reactors.
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Thèses sur le sujet "NUCLEAR REACTOR SYSTEMS"

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Anadani, Mohamed. « Decision support systems for nuclear reactor control ». Thesis, University of Sheffield, 2000. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.341828.

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Presby, Andrew L. « Thermophotovoltaic energy conversion in space nuclear reactor power systems ». Thesis, Monterey, Calif. : Naval Postgraduate School, 2004. http://edocs.nps.edu/npspubs/scholarly/theses/2004/Dec/04Dec%5FPresby.pdf.

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Thesis (Astronautical Engineer and M. S. in Astronautical Engineering)--Naval Postgraduate School, December 2004.
Thesis Advisor(s): Gopinath, Ashok ; Michael, Sherif. "December 2004." Description based on title screen as viewed on March 13, 2009. Includes bibliographical references (p. 123-127). Also available in print.
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CARVALHO, LUIZ S. « Frequencia de danos no nucleo por blecaute em reator nuclear de concepcao avancada ». reponame:Repositório Institucional do IPEN, 2004. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11147.

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Kim, Choong Seok. « Reliability assessment of pressurized water reactor auxiliary feedwater systems ». Diss., Georgia Institute of Technology, 1985. http://hdl.handle.net/1853/13374.

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Persson, Carl-Magnus. « Reactivity Assessment in Subcritical Systems ». Licentiate thesis, Stockholm : Fysiska institutionen, Kungliga Tekniska högskolan, 2007. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-4363.

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Witter, Jonathan Keay. « Modeling for the simulation and control of nuclear reactor rocket systems ». Thesis, Massachusetts Institute of Technology, 1993. http://hdl.handle.net/1721.1/12755.

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Wu, Xiao. « Design of a Tritium Mitigation and Control System for Fluoride-salt-cooled High-temperature Reactor Systems ». The Ohio State University, 2016. http://rave.ohiolink.edu/etdc/view?acc_num=osu1452249907.

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Johnson, Kyle D. « High Performance Fuels for Water-Cooled Reactor Systems ». Doctoral thesis, KTH, Reaktorfysik, 2016. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-201604.

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Investigation of nitride fuels and their properties has, for decades, been propelled on the basis of their desirable high metal densities and high thermal conductivities, both of which oer intrinsic advantages to performance, economy, and safety in fast and light water reactor systems. In this time several key obstacles have been identied as impeding the implementation of these fuels for commercial applications; namely chemical interactions with air and steam, the noted diculty in sintering of the material, and the high costs associated with the enrichment of 15N. The combination of these limitations, historically, led to the well founded conclusion that the most appropriate use of nitride fuels was in the fast reactor fuel cycle, where the cost burdens associated with them is substantially less. Indeed, it is within this context that the vast majority of work on nitrides has been and continues to be done. Nevertheless, following the 2011 Fukushima-Daiichi nuclear accident, a concerted governmental-industrial eort was embarked upon to explore the alternatives of so-called \accident tolerant" and \high performance" fuels. These fuels would, at the same time, improve the response of the fuel-clad system to severe accidents and improve the economy of operation for light water reactor systems. Among the various candidates proposed are uranium nitride, uranium silicide, and a third \uranium nitride-silicide" composite featuring a mixture of the former. In this thesis a method has been established for the synthesis, fabrication, and characterization of high purity uranium nitride, and uranium nitride-silicide composites, prepared by the spark plasma sintering (SPS) technique. A specic result has been to isolate the impact of the processing parameters on the microstructure of representative fuel pellets, essentially permitting any conceivable microstructure of interest to be fabricated. This has enabled the development of a highly reproducible technique for the production of pellets with microstructures tailored towards any desired porosity between 88-99.9%TD, any grain size between 6-24 μm, and, in the case of  the uranium nitride-silicide composite, a silicide-coated UN matrix. This has permitted the evaluation of these microstructural characteristics on the performance of these materials, specically with respect to their role as accident tolerant fuels. This has generated results which have tightly coupled nitride performance with pellet microstructure, with important implications for the use of nitrides in water-cooled reactors.
Under artionden har forskning om nitridbranseln och dess egenskaper bedrivits pa grundval av nitridbransletsatravarda egenskaper avseende dess hoga metall tathet och hog varmeledningsformaga. Dessa egenskaper besitter vasentliga fordelar avseende prestanda, ekonomi och sakerhet for metallkylda som lattvatten reaktorer. Genom forskning har aven centrala begr ansningar identierats for implementering av nitridbranslen for kommersiellt bruk. Begransningar avser den kemiska interaktionen med luft och vattenanga, en uppmarksammad svarighet att sintring av materialet samt hoga kostnader forknippade med den nodvandiga anrikningen av 15-N. Kombinationen av dessa begransningar resulterade, tidigare, i en valgrundad slutsats att nitridbranslet mest andamalsenliga anvandningsomrade var i karnbranslecykeln for snabba reaktorer. Detta da kostnaderna forenade med implementeringen av branslet ar avsevart lagre. Inom detta sammanhang har majoriteten av forskning avseende nitrider bedrivits och fortskrider an idag. Dock, efter karnkraftsolyckan i Fukushima-Daiichi 2011, inleddes en samlad industriell och statlig anstrangning for att undersoka alternativ till sa kallade \olyckstoleranta" och \hogpresterande" branslen. Dessa branslen skulle samtidigt forbattra reaktionstiden for bransleinkapsling systemet mot allvarliga olyckor samt forbattra driftsekonomin av lattvattenreaktorer. Foreslagna kandidater ar urannitrid, uransilicid och en tredje \uran nitrid-silicid", komposit bestaende av en blandning av de foregaende. Genom denna avhandling har en metod faststallts for syntes, tillverkning och karaktarisering av uran nitrid av hog renhet samt uran nitrid-silicid kompositer, forberedda med tekniken SPS (Spark Plasma Sintering). Ett specikt resultat har varit att isolera eekten av processparametrar pa mikrostrukturen pa representativa branslekutsar. Detta mojliggor, i princip, framstallningen av alla tankbara mikrostrukturer utav intresse for tillverkning. Vidare har detta mojliggjort utvecklingen av en hogeligen reproducerbar  teknik for framstallningen av branslekutsar med mikrostrukturer skraddarsydda for onskad porositet mellan 88 och 99.9 % TD, och kornstorlek mellan 6 och 24 μm. Dartill har en metod for att belagga en matris av uran nitrid-silicid framarbetats. Detta har mojliggjort utvarderingen av dessa mikrostrukturella parametrars paverkan pa materialens prestanda, sarskilt avseende dess roll som olyckstoleranta branslen. Detta har genererat resultat som ar tatt sammanlankat nitridbranslets prestanda till kutsens mikrostruktur, med viktiga konsekvenser for den potentiella anvandningen av nitrider i lattvatten reaktorer.

QC 20170210

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CONCEICAO, JUNIOR OSMAR. « Aplicacao da tecnica de analise de modos de falha e efeitos ao sistema de resfriamento de emergencia de uma instalacao nuclear experimental ». reponame:Repositório Institucional do IPEN, 2009. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9367.

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IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Morrison, Jonathan J. « Corrosion, transport, and deposition in pressurised water nuclear reactor primary coolant systems ». Thesis, University of Birmingham, 2016. http://etheses.bham.ac.uk//id/eprint/6816/.

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Several unscheduled shut downs of the Cruas nuclear power plant in France have been caused by the deposition of corrosion products in flow broaches of the steam generator tube support sheets. The depositions are theorised to be the result of electrokinetically stimulated deposition. In this work, a hot water loop to replicate these depositions in the laboratory was built, along with rigs to characterise supporting phenomena – the corrosion rate of stainless steel and the solubility of the corrosion products. While the data obtained from the hot water loop did not provide conclusive proof of the existence or prevalence of the electrokinetically stimulated deposition mechanism, evidence of deposition caused by cavitation was found. The corrosion rate of stainless steel was measured at high temperatures in solutions of lithium hydroxide at various concentrations. Surface finish was found to have an effect on the corrosion rate, though the difference between mechanically ground surfaces with an order of magnitude difference in roughness was found to be minimal. The solubility of the corrosion products formed was measured and found to be of similar order to that reported in the literature, however the minor alloying elements were found to leach from the surface in substantial quantities.
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Livres sur le sujet "NUCLEAR REACTOR SYSTEMS"

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Glasstone, Samuel. Nuclear Reactor Engineering : Reactor Systems Engineering. Boston, MA : Springer US, 1994.

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Zohuri, Bahman. Neutronic Analysis For Nuclear Reactor Systems. Cham : Springer International Publishing, 2019. http://dx.doi.org/10.1007/978-3-030-04906-5.

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Zohuri, Bahman. Neutronic Analysis For Nuclear Reactor Systems. Cham : Springer International Publishing, 2017. http://dx.doi.org/10.1007/978-3-319-42964-9.

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Myers, Lynne C. Nuclear power systems : Their safety. [Ottawa] : Library of Parliament, Research Branch, 1989.

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Myers, Lynne C. Nuclear power systems : Their safety. [Ottawa] : Library of Parliament, Research Branch, 1996.

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United States. National Aeronautics and Space Administration., dir. Small space reactor power systems for unmanned solar system exploration missions. [Washington, D.C.] : National Areonautics and Space Administration, 1987.

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Lawrence, J. D. Software reliability and safety in nuclear reactor protection systems. Washington, DC : Division of Reactor Controls and Human Factors, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1993.

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Jalel, N. A. Expert systems for fault diagnosis in nuclear reactor control. Sheffield : University of Sheffield, Dept. of Control Engineering, 1990.

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Lawrence, Jim. Software reliability and safety in nuclear reactor protection systems. Washington, DC : Division of Reactor Controls and Human Factors, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1993.

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U.S. Nuclear Regulatory Commission. Division of Reactor Controls and Human Factors. et Lawrence Livermore National Laboratory, dir. Reviewing real-time performance of nuclear reactor safety systems. Washington, DC : Division of Reactor Controls and Human Factors, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, 1993.

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Chapitres de livres sur le sujet "NUCLEAR REACTOR SYSTEMS"

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Glasstone, Samuel, et Alexander Sesonske. « Power Reactor Systems ». Dans Nuclear Reactor Engineering, 759–90. Boston, MA : Springer US, 1994. http://dx.doi.org/10.1007/978-1-4615-2083-2_6.

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Glasstone, Samuel, et Alexander Sesonske. « Power Reactor Systems ». Dans Nuclear Reactor Engineering, 759–90. Boston, MA : Springer US, 1994. http://dx.doi.org/10.1007/978-1-4615-7525-2_13.

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Glasstone, Samuel, et Alexander Sesonske. « The Systems Concept, Design Decisions, and Information Tools ». Dans Nuclear Reactor Engineering, 487–500. Boston, MA : Springer US, 1994. http://dx.doi.org/10.1007/978-1-4615-2083-2_1.

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Glasstone, Samuel, et Alexander Sesonske. « The Systems Concept, Design Decisions, and Information Tools ». Dans Nuclear Reactor Engineering, 487–500. Boston, MA : Springer US, 1994. http://dx.doi.org/10.1007/978-1-4615-7525-2_8.

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Todreas, Neil E., et Mujid S. Kazimi. « Reactor Energy Distribution ». Dans Nuclear Systems Volume I, 71–122. Third edition. | Boca Raton : CRC Press, 2021- | : CRC Press, 2021. http://dx.doi.org/10.1201/9781351030502-3.

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Zohuri, Bahman. « Reactor Dynamics ». Dans Neutronic Analysis For Nuclear Reactor Systems, 435–57. Cham : Springer International Publishing, 2016. http://dx.doi.org/10.1007/978-3-319-42964-9_12.

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Zohuri, Bahman. « Reactor Stability ». Dans Neutronic Analysis For Nuclear Reactor Systems, 459–74. Cham : Springer International Publishing, 2016. http://dx.doi.org/10.1007/978-3-319-42964-9_13.

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Zohuri, Bahman. « Reactor Dynamics ». Dans Neutronic Analysis For Nuclear Reactor Systems, 427–49. Cham : Springer International Publishing, 2019. http://dx.doi.org/10.1007/978-3-030-04906-5_12.

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Zohuri, Bahman. « Reactor Stability ». Dans Neutronic Analysis For Nuclear Reactor Systems, 451–65. Cham : Springer International Publishing, 2019. http://dx.doi.org/10.1007/978-3-030-04906-5_13.

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Todreas, Neil E., Mujid S. Kazimi et Mahmoud Massoud. « Fundamentals of Reactor Transient Simulation ». Dans Nuclear Systems Volume II, 443–75. 2e éd. Boca Raton : CRC Press, 2021. http://dx.doi.org/10.1201/9780429157608-12.

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Actes de conférences sur le sujet "NUCLEAR REACTOR SYSTEMS"

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Yamamoto, Takahisa, Koshi Mitachi et Masatoshi Nishio. « Reactor Controllability of 3-Region-Core Molten Salt Reactor System : A Study on Load Following Capability ». Dans 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89440.

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The Molten Salt Reactor (MSR) systems are liquid-fueled reactors that can be used for actinide burning, production of electricity, production of hydrogen, and production of ssile fuels (breeding). Thorium (Th) and uranium-233 (233U) are fertile and ssile of the MSR systems, and dissolved in a high-temperature molten fluoride salt (fuel salt) with a very high boiling temperature (up to 1650K), that is both the reactor nuclear fuel and the coolant. The MSR system is one of the six advanced reactor concepts identified by the Generation IV International Forum (GIF) as a candidate for cooperative development [1]. In the MSR system, fuel salt flows through a fuel duct constructed around a reactor core and fuel channel of a graphite moderator accompanied by fission reaction and heat generation, and flows out to an external-loop system consisted of a heat exchanger and a circulation pump. Due to the motion of fuel salt, delayed neutron precursors that are one of the source of neutron production make to change their position between the ssion reaction and neutron emission events and decay even occur in the external loop system. Hence the reactivity and effective delayed neutron precursor fraction of the MSR system are lower than those of solid fuel reactor systems such as Boiling Water Reactors (BWRs) and Pressurised Water Reactor (PWRs). Since all of the presently operating nuclear power reactors utilize solid fuel, little attention had been paid to the MSR analysis of the reactivity loss and reactor characteristics change caused by the fuel salt circulation. Sides et al. [2] and Shimazu et al. [3] developed MSR analytical models based on the point reactor kinetics model to consider the effect of fuel salt flow. Their models represented a reactor as having six zones for fuel salt and three zones for the graphite moderator. Since their models employed the point reactor kinetics model and the rough temperature approximation, their results were not sufficiently accurate to consider the effect of fuel salt flow.
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Bakhmach, Ievgenii S., Alexander A. Siora, Volodymyr T. Bezsalyi et Mikhail A. Yastrebenetsky. « Digital Systems for Reactor Control : Design, Experience of Operation ». Dans 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48205.

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Conversion of traditional analog NPP I&C systems to digital systems is a common tendency for many countries. Digital systems for reactor control designed by «Radiy» Company (Kirovograd, Ukraine) are described below. FPGA (Field Programmable Gates Arrays) were used for implementation of control algorithms. An equivalence between FPGA-projects implementation and schemes of control technological algorithms permitted to simplify development and verification processes and decrease the number of development errors. The platform was used for implementation of different safety important systems: reactor protection systems, automatic reactor power control and limitation systems, rods control systems, control safety systems. The main peculiarity of the reactor protection system is different types of diversity (apparatus and program diversity due to different hardware and different languages in main and diverse divisions; functional diversity; difference of CASE-tools). These systems have been used at 10 units. Reliability measures of systems and their components were determined using operational statistical data. Possible uses for these systems: modernization of different types of existing reactors (not only WWER); as full system or as subsystems; not only for Ukraine, but for other countries; for reactors III+ and IV generations.
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Yastrebenetsky, Mikhail A., Alexander A. Siora et Volodymyr I. Tokarev. « Reliability of Reactor Control Digital Systems ». Dans 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75157.

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In spite of wide expansion of digital systems for NPP control, information about realistic operating reliability measures of these systems is still lacking. The paper is a continuation of report [1] and contains results of analysis of operating reliability of digital control systems used in Ukrainian nuclear power plants. This paper contains: - reliability measures of digital instrumentation and control systems of first generation (designed in 1979–1983); - reliability measures of digital reactor control systems of second generation (designed after 2000) including reasons of failures; - analysis of the reasons of NPP violations due to digital instrumentation and control systems.
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Lawrence, Leo A., Bruce J. Makenas et Les L. Begg. « Performance of fast reactor irradiated fueled emitters for thermionic reactors ». Dans Proceedings of the ninth symposium on space nuclear power systems. AIP, 1992. http://dx.doi.org/10.1063/1.41886.

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Gese, Augustin, Michal Kopcek et Alojz Meszaros. « Signal filtering for a nuclear reactor restarting ». Dans 2012 6th IEEE International Conference Intelligent Systems (IS). IEEE, 2012. http://dx.doi.org/10.1109/is.2012.6335162.

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Kononov, V. N., M. V. Bokhovko, P. P. Dyachenko, V. V. Korobkin, Yu A. Prokhorov, V. I. Regushevskiv et V. N. Smolskiy. « Nuclear Reactor Pumped Lasing Experiments on Fast Burst Reactor ». Dans The European Conference on Lasers and Electro-Optics. Washington, D.C. : Optica Publishing Group, 1996. http://dx.doi.org/10.1364/cleo_europe.1996.ctud7.

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The concept of using nuclear energy to pump a laser by fission fragments of uranium can potentially be used to create large high power lasers that cannot be matched by other laser svstems[1]. That is possible owing to a high energy capacity of fission reactor and high penetration ability of neutrons in uranium multiplication systems. In the framework of designing of Energy Model of a Pulse Nuclear Reactor Pumped Laser System in IPPE[1] it was performed lasing experiments on 235U fission fragments pumping of 5d-6p atomic xenon 1.73 µm transition with 2 core fast burst reactor BARS-6 as a neutron source. The laser cell is a thin wall stainless steel tube 49 mm outer diameter and 400 mm lenth with internal 2.7 mg/cm2 235UO2 coating and filled Ar-Xe mixture (200: l) at 380 Torr. The laser cavity was formed by 6 m radius gold mirror and flat dielectric output coupler with reflectivity of 95% at 1.73 µm. The laser cell was surrounded by the 5 cm thick poliethylen moderator. In Fig. 1 thermal neutrons and laser output signal are shown. Thermal neutrons fluence was measured using Au radioactive indicator and was 1.7·1012 n/cm2. In Fig. 2 results of "master oscillator - two round trip amplifier" experiment are shown. As amplifier it was used a similar tube filled Ar-Xe mixture 2500 mm length with internal 5 µm thickness metal 235U coating and with optical windows. The energy gain in amplifier tube was ~ 4.
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Vijayakumaran, P. M., C. P. Nagaraj, C. Paramasivan Pillai, R. Ramakrishnan et M. Sivaramakrishna. « Nuclear Instrumentation Systems in Prototype Fast Breeder Reactor ». Dans 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49354.

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The nuclear instrumentation systems of the Prototype Fast Breeder Reactor (PFBR) primarily comprise of global Neutron Flux Monitoring, Failed Fuel Detection & Location, Radiation Monitoring and Post-Accident Monitoring. High temperature fission chambers are provided at in-vessel locations for monitoring neutron flux. Failed fuel detection and location is by monitoring the cover gas for fission gases and primary sodium for delayed neutrons. Signals of the core monitoring detectors are used to initiate SCRAM to protect the reactor from various postulated initiating events. Radiation levels in all potentially radioactive areas are monitored to act as an early warning system to keep the release of radioactivity to the environment and exposure to personnel well below the permissible limits. Fission Chambers and Gamma Ionisation Chambers are located in the reactor vault concrete for monitoring the neutron flux and gamma radiation levels during and after an accident.
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Deffrennes, Marc, Michel Hugon, Panagiotis Manolatos, Georges Van Goethem et Simon Webster. « Euratom Research Framework Programme on Reactor Systems ». Dans 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89502.

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The activities of the European Commission (EC) in the field of nuclear energy are governed by the Treaty establishing the European Atomic Energy Community (EURATOM). The research activities of the European Union (EU) are designed as multi-annual Framework Programmes (FP). The EURATOM 6th Framework Programme (EURATOM FP-6), covering the period 2002–2006, is funded with a budget of 1, 230 million Euros and managed by the European Commission. Beyond the general strategic goal of the EURATOM Framework Programmes to help exploit the potential of nuclear energy, in a safe and sustainable manner, FP-6 is designed to contribute also to the development of the “European Research Area” (ERA), a concept described in the Commission’s Communication COM(2000)6, of January 2000. Moreover EURATOM FP-6 contributes to the creation of the conditions for sharing the same nuclear safety culture throughout the EU-25 and the Candidate Countries, fostering the acceptance of nuclear power as an element of the energy mix. This paper gives an overview of the research activities undertaken through EURATOM FP-6 in the area of Reactor Systems, covering the safety of present reactors, the development of future safe reactors, and the needs in terms of research infrastructures and education & training. The actions under FP-6 are presented in their continuity of a ctions under FP-5. The perspectives under FP-7 are also provided. Other parts of the EURATOM FP, covering Waste Handling and Radiation Protection, as well as Fusion Energy, are not detailed in this paper.
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Cheatham, Jesse, Bao Truong, Nicholas Touran, Ryan Latta, Mark Reed et Robert Petroski. « Fast Reactor Design Using the Advanced Reactor Modeling Interface ». Dans 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16815.

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The Advanced Reactor Modeling Interface (ARMI) code system has been developed at TerraPower to enable rapid and robust core design. ARMI is a modular modeling framework that loosely couples nuclear reactor simulations to provide high-fidelity system analysis in a highly automated fashion. Using a unified description of the reactor as input, a wide variety of independent modules run sequentially within ARMI. Some directly calculate results, while others write inputs for external simulation tools, execute them, and then process the results and update the state of the ARMI model. By using a standardized framework, a single design change, such as the modification of the fuel pin diameter, is seamlessly translated to every module involved in the full analysis; bypassing error-prone multi-analyst, multi-code approaches. Incorporating global flux and depletion solvers, subchannel thermal-hydraulics codes, pin-level power and flux reconstruction methods, detailed fuel cycle and history tracking systems, finite element-based fuel performance coupling, reactivity coefficient generation, SASSYS-1/SAS4A transient modeling, control rod worth routines, and multi-objective optimization engines, ARMI allows “one click” steady-state and transient assessments throughout the reactor lifetime by a single user. This capability allows a user to work on the full-system design iterations required for reactor performance optimizations that has traditionally required the close attention of a multi-disciplinary team. Through the ARMI framework, a single user can quickly explore a design concept and then consult the multi-disciplinary team for model validation and design improvements. This system is in full production use for reactor design at TerraPower, and some of its capabilities are demonstrated in this paper by looking at how design perturbations in fast reactor core assemblies affect steady-state performance at equilibrium as well as transient performance. Additionally, the pin-power profile is examined in the high flux gradient portion of the core to show the impact of the perturbations on pin peaking factors.
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Lewis, Bryan R., Ronald A. Pawlowski, Kevin J. Greek et Andrew C. Klein. « Advanced thermionic reactor systems design code ». Dans Proceedings of the eighth symposium on space nuclear power systems. AIP, 1991. http://dx.doi.org/10.1063/1.40043.

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Rapports d'organisations sur le sujet "NUCLEAR REACTOR SYSTEMS"

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Meyer, L. C. Nuclear plant-aging research on reactor protection systems. Office of Scientific and Technical Information (OSTI), janvier 1988. http://dx.doi.org/10.2172/5122402.

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Shannon Bragg-Sitton, J. Michael Doster et Alan Rominger. Reactor Subsystem Simulation for Nuclear Hybrid Energy Systems. Office of Scientific and Technical Information (OSTI), septembre 2012. http://dx.doi.org/10.2172/1060985.

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Lawrence, J. D. Software reliability and safety in nuclear reactor protection systems. Office of Scientific and Technical Information (OSTI), novembre 1993. http://dx.doi.org/10.2172/10108329.

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Preckshot, G. G. Reviewing real-time performance of nuclear reactor safety systems. Office of Scientific and Technical Information (OSTI), août 1993. http://dx.doi.org/10.2172/10178197.

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Hawrylak, Peter, John Hale, Mauricio Papa, Donald Wall, Corey Hines, Thomas Edgar et Philip Craig. Cyber Security Analysis for Nuclear Reactor Control Systems. Final Technical Report. Office of Scientific and Technical Information (OSTI), décembre 2020. http://dx.doi.org/10.2172/1650024.

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Palomar, J., et R. Wyman. The Programmable Logic Controller and its application in nuclear reactor systems. Office of Scientific and Technical Information (OSTI), septembre 1993. http://dx.doi.org/10.2172/10185827.

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Brittain, C. R., P. J. Otaduy et R. B. Perez. Development of a general learning algorithm with applications in nuclear reactor systems. Office of Scientific and Technical Information (OSTI), décembre 1989. http://dx.doi.org/10.2172/5273390.

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Anderson, Mark, Kumar Sridharan, Dane Morgan, Per Peterson, Pattrick Calderoni, Randall Scheele, Andrew Casekka et Bruce McNamara. Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling. Office of Scientific and Technical Information (OSTI), janvier 2015. http://dx.doi.org/10.2172/1169921.

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Gallup, D. R. The scalability of OTR (out-of-core thermionic reactor) space nuclear power systems. Office of Scientific and Technical Information (OSTI), mars 1990. http://dx.doi.org/10.2172/6923465.

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Professor Neill Todreas. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design. Office of Scientific and Technical Information (OSTI), octobre 2001. http://dx.doi.org/10.2172/839057.

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