Bibliografías temáticas / Nuclear reactors – Computer simulation

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Literatura académica sobre el tema "Nuclear reactors – Computer simulation"

Autor: Grafiati
Publicado: 4 de junio de 2021
Última modificación: 7 de febrero de 2022

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Artículos de revistas sobre el tema "Nuclear reactors – Computer simulation"

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Wulff, Wolfgang. "Computer simulation of two-phase flow in nuclear reactors". Nuclear Engineering and Design 141, n.º 1-2 (junio de 1993): 303–13. http://dx.doi.org/10.1016/0029-5493(93)90108-l.

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Bakhshayesh, Moshkbar y Naser Vosoughi. "A simulation of a pebble bed reactor core by the MCNP-4C computer code". Nuclear Technology and Radiation Protection 24, n.º 3 (2009): 177–82. http://dx.doi.org/10.2298/ntrp0903177b.

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Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results), chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.
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3

Salcedo, L. L., E. Oset, M. J. Vicente-Vacas y C. Garcia-Recio. "Computer simulation of inclusive pion nuclear reactions". Nuclear Physics A 484, n.º 3-4 (julio de 1988): 557–92. http://dx.doi.org/10.1016/0375-9474(88)90310-7.

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Lee, Kim, Moon, Lim y Cho. "Heat-Absorbing Capacity of High-Heat-Flux Components in Nuclear Fusion Reactors". Energies 12, n.º 19 (3 de octubre de 2019): 3771. http://dx.doi.org/10.3390/en12193771.

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Nuclear fusion energy is a solution to the substitution of fossil fuels and the global energy deficit. However, among the several problems encountered for realizing a nuclear fusion reactor, the divertor presents difficulties due to the tremendous heat flux (~10 MW/m2) from high-temperature plasma. Also, neutrons produce additional heat (~17.5 MW/m3) from collisions with the materials’ atoms. This may lead to unexpected effects such as thermal failure. Thus, a comprehensive investigation on the divertor module is needed to determine the heat-absorbing capacity of the divertor module so to maintain the effect of incident heat flux. In this study, using an analytical approach and a simulation, the quantitative effect of heat generation on the thermophysical behavior, such as temperature and thermal stress, was analyzed while maintaining the incident heat flux. Then, a correlated equation was derived from the thermal design criteria, namely, the maximum thimble temperature and the safety factor at the vulnerable point. Finally, on the basis of the thermal design criteria, the heat-absorbing capacity of a nuclear fusion reactor in operating conditions was determined. This study contributes to the understanding of the divertor’s effects in nuclear fusion reactors for high-heat-flux and high-temperature applications.
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5

Okunev, V. S. "Fundamentally New Composite Materials of Fast Reactors Made on the Basis of Nanotechnology". Key Engineering Materials 887 (mayo de 2021): 159–64. http://dx.doi.org/10.4028/www.scientific.net/kem.887.159.

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The main goal of the work is to identify the advantages of fast reactors when using nanotechnology in the manufacture of core materials. The research methods are based on the adaptation of known technologies (including powder metallurgy) to the design of fast reactors and on the numerical simulation of physical processes carried out using computer programs for the analysis of emergency conditions of fast reactors (including anticipated transient without scram - ATWS). The results of the research show that the use of structural materials based on steels hardened by nanooxides in combination with fundamentally new types of fuel based on composite materials can significantly improve the safety of nuclear technics. Sintered mixtures of ceramic microgranules (oxide, nitride) and nanoadditives of metallic beryllium or uranium are considered as nuclear fuel. Such composite nuclear fuel improves reactor safety and power. The following types of composite fuel were analyzed: mixed oxide with additives of a beryllium or uranium nanopowder, mixed mononitride with additives of a beryllium or uranium nanopowder. Most preferably, a ceramic-metal pellet fuel based on mononitride microgranules and uranium metal nanopowder. The use of such fuel (with a volume fraction of metallic uranium up to 20%) significantly increases the safety of the reactor, combining the advantages of metal and ceramics and completely neutralizing their disadvantages. The proposed materials are of practical importance in the development of new concepts of nuclear technics, in the transition to large-scale nuclear power and high-power reactors. The use of a new cermet-based composite fuel increases the power of the reactor and significantly increases the safety of the reactor.
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Sadek, I. S. y R. Vedantham. "Optimal control of distributed nuclear reactors with pointwise controllers". Mathematical and Computer Modelling 32, n.º 3-4 (agosto de 2000): 341–48. http://dx.doi.org/10.1016/s0895-7177(00)00139-4.

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Khorshidi, Abdollah. "Accelerator-Based Methods in Radio-Material 99Mo/99mTc Production Alternatives by Monte Carlo Method: The Scientific-Expedient Considerations in Nuclear Medicine". Journal of Multiscale Modelling 11, n.º 01 (14 de enero de 2019): 1930001. http://dx.doi.org/10.1142/s1756973719300016.

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Some accelerator technologies are already used for commercial [Formula: see text]Mo-99mTc production, as the economic criteria are considered representative of the main differences between diverse technologies including accelerators and reactors. This study has provided a review of known and potential [Formula: see text]Mo production using conventional medical facilities. Accelerator-based method in 99mTc production via ([Formula: see text], [Formula: see text]) direct reaction on [Formula: see text]Mo was simulated using 18[Formula: see text]MeV proton beam. Meanwhile, a conceptual design for indirect [Formula: see text]Mo production via [Formula: see text]Mo([Formula: see text])[Formula: see text]Mo and [Formula: see text]Mo(n,[Formula: see text]2n)[Formula: see text]Mo reactions was investigated when an electron source of 35[Formula: see text]MeV by accelerator is used. These indirect reactions were explored via inserted [Formula: see text]Mo samples at different positions inside the lead region. Furthermore, Adiabatic Resonance Crossing (ARC) method based on proton accelerator via transmutation in [Formula: see text]Mo([Formula: see text]Mo was examined when the 30[Formula: see text]MeV proton beam is used. Saturation activity and yield were investigated using alternative proposed methods. The potential proliferation risk associated with accelerator technetium production is minimal. While accelerators could be turned into neutron sources which could in turn be used to irradiate [Formula: see text]U to breed plutonium, and centrifuges used to enrich [Formula: see text]Mo for targets could conceivably be turned to enriching uranium, this would result in very tiny global production capability particularly compared with research or power reactors. The potential of the fresh methods could provide a replacement or complement over current reactor-based supply sources in various radioisotopes production purposes.
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Gabbar, Hossam A., Muhammad R. Abdussami y Md Ibrahim Adham. "Micro Nuclear Reactors: Potential Replacements for Diesel Gensets within Micro Energy Grids". Energies 13, n.º 19 (5 de octubre de 2020): 5172. http://dx.doi.org/10.3390/en13195172.

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Resilient operation of medium/large scale off-grid energy systems, which is a key challenge for energy crisis solutions, requires continuous and sustainable energy resources. Conventionally, micro energy grids (MEGs) are adopted to supply electricity and thermal energy simultaneously. Fossil-fired gensets, such as diesel generators, are indispensable components for off-grid MEGs due to the intermittent nature of renewable energy sources (RESs). However, fossil-fired gensets emit a significant amount of greenhouse gases (GHGs). Therefore, this study investigates an alternative source as an economical and environmental replacement for diesel gensets that can reduce GHG emissions and ensure system reliability. A MEG is developed in this paper to support a considerably large-scale electric and thermal demand at Ontario Tech University (UOIT). Different sizes of diesel gensets and RESs, such as solar, wind, hydro, and biomass, are combined in the MEG for off-grid applications. To evaluate diesel gensets’ competency, the diesel genset is substituted by an emission-free generation source named microreactor (MR). The fossil-fired MEG and MR-based MEG are optimized by an intelligent optimization technique, namely particle swarm optimization (PSO). The objective of the PSO is to minimize the net present cost (NPC). The simulation results show that MR-based MEG could be an excellent replacement for a diesel genset in terms of NPC and selected key performance indicators (KPIs). A comprehensive sensitivity analysis is also carried out to validate the simulation results.
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Prošek, Andrej y Marko Matkovič. "RELAP5/MOD3.3 Analysis of the Loss of External Power Event with Safety Injection Actuation". Science and Technology of Nuclear Installations 2018 (2018): 1–14. http://dx.doi.org/10.1155/2018/6964946.

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The code assessment typically comprises basic tests cases, separate effects test, and integral effects tests. On the other hand, the thermal hydraulic system codes like RELAP5/MOD3.3 are primarily intended for simulation of transients and accidents in light water reactors. The plant measured data come mostly from startup tests and operational events. Also, for operational events the measured plant data may not be sufficient to explain all details of the event. The purpose of this study was therefore besides code assessment to demonstrate that simulations can be very beneficial for deep understanding of the plant response and further corrective measures. The abnormal event with reactor trip and safety injection signal actuation was simulated with the latest RELAP5/MOD3.3 Patch 05 best-estimate thermal hydraulic computer code. The measured and simulated data agree well considering the major plant system responses and operator actions. This suggests that the RELAP5 code simulation is good representative of the plant response and can complement not available information from plant measured data. In such a way, an event can be better understood.
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Pakari, O., V. Lamirand, B. Vandereydt, F. Vitullo, M. Hursin, C. Kong y A. Pautz. "Design and Simulation of Gamma Spectrometry Experiments in the CROCUS Reactor". EPJ Web of Conferences 225 (2020): 04016. http://dx.doi.org/10.1051/epjconf/202022504016.

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Gamma rays in nuclear reactors, arising either from fission or decay processes, significantly contribute to the heating and dose of the reactor components. Zero power research reactors offer the possibility to measure gamma rays in a purely neutronic environment, allowing for validation experiments of computed spectra, dose estimates, reactor noise and prompt to delayed gamma ratios. This data then contributes to models, code validation and photo atomic nuclear data evaluation. In order to contribute to aforementioned experimental data, gamma detection capabilities are being added to the CROCUS reactor facility. The CROCUS reactor is a two-zone, uranium-fueled light water moderated facility operated by the Laboratory for Reactor Physics and Systems Behaviour (LRS) at the Swiss Federal Institute of Technology Lausanne (EPFL). With a maximum power of 100W, it is a zero power reactor used for teaching and research, most recently for intrinsic and induced neutron noise studies. For future gamma detection applications in the CROCUS reactor, an array of four detectors - two large 5”x10” Bismuth Germanate (BGO) and two smaller Cerium Bromide (CeBr3) scintillators - was acquired. The BGO detectors are to be arbitrarily positioned in the core reflector and out of the vessel for measurements at arbitrary distances. The CeBr3 detectors on the other hand are small enough to be set in the guide tubes of the control rods for in-core measurements. We present a study of the neutron and gamma flux in the core and reflector using the MCNP 6.2 and Serpent 2 Monte Carlo codes for coupled neutron and photon transport criticality calculations. More specifically, we investigate and compare predicted spectra as well as reactivity worth of different envisioned experimental setups. We further predict pulse height spectra as well as doses to the crystals with and without cadmium shielding to estimate allowable reactor powers with respect to detector radiation hardness. The results serve as basis for calibration and aid in the design and regulatory approval of the experiments.
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Tesis sobre el tema "Nuclear reactors – Computer simulation"

1

OLIVEIRA, JOSE R. de. "Programa computacional para estudo da estrategia de controle de um reator nuclear do tipo PWR". reponame:Repositório Institucional do IPEN, 2002. http://repositorio.ipen.br:8080/xmlui/handle/123456789/11060.

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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2

Huntington, James E. "Computer simulation studies of nuclear reactor fuel and related uranium phases". Thesis, Keele University, 1994. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.384936.

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Biaty, Patricia Andrea Paladino. "Pré-processador matemático para o código Relap5 utilizando o Microsoft Excel". Universidade de São Paulo, 2006. http://www.teses.usp.br/teses/disponiveis/85/85133/tde-14052007-141446/.

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O estudo termo-hidráulico, utilizado para análise de acidentes e transientes em reatores nucleares, é feito com o uso de algumas ferramentas computacionais sofisticadas. Esses programas utilizam uma filosofia realista (best estimate) para análise de acidentes e transientes em reatores refrigerados à água leve do tipo PWR (Pressurized Water Reactor) e sistemas associados. O código RELAP5, objeto de nosso estudo, tem sido usado como uma ferramenta para o licenciamento de instalações nucleares no nosso país. Uma das maiores dificuldades na simulação de acidentes e transientes em uma instalação nuclear com o código RELAP5 é a quantidade de informações necessárias, que na maioria dos casos é muito grande. Além disso, existe a necessidade de uma quantidade razoável de operações matemáticas para os cálculos da geometria dos componentes. Portanto, a fim de facilitar a manipulação destas informações, percebeu-se a necessidade do desenvolvimento de um pré-processador amigável com o usuário, para realização desses cálculos e para elaboração dos dados de entrada do RELAP5. A ferramenta escolhida foi o MS-EXCEL, que apresentou grande potencialidade no desenvolvimento do pré-processador desejado.
Computational program are used for thermal hydraulic analysis of accidents and transients conditions in nuclear power plants. The RELAP5 code has been developed to simulate accidents and transients conditions, performing a best estimate analysis, in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code, which has been used as a tool for licensing nuclear facilities in Brazil, is the objective of the study performed in this work. The main problem in using the RELAP5 code is the huge amount of information necessary to model the nuclear reactor and thus to simulate thermal-hydraulic accidents. Moreover, the RELAP5 code input data requires a large amount of mathematical operations to calculate the geometry of the plant components. Therefore, in order to make easier the data input for the RELAP5 code a friendly preprocessor has been developed. The preprocessor accepts basic information about the geometry of the plant components and performs all the calculations needed for the RELAP5 input. This preprocessor has been developed based on the MS-EXCEL software.
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4

Dalle, Hugo Moura. "Simulação do reator TRIGA IPR-R1 utilizando metodos de transporte por Monte Carlo". [s.n.], 2005. http://repositorio.unicamp.br/jspui/handle/REPOSIP/267210.

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Orientador: Elias Basile Tambourgi
Tese (doutorado) - Universidade Estadual de Campinas, Faculdade de Engenharia Quimica
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Resumo: A utilização do método Monte Carlo na simulação do transporte de partículas em reatores nucleares é crescente e constitui uma tendência mundial. O maior inconveniente dessa técnica, a grande exigência de capacidade de processamento, vem sendo superado pelo contínuo desenvolvimento de processadores cada vez mais rápidos. Esse contexto permitiu o desenvolvimento de metodologias de cálculo neutrônico de reatores nas quais se acopla a parte do transporte de partículas, feita com um código de Monte Carlo, ao cálculo de queima e decaimento radioativo. Neste trabalho tal metodologia de simulação é implantada, validada para reatores de pesquisas, notadamente os do tipo TRIGA e finalmente utilizada na simulação neutrônica do reator TRIGA IPR ¿ RI do CDTN/CNEN. O sistema de códigos empregados é constituído pelos amplamente utilizados códigos MCNP4B (transporte por método Monte Carlo) e ORIGEN2.1 (queima e decaimento radioativo). Apesar dos esforços recentes no sentido de agrupar as duas etapas de cálculo, transporte e queima, em um único código, até o momento esta opção não está disponível e, portanto, um terceiro código é utilizado para realizar o acoplamento transporte/queima. Neste trabalho utilizou-se para tal o código MONTEBURNS. O sistema formado por estes três códigos permitiu obter os parâmetros neutrônicos de interesse do IPR ¿ R1 através apenas de simulação teórica, sem a necessidade de qualquer tipo de ajuste baseado em dados experimentais, em boa concordância com os valores medidos... Observação: O resumo, na íntegra, poderá ser visualizado no texto completo da tese digital
Abstract: The use of Monte Carlo methods in particles transport simulations of nuclear reactor is growing fast and constitutes a strong tendency all over the world. The major inconvenient of such techniques is the huge demand of processing power which has been surpassed the development of reactor physics calculation methodologies in which the particles transport part, made by a Monte Carlo transport code, is linked with the burnup and radioactive decay part of the simulation. On this work a such simulation methodology is made operational, validated for research reactors, mainly for TRIGA reactor and finally utilized for reactor physics simulation of the CDTN¿s TRIGA IPR ¿ R1. The adopted codes system is constituted by the widespreadly used codes MCNP4B (Monte Carlo transport) and ORIGEN2.1 (burnup and radioactive decay). In spite of the very recent efforts toward get together both, transport and burnup, in only one code at the moment this is a not available option and therefore, a third code is needed to carry out the linkage transport/burnup. MONTEBURS code was used to this purpose. This three codes system has allowed to obtain the physical parameters of IPR ¿ R1 calculated using only theoretical simulation without any kind of experimental adjustment or interaction between experiments and calculation in good agreement with measured values... Note: The complete abstract is available with the full electronic digital thesis or dissertations
Doutorado
Sistemas de Processos Quimicos e Informatica
Doutor em Engenharia Química
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5

Bollen, Rob. "Verification and validation of computer simulations with the purpose of licensing a pebble bed modular reactor". Thesis, Stellenbosch : Stellenbosch University, 2002. http://hdl.handle.net/10019.1/53216.

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Thesis (MBA)--Stellenbosch University, 2002.
ENGLISH ABSTRACT: The Pebble Bed Modular Reactor is a new and inherently safe concept for a nuclear power generation plant. In order to obtain the necessary licenses to build and operate this reactor, numerous design and safety analyses need to be performed. The results of these analyses must be supported with substantial proof to provide the nuclear authorities with a sufficient level of confidence in these results to be able to supply the required licences. Beside the obvious need for a sufficient level of confidence in the safety analyses, the analyses concerned with investment protection also need to be reliable from the investors’ point of view. The process to be followed to provide confidence in these analyses is the verification and validation process. It is aimed at presenting reliable material against which to compare the results from the simulations. This material for comparison will consist of a combination of results from experimental data, extracts from actual plant data, analytical solutions and independently developed solutions for the simulation of the event to be analysed. Besides comparison with these alternative sources of information, confidence in the results will also be built by providing validated statements on the accuracy of the results and the boundary conditions with which the simulations need to comply. Numerous standards exist that address the verification and validation of computer software, for instance by organisations such as the American Society of Mechanical Engineers (ASME) and the Institute of Electrical and Electronics Engineers (IEEE). The focal points of the verification and validation of the design and safety analyses performed on typical PBMR modes and states, and the requirements imposed by both the local and overseas nuclear regulators, are not entirely enveloped by these standards. For this reason, PBMR developed a systematic and disciplined approach for the preparation of the Verification and Validation Plan, aimed at capturing the essence of the analyses. This approach aims to make a definite division between software development and the development of technical analyses, while still using similar processes for the verification and validation. The reasoning behind this is that technical analyses are performed by engineers and scientists who should only be responsible for the verification and validation of the models and data they use, but not for the software they are dependent on. Software engineers should be concerned with the delivery of qualified software to be used in the technical analyses. The PBMR verification and validation process is applicable to both hand calculations and computer-aided analyses, addressing specific requirements in clearly defined stages of the software and Technical Analysis life cycle. The verification and validation effort of the Technical Analysis activity is divided into the verification and validation of models and data, the review of calculational tasks, and the verification and validation of software, with the applicable information to be validated, captured in registers or databases. The resulting processes are as simple as possible, concise and practical. Effective use of resources is ensured and internationally accepted standards have been incorporated, aiding in faith in the process by all stakeholders, including investors, nuclear regulators and the public.
AFRIKAASE OPSOMMING: Die Modulêre Korrelbedreaktor is ’n nuwe konsep vir ’n kernkragsentrale wat inherent veilig is. Dit word deur PBMR (Edms.) Bpk. ontwikkel. Om die nodige vergunnings om so ’n reaktor te kan bou en bedryf, te bekom, moet ’n aansienlike hoeveelheid ontwerp- en veiligheidsondersoeke gedoen word. Die resultate wat hierdie ondersoeke oplewer, moet deur onweerlegbare bewyse ondersteun word om vir die owerhede ’n voldoende vlak van vertroue in die resultate te gee, sodat hulle die nodigde vergunnings kan maak. Benewens die ooglopende noodsaak om ’n voldoende vlak van vertroue in die resultate van die veiligheidsondersoeke te hê, moet die ondersoeke wat met die beskerming van die beleggers se beleggings gepaard gaan, net so betroubaar wees. Die proses wat gevolg word om vertroue in die resultate van die ondersoeke op te bou, is die proses van verifikasie en validasie. Dié proses is daarop gerig om betroubare vergelykingsmateriaal vir simulasies voor te lê. Hierdie vergelykingsmateriaal vir die gebeurtenis wat ondersoek word, sal bestaan uit enige kombinasie van inligting wat in toetsopstellings bekom is, wat in bestaande installasies gemeet is, wat analities bereken is; asook dit wat deur ’n derde party onafhanklik van die oorspronklike ontwikkelaars bekom is. Vertroue in die resultate van die ondersoeke sal, behalwe deur vergelyking met hierdie alternatiewe bronne van inligting, ook opgebou word deur die resultate te voorsien van ’n gevalideerde verklaring wat die akkuraatheid van die resultate aantoon en wat die grensvoorwaardes waaraan die simulasies ook moet voldoen, opsom. Daar bestaan ’n aansienlike hoeveelheid internasionaal aanvaarde standaarde wat die verifikasie en validasie van rekenaarsagteware aanspreek. Die standaarde kom van instansies soos die Amerikaanse Vereniging vir Meganiese Ingenieurs (ASME) en die Instituut vir Elektriese en Elektroniese Ingenieurs (IEEE) – ook van Amerika. Die aandag wat deur die Suid-Afrikaanse en oorsese kernkragreguleerders vereis word vir die toestande wat spesifiek geld vir korrelbedreaktors, word egter nie geheel en al deur daardie standaarde aangespreek nie. Daarom het die PBMR maatskappy ’n stelselmatige benadering ontwikkel om verifikasie- en validasieplanne voor te berei wat die essensie van die ondersoeke kan ondervang. Hierdie benadering is daarop gemik om ’n duidelike onderskeid te maak tussen die ontwikkeling van sagteware en die ontwikkeling van tegniese ondersoeke, terwyl steeds gelyksoortige prosesse in die verifikasie en validasie gebruik sal word. Die rede hiervoor is dat tegniese ondersoeke uitgevoer word deur ingenieurs en wetenskaplikes wat net vir verifikasie en validasie van hulle eie modelle en die gegewens verantwoordelik gehou kan word, maar nie vir die verifikasie en validasie van die sagteware wat hulle gebruik nie. Ingenieurs wat spesialiseer in sagteware-ontwikkeling behoort verantwoordelik te wees vir die daarstelling van sagteware wat deur die reguleerders gekwalifiseer kan word, sodat dit in tegniese ondersoeke op veiligheidsgebied gebruik kan word. Die verifikasie- en validasieproses van die PBMR is sowel vir handberekeninge as vir rekenaarondersteunde-ondersoek geskik. Hierdie proses spreek spesifieke vereistes in onderskeie stadiums gedurende die lewenssiklusse van die ontwikkeling van sagteware en van tegniese ondersoeke aan. Die verifikasie- en validasiewerk vir tegniese ondersoeksaktiwiteite is verdeel in die verifikasie en validasie van modelle en gegewens, die nasien van berekeninge en die verifikasie en validasie van sagteware, waarby die betrokke inligting wat gevalideer moet word, versamel word in registers of databasisse. Die prosesse wat hieruit voortgevloei het, is so eenvoudig as moontlik, beknop en prakties gehou. Hierdeur is ’n effektiewe benutting van bronne verseker. Internasionaal aanvaarde standaarde is gebruik wat die vertroue in die proses deur alle betrokkenes, insluitende beleggers, die owerhede en die publiek, sal bevorder.
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6

Revel, Aldric. "Nuclear forces at the extremes". Thesis, Normandie, 2018. http://www.theses.fr/2018NORMC227/document.

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L’émission de paires de neutrons par les noyaux riches en neutrons 18C et 20O (isotones N = 12) est étudié par réactions de knock-out d’un nucléon des faisceaux secondaires 19N et 21O, peuplant ainsi des états non liés jusqu’à 15 MeV au-dessus de leur seuil d’émission deux neutrons. L’analyse des corrélations des triples coïncidences fragment+n+n montre que la décroissance 19N(−1p)→18C → 16C+n+n est clairement dominée par l’émission directe de paires. Les corrélations n-n, les plus grandes jamais observées, suggèrent la prédominance d’un coeur de 14C entouré de quatre neutrons arrangés en paires très corrélées. De plus, une importante compétition du mode de décroissance séquentiel est observée dans la décroissance 21O(−1n) → 20O → 18O+n+n, interprétée par la déformation causée par le knock-out d’un neutron très lié ayant pour effet de casser le cœur de 16O et ainsi de réduire le nombre de paires.De plus, les états non liés du 26F et 28F sont étudiés. Les deux systèmes étant peuplés par knock-out d’un nucléon du 27F dans le cas du 26F et du 29Ne ou du 29F pour 28F. Cinq états ont été observés pour 26F avec en particulier l’état de plus basse énergie (0.39 MeV) identifié comme l’état 3+ résultant du couplage d5/2 ⊗ d3/2 . Pour 28F, cinq états ont aussi été observés et l’état fondamental (200 keV) a été identifié comme étant de parité négative, plaçant ainsi 28F dans l’îlot d’inversion
The emission of neutron pairs from the neutron-rich N = 12 isotones 18C and 20O has been studied by high-energy nucleon knockout from 19N and 21O secondary beams, populating unbound states of the two isotones up to 15 MeV above their two-neutron emission thresholds. The analysis of triple fragment-n-n correlations shows that the decay 19N(−1p) → 18C → 16C+n+n is clearly dominated by direct pair emission. The two-neutron correlation strength, the largest ever observed, suggests the predominance of a 14C core surrounded by four neutrons arranged in strongly correlated pairs. On the other hand, a significant competition of a sequential branch is found in the decay 21O(−1n) → 20O → 18O+n+n, attributed to its formation through the knockout of a deeply-bound neutron that breaks the 16O core and reduces the number of pairs.Moreover, unbound states in 26F and 28F have been studied. The two systems were probed using single-nucleon knockout reaction from secondary beams of 27F respectively in the case of 26F, and 29Ne and 29F for 28F. Five possible states have been identified in 26F, with in particular the lowest energy one (0.39 MeV) being identified as the 3+ state resulting from the d5/2 ⊗ d3/2 coupling. In the case of 28F, five unbound state have also been observed and in particular its ground state (200 keV) has been identified as a negative parity state, meaning that 28F is located inside the island of inversion
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Tshamala, Mubenga Carl. "Simulation and control implications of a high-temperature modular reactor (HTMR) cogeneration plant". Thesis, Stellenbosch : Stellenbosch University, 2014. http://hdl.handle.net/10019.1/86264.

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Thesis (MScEng)--Stellenbosch University, 2014.
ENGLISH ABSTRACT: Traditionally nuclear reactor power plants have been optimised for electrical power generation only. In the light of the ever-rising cost of dwindling fossil fuel resources as well the global polluting effects and consequences of their usage, the use of nuclear energy for process heating is becoming increasingly attractive. In this study the use of a so-called cogeneration plant in which a nuclear reactor energy source is optimised for the simultaneous production of superheated steam for electrical power generation and process heat is considered and analysed. The process heat superheated steam is generated in a once-through steam generator of heat pipe heat exchanger with intermediate fluid while steam for power generation is generated separately in a once-through helical coil steam generator. A 750 °C, 7 MPa helium cooled HTMR has been conceptually designed to simultaneously provide steam at 540 °C, 13.5 MPa for the power unit and steam at 430 °C, 4 MPa for a coal-to-liquid fuel process. The simulation and dynamic control of such a typical cogeneration plant is considered. In particular, a theoretical model of a typical plant will be simulated with the aim of predicting the transient and dynamic behaviour of the HTMR in order to provide guideline for the control of the plant under various operating conditions. It was found that the simulation model captured the behaviour of the plant reasonably well and it is recommended that it could be used in the detailed design of plant control strategies. It was also found that using a 1500 MW-thermal HTMR the South African contribution to global pollution can be reduced by 1.58%.
AFRIKAANSE OPSOMMING: Tradisioneel is kernkragaanlegte vir slegs elektriese kragopwekking geoptimeer. In die lig van die immer stygende koste van uitputbare fossielbrandstohulpbronne asook die besoedelingsimpak daarvan wêreldwyd, word die gebruik van kernkrag vir prosesverhitting al hoe meer aanlokliker. In hierdie studie word die gebruik van ‘n sogenaamde mede-opwekkingsaanleg waarin ‘n kernkragreaktor-energiebron vir die gelyktydige produksie van oorverhitte stoom vir elektriese kragopwekking en proseshitte oorweeg ontleed word. Die oorvehitte stoom word in ‘n enkeldeurvloei-stoomopwekking van die hittepyp-hitteruiler met tussenvloeistof opgewek en stoom vir kragopwekking word apart in ‘n enkeldeurvloei-spiraalspoel-stoomopwekker opgewek. ‘n 750 °C, 7 MPa heliumverkoelde HTMR is konseptueel ontwerp vir die gelytydige veskaffing van stoom by 540 °C, 13.5 MPa, vir die kragopwekkings eenheid, en stoom by 430 °C, 4 MPa, vir ‘n steenkool-tot-vloeibare (CTL) brandstoff proses. Die simulasie en dinamiese beheer van ‘n tipiese HTMR mede-opwekkingsaanleg word beskou. ‘n die besonder word ‘n teoretiese model van die transiënte en dinamiese gedrag van die aanleg gesimuleer om sodoene riglyne te identifiseer vir die ontwikkeling van dinamiese beheer strategië vir verskillende werkstoestande van die aanleg. Daar was ook gevind dat die simulasie model van die aanleg se gedrag goed nageboots word en dat dit dus gebruik kan word vir beheer strategie doeleindes. Indien so ‘n 1500 MW-termies HTMR gebruik word sal dit die Suid Afrikaanse besoedling met 1.58% sal kan verminder.
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SILVESTRE, LARISSA J. B. "PCRELAP5 - Programa de cálculo para os dados de entrada do código RELAP5". reponame:Repositório Institucional do IPEN, 2016. http://repositorio.ipen.br:8080/xmlui/handle/123456789/26393.

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IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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PERRENOUD, HELENA G. "Modulo de extracao de eventos em assinaturas de potencia de valvulas moto-operadas, usando um sistema especialista para o sistema de diagnostico de MOV's utilizado em reatores nucleares". reponame:Repositório Institucional do IPEN, 2001. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10967.

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IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Kruckenberg, Norman E. "A piping network model program for small computer". Ohio : Ohio University, 1986. http://www.ohiolink.edu/etd/view.cgi?ohiou1183138191.

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Libros sobre el tema "Nuclear reactors – Computer simulation"

1

Puska, Eija Karita. Nuclear reactor core modelling in multifunctional simulators. Espoo [Finland]: Technical Research Centre of Finland, 1999.

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Klein, M. E. Simulation of in-reactor experiments with the ELOCA.Mk5 code. Chalk River, Ont: Chalk River Laboratories, 1994.

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Kim, K. Assessment of RELAP5/MOD2 critical flow model using Marviken test data 15 and 24. Washington, DC: Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1992.

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Kim, Kyu-Soo. Assessment of RELAP5/MOD2 critical flow model using Marviken test data 15 and 24. Washington, DC: Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1992.

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Mayya, Y. S. Containment aerosol behaviour simulation studies in the BARC nuclear aerosol test facility. Mumbai: Bhabha Atomic Research Centre, 2005.

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Hämäläinen, A. Applying thermal hydraulics modeling in coupled processes of nuclear power plants. [Espoo, Finland]: VTT Technical Research Centre of Finland, 2005.

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Ross, Kyle. MELCOR best practices as applied in the State-of-the-Art Reactor Consequence Analyses (SOARCA) Project. Washington, DC: U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, 2014.

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Llopis, C. Assessment of RELAP5/MOD3.2-NPA3.4 against an inadvertent closure of all three MSIV's in VANDELLOS-II nuclear power plant. Washington, DC: Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 2001.

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Weingardt, Jay J. TAC2D studies of Mark I containment drywell shell melt-through. Washington, DC: Division of Reactor Accident Analysis, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1988.

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Analytis, G. Th. Assessment of interfacial shear and wall heat transfer of RELAP5/MOD2/36.02 during reflooding. Washington, DC: Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1989.

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Capítulos de libros sobre el tema "Nuclear reactors – Computer simulation"

1

Fauquet-Alekhine, Philippe y Carole Maridonneau. "Piloting Nuclear Reactors". En Simulation Training: Fundamentals and Applications, 59–85. Cham: Springer International Publishing, 2015. http://dx.doi.org/10.1007/978-3-319-19914-6_3.

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Lopez-Munguia, Agustin. "Simulation of Batch Enzyme Reactors". En Computer and Information Science Applications in Bioprocess Engineering, 179–89. Dordrecht: Springer Netherlands, 1996. http://dx.doi.org/10.1007/978-94-009-0177-3_15.

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Kohler, W. y M. Schindler. "Two Phase Flow Analysis Capability of Advanced Computer Codes". En Nuclear Simulation, 133–41. Berlin, Heidelberg: Springer Berlin Heidelberg, 1990. http://dx.doi.org/10.1007/978-3-642-84279-5_10.

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Graf, U. y W. F. Werner. "Experience with Simulation of Nuclear Systems on Parallel Processing Computer Systems". En Nuclear Simulation, 3–14. Berlin, Heidelberg: Springer Berlin Heidelberg, 1990. http://dx.doi.org/10.1007/978-3-642-84279-5_1.

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Müller, R., R. Böer y H. Finnemann. "Nuclear Core and Power Plant Simulation on High Performance Parallel Computer Systems". En Nuclear Simulation, 104–15. Berlin, Heidelberg: Springer Berlin Heidelberg, 1990. http://dx.doi.org/10.1007/978-3-642-84279-5_8.

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Bakker, R., E. Große-Dunker y P. Leishman. "Hydraulic Network Modelling for Real-Time Power Plant Simulation with Computer Aided Code Generation". En Nuclear Simulation, 25–37. Berlin, Heidelberg: Springer Berlin Heidelberg, 1990. http://dx.doi.org/10.1007/978-3-642-84279-5_3.

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Wallner, M. "Computer Simulation of the Long-Term Stability of a NuclearWaste Repository in a Salt Dome". En Nuclear Simulation, 326–43. Berlin, Heidelberg: Springer Berlin Heidelberg, 1987. http://dx.doi.org/10.1007/978-3-642-83221-5_21.

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Hughes, G. y R. S. Overton. "The Monte Carlo Simulation of Thermal Noise in Fast Reactors". En Noise and Nonlinear Phenomena in Nuclear Systems, 313–38. Boston, MA: Springer US, 1989. http://dx.doi.org/10.1007/978-1-4684-5613-4_26.

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Jurs, Peter C. y Debra S. Egolf. "Carbon-13 Nuclear Magnetic Resonance Spectrum Simulation". En Computer-Enhanced Analytical Spectroscopy, 163–82. Boston, MA: Springer US, 1987. http://dx.doi.org/10.1007/978-1-4684-5368-3_8.

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Rodríguez-Hernandez, Andrés, Armando M. Gómez-Torres, Edmundo del Valle-Gallegos, Javier Jimenez-Escalante, Nico Trost y Victor H. Sanchez-Espinoza. "Accelerating AZKIND Simulations of Light Water Nuclear Reactor Cores Using PARALUTION on GPU". En Communications in Computer and Information Science, 419–31. Cham: Springer International Publishing, 2016. http://dx.doi.org/10.1007/978-3-319-32243-8_29.

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Actas de conferencias sobre el tema "Nuclear reactors – Computer simulation"

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Ablay, Gunyaz. "Modeling and simulation of advanced nuclear reactors". En 2013 International Conference on Electronics, Computer and Computation (ICECCO). IEEE, 2013. http://dx.doi.org/10.1109/icecco.2013.6718250.

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Chegrani, Yacine, Corinne d’Aletto, Jacques Di Salvo y Evgeny Ivanov. "Validation of SIMMER-III Neutronics Module for the Simulation of Reactivity Injection Accident in Material Testing Reactors". En 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29188.

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The “Institut de Radioprotection et de Suˆrete´ Nucle´aire”, as the technical support of the French Safety Authority, carries out studies and research to analyze and assess the safety of all nuclear plants. In this frame IRSN studies the feasibility of modeling Material Testing Reactor core with SIMMER-III code, for simulation of reactivity initiated accidental transients. The SIMMER-III multi-physics code system was initially developed for mechanistic safety analyses of liquid metal cooled fast reactors while employing coupled spatial neutron kinetics and thermal hydraulics models. Neutronics and thermal-hydraulics SIMMER-III models have been extended to safety analyses for water cooled and moderated reactors. The use of a code like SIMMER-III requires approximations; it computes a simplified R-Z geometry and chemistry description of the core that must be validated. The methods applied consist here in developing models of the same reactor on several scales of detail. The first step is the validation of the cross section condensation for deterministic APOLLO2 calculation against Monte Carlo TRIPOLI4 2D model. Temperature effects, kinetic parameters and void coefficients on the whole core are then calculated on a 2D APOLLO2 model, using the Method of Characteristics. These parameters are also computed with a 3D combined transport and diffusion calculations by means of APOLLO2/CRONOS2 calculations, validated against a TRIPOLI4 3D precise reference model. The final step is the validation of the simmer-like R-Z geometry in APOLLO2 Sn and Pij. Finally, an R-Z geometry has been computed in SIMMER-III, for the calculation of the kinetic parameters and temperature coefficients. This validation method has been applied to Jules Horowitz Reactor, a French Material Testing Reactor currently in commissioning by the CEA. This leads to conclude that discrepancies due to simplifications are acceptable. Moreover SIMMER-III shows quite a good agreement with CEA ring calculation on the kinetic parameters. Concerning neutronics feedbacks coefficients, further analyses remain necessary.
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Peter, Geoffrey J. "Numerical Simulation of Accident Scenario in HTGR (Pebble Bed Reactor) Using COMSOL® Code". En 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-16535.

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High Temperature Gas Cooled Reactor (HTGR) development and operation is expanding in the United Kingdom, Russia, USA (Generation IV Reactors), and France (Pebble Bed Modular Reactor, PBMR). A prototype pebble bed reactor producing 10 MW thermal, High Temperature Reactor (HTR-10) is in operation in China. However, the general public remains skeptical of the safety and the perceived dangers of possible accidents. Of particular concern are blockages caused by local variations in flow and heat transfer that lead to hot spots within the bed. This paper models the accident scenario resulting from blockages due to the retention of dust in the coolant gas or from the rupture of one or more fuel particles used in the High Temperature Gas Cooled (Pebble Bed) Nuclear Reactors using the commercially available computer code COMSOL. Numerical modeling of flow and heat transfer in a packed bed produces an Elliptical Non-Linear Partial Differential equation that requires custom made computer codes. Previously published results obtained from the use of a custom-made verified computer code are limited to one accident scenario and involve considerable modification to study different accident scenarios. Thus the use of a commercially available computer code that can simulate many different accident scenarios is of considerable advantage. Further, this paper compares numerical solutions obtained from custom-made computer code with COMSOL simulation and discusses the advantages and limitations of both codes.
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Ohshima, Hiroyuki y Masahiko Ohtaka. "Development of Computer Program for Whole Core Thermal-Hydraulic Analysis of Fast Reactors". En 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22034.

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A whole core thermal-hydraulic analysis program ACT was developed for the purpose of evaluating detailed in-core thermal-hydraulic phenomena of sodium cooled fast reactors under various reactor operation conditions. ACT consists of four kinds of calculation modules, i.e., fuel-assembly, inter-wrapper gap (core barrel), upper plenum and heat transport system modules. The latter two modules give proper boundary conditions for the reactor core thermal-hydraulic analysis. These four modules are coupled with each other by using MPI and calculate simultaneously on a cluster workstation. ACT was applied to analyzing a sodium experiment performed at JNC, which simulated the natural circulation decay heat removal under PRACS and DRACS operation condition. In the experiment, not only inter-wrapper flows but also reverses flows in the fuel assemblies were observed. ACT succeeded in simulating such complicated phenomena.
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Bottoni, Maurizio, Claudio Bottoni y John Scanu. "Molecular Dynamic Simulation of Sodium in 7-Pin LMFBR Bundle Under Hypothetical Accident Conditions". En 14th International Conference on Nuclear Engineering. ASMEDC, 2006. http://dx.doi.org/10.1115/icone14-89144.

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In the frame of safety analysis of liquid metal fast breeder reactors (LMFBRs) under hypothetical Unprotected Loss of Flow (ULOF) conditions two-phase flow of sodium is simulated in a 7-pin bundle, with hexagonal lattice. Molecular dynamics, with the application of the Direct Simulation Monte Carlo (DSMC) method, and a macroscopic model describing rewetting sequences due to the flow of a sodium liquid film along the pin surfaces, are applied to simulate the coolant in the bundle. The pin surfaces and the inner surface of the hexagonal canning are treated in the Monte Carlo simulation as diffusively reflecting surfaces. Collisions of sodium molecules are computed with the “hard-sphere” model. With respect to previous work the following improvements of the computational code were made: i) The full bundle is simulated, thus allowing for asymmetries, like a skewed power distribution, to be accounted for; ii) A pin model calculates detailed temperature distributions in the pins, so that temperature boundary conditions are computed and not imposed; iii) Post processing visualisation of computed results was developed. An out of pile sodium boiling experiment run at the Nuclear Research Center of Karlsruhe, Germany, is simulated and conclusions are drawn about the applicability of the methodology in computer codes dedicated to breeder reactors safety analysis.
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Madaras, Attila, Sa´ndor Deme, Zolta´n Ho´zer, Edit La´ng, Istva´n Ne´meth, Tama´s Pa´zma´ndi y Pe´ter Sza´nto´. "A New Simulation Code for Analyzing Loss of Coolant Accidents in VVER-440/213 Reactors Concerning Activity Transport". En 17th International Conference on Nuclear Engineering. ASMEDC, 2009. http://dx.doi.org/10.1115/icone17-75306.

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In the Atomic Energy Research Institute, Budapest, Hungary a computer code for modelling the in-containment fission product related processes of a design basis LOCA in VVER-440/213 type nuclear reactors is under development. The model is based on the lumped-parameter approach (the total volume of the simulated containment is divided into distinct, connected sub-volumes in which the parameters are assumed to be homogenous). The structural and functional models of the adequate reactor units are implemented in the code. The main considered physical processes of the fission product elements are radioactive decay, transport by gas flows, removal from the containment atmosphere by adsorption to wall surfaces and wash-out. In order to test the abilities of the code we performed sample calculations for the units of the Paks Nuclear Power Plant, Hungary. In this study the discussion of the first results is presented following a summary of the basics of the physical models implemented in the code.
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Huang, Shanfang, Yaopeng Gong, Chao Li, Ruilong Liu, Jiageng Wang y Kan Wang. "Numerical Simulation for Nuclear Engineering Education: A Case Study in a Course “Advanced Nuclear Reactor Thermal Analysis”". En 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81042.

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Nuclear energy is an important way to solve energy shortage and pollution problems today. Therefore, China is vigorously developing nuclear energy and is facing huge demands for talents. However, nuclear engineering education has been severely hampered, particularly in its experiment aspect because of the Fukushima accident. The original sub-critical nuclear reactor at Tsinghua University (THU) was stopped, forcing students to use computers to conduct relevant nuclear simulation experiments. The emerging of supercomputers and commercial numerical simulation softwares has provided enough hardware and software support for complex calculations required in numerical simulation of nuclear reactors. Thus numerical simulation could be integrated into nuclear engineering education. With its ease of use, quick acquisition and direct visualization of results, numerical simulation can save much time and money. Besides, it is convenient to change simulation conditions which is helpful for academic research and talent development. This paper starts with THU’s nuclear engineering talent training mode, and taking the course “Advanced Nuclear Reactor Thermal Analysis” as an example, discusses the applications of numerical simulation softwares FLUENT and COBRA-TF in this course. Finally, the analysis shows that numerical simulation performs well in conceptual understanding and experimental design. It plays a significant role in nuclear engineering education, which provides an important reference for the new mode of nuclear engineering education.
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Janosy, Janos Sebestyen, Andras Kereszturi, Gabor Hazi, Jozsef Pales y Endre Vegh. "Real-Time 3D Simulation of a Pressurized Water Nuclear Reactor". En 2010 12th International Conference on Computer Modelling and Simulation. IEEE, 2010. http://dx.doi.org/10.1109/uksim.2010.83.

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Mladin, Mirea y Daniela Mladin. "Simulation of B9401 Test in the RD-14M Experimental Facility With CATHARE2". En 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29320.

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RD-14M experimental facility is a full vertical-scale representation of a CANDU heat transport system, that was used as a benchmark data generating facility within the frame of IAEA’s Technical Working Group on Advanced Technologies for HWRs. RD-14M Large-Loss Of Coolant Accident (LOCA) test B9401 simulating HWR LOCA behaviour that was conducted by Atomic Energy of Canada Ltd (AECL) was selected for an international standard problem exercise. The aim was the intercomparison and validation of computer codes for thermalhydraulics safety analyses, using both codes originating within the HWR technology and also different versions of RELAP5 code, the later being developed for transient simulation of light water reactor coolant systems during postulated accidents. A report was published by IAEA in 2004. The Code for Analysis of THermalhydraulics during an Accident of Reactor and safety Evaluation (CATHARE), is also a LWR safety analysis code, developed jointly by AREVA_NP (reactors vendor), CEA (the French Atomic Energy Commission), EDF (the French electricity utility) and IRSN (the French Nuclear Safety Institute). The paper presents a model created for the RD-14M facility with CATHARE2, and the code application to the B9401 test. From the extensive series of test results available, several were selected to be compared to corresponding calculated evolutions. In some cases, our results are placed among those produced by the participants to the international standard problem with other codes.
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Qi, Yingxia y Minoru Takahashi. "Computer Simulation of Diffusion of Pb-Bi Eutectic in Liquid Sodium by Molecular Dynamics Method". En 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22236.

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Lead-bismuth eutectic is a potential candidate for coolant of secondary loops of sodium-cooled fast breeder reactors (FBR). The studies on the diffusion of liquid Pb-Bi in liquid Na are carried out corresponding to the case that liquid Pb-Bi leaks to liquid Na by accident. As the diffusion processes are the results of atomic motions, molecular dynamics method has been used to study the diffusion process. The self-diffusion coefficients of pure liquid Pb and Na, and liquid Pb-Bi are calculated and compared with ones by the empirical equations. The discrepancy between them could be eliminated by changing the densities of the liquids. The diffusion of lead-bismuth in sodium is simulated based on the changed densities under which the self-diffusion coefficients of individual liquid metals are close to those by the empirical equations. The simulation results show that the diffusion process of liquid Pb-Bi in liquid Na is a heat releasing process and the density of ternary liquid Na-Pb-Bi is higher than the average value of the densities of liquid Na and liquid Pb-Bi. It is also found that the diffusion coefficients of liquid Pb-Bi in liquid Na are much higher than their self-diffusion coefficients, indicating that liquid Pb-Bi are easy and quickly to diffuse in liquid Na. However, the diffusion coefficient of liquid Na is decreased due to the existence of liquid Pb-Bi, implying that liquid Na-Pb-Bi have a higher viscosity than that of pure liquid Na.
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Informes sobre el tema "Nuclear reactors – Computer simulation"

1

Kamegai, M. Computer simulation of underwater nuclear events. Office of Scientific and Technical Information (OSTI), septiembre de 1986. http://dx.doi.org/10.2172/5275001.

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Martin, William. Implementation of On-the-Fly Doppler Broadening in MCNP5 for Multiphysics Simulation of Nuclear Reactors. Office of Scientific and Technical Information (OSTI), noviembre de 2012. http://dx.doi.org/10.2172/1058919.

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Kroeger, P. G., R. J. Kennett, J. Colman y T. Ginsberg. THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors. Office of Scientific and Technical Information (OSTI), octubre de 1991. http://dx.doi.org/10.2172/6239042.

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Ferguson, Jim, Peter Brown y Larry Jacobsen. Computer Simulation of Nuclear Well Logging Devices Final Report CRADA No. TC-824-94F. Office of Scientific and Technical Information (OSTI), abril de 1998. http://dx.doi.org/10.2172/1438796.

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Ferguson, James M., Larry Jacobson y Daniel Johnson. Computer Simulation of Nuclear Well Logging Devices: Final Report CRADA No. TC-1114-95. Office of Scientific and Technical Information (OSTI), octubre de 2000. http://dx.doi.org/10.2172/1410084.

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Lee, C. H. y H. C. Lee. Verification and Validation of High-Fidelity Multi-Physics Simulation Codes for Advanced Nuclear Reactors (Year 2). Office of Scientific and Technical Information (OSTI), junio de 2014. http://dx.doi.org/10.2172/1134005.

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Rosa, M. P. y M. Z. Podowski. Modeling and numerical simulation of oscillatory two-phase flows, with application to boiling water nuclear reactors. Office of Scientific and Technical Information (OSTI), septiembre de 1995. http://dx.doi.org/10.2172/107760.

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