Relevant bibliographies by topics / Nuclear fuels – Computer programs

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Academic literature on the topic 'Nuclear fuels – Computer programs'

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

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Journal articles on the topic "Nuclear fuels – Computer programs"

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FAGHIHI, F., and M. R. NEMATOLLAHI. "NUCLEAR FUEL DEPLETION ANALYSIS USING MATLAB SOFTWARE." International Journal of Modern Physics C 17, no. 06 (June 2006): 805–15. http://dx.doi.org/10.1142/s0129183106009229.

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Coupled first order IVPs are frequently used in many parts of engineering and sciences. In this article, we presented a code including three computer programs which are joint with the Matlab software to solve and plot the solutions of the first order coupled stiff or non-stiff IVPs. Some engineering and scientific problems related to IVPs are given and fuel depletion (production of the 239 Pu isotope) in a Pressurized Water Nuclear Reactor (PWR) are computed by the present code.
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Seydaliev, M., and D. Caswell. "CORBA AND MPI-BASED “BACKBONE” FOR COUPLING ADVANCED SIMULATION TOOLS." AECL Nuclear Review 3, no. 2 (December 1, 2014): 83–90. http://dx.doi.org/10.12943/anr.2014.00036.

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There is a growing international interest in using coupled, multidisciplinary computer simulations for a variety of purposes, including nuclear reactor safety analysis. Reactor behaviour can be modeled using a suite of computer programs simulating phenomena or predicting parameters that can be categorized into disciplines such as Thermalhydraulics, Neutronics, Fuel, Fuel Channels, Fission Product Release and Transport, Containment and Atmospheric Dispersion, and Severe Accident Analysis. Traditionally, simulations used for safety analysis individually addressed only the behaviour within a single discipline, based upon static input data from other simulation programs. The limitation of using a suite of stand-alone simulations is that phenomenological interdependencies or temporal feedback between the parameters calculated within individual simulations cannot be adequately captured. To remove this shortcoming, multiple computer simulations for different disciplines must exchange data during runtime to address these interdependencies. This article describes the concept of a new framework, which we refer to as the “Backbone,” to provide the necessary runtime exchange of data. The Backbone, currently under development at AECL for a preliminary feasibility study, is a hybrid design using features taken from the Common Object Request Broker Architecture (CORBA), a standard defined by the Object Management Group, and the Message Passing Interface (MPI), a standard developed by a group of researchers from academia and industry. Both have well-tested and efficient implementations, including some that are freely available under the GNU public licenses. The CORBA component enables individual programs written in different languages and running on different platforms within a network to exchange data with each other, thus behaving like a single application. MPI provides the process-to-process intercommunication between these programs. This paper outlines the different CORBA and MPI configurations examined to date, as well as the preliminary configuration selected for coupling 2 existing safety analysis programs used for modeling thermal–mechanical fuel behavior and fission product behavior respectively. In addition, preliminary work in hosting both the Backbone and the associated safety analysis programs in a cluster environment are discussed.
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Okunev, V. S. "Fundamentally New Composite Materials of Fast Reactors Made on the Basis of Nanotechnology." Key Engineering Materials 887 (May 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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Byers, J. L. "Desktop Failure Analysis on a Microcomputer Using Weibull, Lognormal, and Renewal Data." Journal of Engineering for Gas Turbines and Power 112, no. 2 (April 1, 1990): 233–36. http://dx.doi.org/10.1115/1.2906168.

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Gas turbine components and parts are widely known to have many failure modes for which the failures correlate in either the Weibull or Lognormal probability distributions. This paper describes a typical case, which is handled by the new computer programs now being used by the U. S. Navy. These programs have brought the capability to make such analyses directly to the designer or analyst’s desk instead of having to be sent off to a central computer to wait in line. The programs are interactive with the user and extremely user friendly. Uses are expanding to cover almost every area in the life cycle of gas turbines where it would be beneficial to forecast future failures. This makes the programs useful to managers, logisticians, life cycle cost analysts, and a host of others. Wide applicability of the methods assures usage outside of the gas turbine field.
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Perz, E. "A Computer Method for Thermal Power Cycle Calculation." Journal of Engineering for Gas Turbines and Power 113, no. 2 (April 1, 1991): 184–89. http://dx.doi.org/10.1115/1.2906543.

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This paper describes a highly flexible computer method for thermodynamic power cycle calculations (PCC). With this method the user can model any cycle scheme by selecting components from a library and connecting them in an appropriate way. The flexibility is not restricted by any predefined cycle schemes. A power cycle is mathematically represented by a system of algebraic equations. The structure of mathematical cycle models as well as different approaches to set up and solve the resulting equations with computer programs are discussed in the first section. The second section describes the developed method. The mass and energy balance equations are set up and solved with a semiparallel algorithm. As input only the cycle’s topology and component parameters must be entered. Information about the calculation sequence and the convergence method can be omitted completely. The example of two simple steam cycles demonstrates the applied technique. The method requires only a few, if any, iterations. Calculation time and storage requirements can be kept low enough to calculate even very complex cycles on personal computers. At the end of the paper input data and results for a complex cycle scheme as it may occur in reality are given to demonstrate the performance finally.
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Fishbach, L. H., and S. Gordon. "NNEPEQ—Chemical Equilibrium Version of the Navy/NASA Engine Program." Journal of Engineering for Gas Turbines and Power 111, no. 1 (January 1, 1989): 114–16. http://dx.doi.org/10.1115/1.3240205.

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The Navy NASA Engine Program, NNEP, developed in 1975, currently is in use at a large number of government agencies, commercial companies, and universities. This computer code has been used extensively to calculate the design and off-design (matched) performance of a broad range of turbine engines, ranging from subsonic turboprops to variable cycle engines for supersonic transports. Recently, there has been increased interest in applications that NNEP was not capable of simulating, namely, high Mach applications, alternate fuels including cryogenics, and cycles such as the gas generator air-turbo-rocket (ATR). In addition, there is interest in cycles employing ejectors such as for military fighters. New engine component models had to be created for incorporation into NNEP, and it was also found necessary to include chemical dissociation effects of high-temperature gases. This paper discusses the incorporation of these extended capabilities of NNEP and illustrates some of the effects of these changes.
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Zhao, Xiaolu. "Stream Function Solution of Transonic Flow Along S2 Streamsurface of Axial Turbomachines." Journal of Engineering for Gas Turbines and Power 108, no. 1 (January 1, 1986): 138–43. http://dx.doi.org/10.1115/1.3239860.

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Based on Wu’s general equations of 3-D turbomachine flow, expressed with respect to nonorthogonal curvilinear coordinates, the conservative stream-function formulations of transonic flow along S2 streamsurface have been discussed. The problem of mixed flow can be solved by the use of the artificial compressibility method, and the passage shock on the S2 streamsurface can be captured. The distribution of the fluid velocity from hub to shroud can be obtained directly by integrating the velocity gradient equation, after the principal equation has been solved, so that the difficulty of the nonuniqueness of density-mass flux relation is avoided. The density is determined after the velocity has been obtained. Two computer programs have been coded; one can be used to compute the hybrid or design problems, the other is suitable to compute the analysis problem. The former has been used to compute the transonic flow field along a mean S2 streamsurface in the DFLVR compressor rotor at design speed. The numerical results agree well with L2F measurement data.
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Barber, D. H. "Implementation of A Gibbs Energy Minimizer In A Fission-Product Release Computer Program." AECL Nuclear Review 2, no. 1 (June 1, 2013): 39–48. http://dx.doi.org/10.12943/anr.2013.00005.

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SOURCE 2.0 is the Canadian computer program for calculating fractional release of fission products from the UO2 fuel matrix. In nuclear accidents, fission-product release from fuel is one of the physical steps required before radiation dose from fission products can affect the public. Fission-product release calculations are a step in the analysis path to calculating dose consequences to the public from postulated nuclear accidents. SOURCE 2.0 contains a 1997 model of fission-product vaporization by B.J. Corse et al. based on lookup tables generated with the FACT computer program. That model was tractable on computers of that day. However, the understanding of fuel thermochemistry has advanced since that time. Additionally, computational resources have significantly improved since the time of the development of the Corse model and now allow incorporation of the more-rigorous thermodynamic treatment. Combining the newer Royal Military College of Canada (RMC) thermodynamic model of irradiated uranium dioxide fuel, a new model for fission-product vaporization from the fuel surface, a commercial user-callable thermodynamics subroutine library (ChemApp), an updated nuclide list, and updated nuclear physics data, a prototype computer program based on SOURCE IST 2.0P11 has been created that performs thermodynamic calculations internally. The resulting prototype code (with updated and revised data) provides estimates of 140La releases that are in better agreement with experiments than the original code version and data. The improvement can be quantified by a reduction in the mean difference between experimental and calculated release fractions from 0.70 to 0.07. 140La is taken to be representative of “low-volatile” fission products. To ensure that the existing acceptable performance for noble gases and volatile fission products is not adversely affected by the changes, comparisons were also made for a representative noble gas, 85Kr, and a representative volatile fission-product, 134Cs. These nuclides have the largest dataset in the SOURCE 2.0 validation test suite. This improvement provides increased confidence in the safety margin for equipment qualification in Loss-of-Coolant Accidents with Loss of Emergency Core Cooling.
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Bachhav, Mukesh, Brandon Miller, Jian Gan, Dennis Keiser, Ann Leenaers, S. Van den Berghe, and Mitchell K. Meyer. "Microstructural Changes and Chemical Analysis of Fission Products in Irradiated Uranium-7 wt.% Molybdenum Metallic Fuel Using Atom Probe Tomography." Applied Sciences 11, no. 15 (July 27, 2021): 6905. http://dx.doi.org/10.3390/app11156905.

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Understanding the microstructural and phase changes occurring during irradiation and their impact on metallic fuel behavior is integral to research and development of nuclear fuel programs. This paper reports systematic analysis of as-fabricated and irradiated low-enriched U-Mo (uranium-molybdenum metal alloy) fuel using atom probe tomography (APT). This study is carried out on U-7 wt.% Mo fuel particles coated with a ZrN layer contained within an Al matrix during irradiation. The dispersion fuel plates from which the fuel samples were extracted are irradiated at Belgian Nuclear Research Centre (SCK CEN) with burn-up of 52% and 66% in the framework of the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) project. The APT studies on U-Mo particles from as-fabricated fuel plates enriched to 19.8% revealed predominantly γ-phase U-Mo, along with a network of the cell boundary decorated with α-U, γ’-U2Mo, and UC precipitates along the grain boundaries. The corresponding APT characterization of irradiated fuel samples showed formation of fission gas bubbles enriched with solid fission products. The intermediate burnup sample showed a uniform distribution of the typical bubble superlattice with a radius of 2 nm arranged in a regular lattice, while the high burnup sample showed a non-uniform distribution of bubbles in grain-refined regions. There was no evidence of remnant α-U, γ’-U2Mo, and UC phases in the irradiated U-7 wt.% Mo samples.
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El-Sayed, Y. M. "A Second-Law-Based Optimization: Part 2—Application." Journal of Engineering for Gas Turbines and Power 118, no. 4 (October 1, 1996): 698–703. http://dx.doi.org/10.1115/1.2816982.

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This paper deals with an application of the second-law-based optimization methodology described in Part 1 of this paper. The methodology is applied to a simple energy system of 10 components and 19 manipulated decision parameters. The system is treated once as a single-purpose combined cycle and once as a power-heat cogenerating system. The results of the application are summarized and evaluated. The required computer programs are described. Further investigations are suggested.
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Dissertations / Theses on the topic "Nuclear fuels – Computer programs"

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LEVY, DENISE S. "Contribuição para informatização dos programas de proteção radiológica para instalações radiativas." reponame:Repositório Institucional do IPEN, 2012. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10144.

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Dissertação (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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YORIYAZ, HELIO. "Implementacao de queima espacial modificando o programa nodal baseado no metodo de elementos finitos e matriz resposta." reponame:Repositório Institucional do IPEN, 1986. http://repositorio.ipen.br:8080/xmlui/handle/123456789/9855.

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Mattos, Carlos Eduardo. "Estudo de modelos para o comportamento a altas queimas de varetas combustível urânio - 7% gadolínio para reatores a água leve pressurizada: avaliação dos parâmetros para prolongamento do tempo de queima do núcleo." Universidade de São Paulo, 2018. http://www.teses.usp.br/teses/disponiveis/85/85134/tde-17052018-160542/.

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O objetivo deste trabalho é verificar os resultados fornecidos pelo programa computacional FRAPCON-3, hoje na versão 5, utilizado no processo de simulação do comportamento de varetas combustíveis de reatores refrigerados a água pressurizada (Pressurized Water Reactor PWR), sob situações operacionais de regime permanente, em condições de alta queima. Para realizar a verificação, foi utilizada a base de dados FUMEX-III, que fornece dados relativos a experimentos realizados com diversos tipos de combustíveis nucleares, submetidos a diversas condições operacionais. Através dos resultados obtidos na simulação do programa FRAPCON-3.5 e da sua comparação com os dados experimentais da base FUMEX-III, foi possível constar que o programa possui boa capacidade de predizer o comportamento operacional da vareta combustível em regime permanente a altas queimas. O trabalho consiste também em verificar a correlação entre UO2 e UO2-7%Gd2O3 na análise dos modelos que simulam o comportamento das pastilhas combustível. A adição do óxido de gadolínio ou gadolínia (Gd2O3), constitui-se na opção tecnológica mais solidamente consagrada e hoje comum em várias centrais nucleares. Por meio dos resultados obtidos nas simulações computacionais foram apresentadas e discutidas a influência das propriedades do UO2 e UO2-7%Gd2O3, quanto à temperatura no centro do combustível, liberação de gás de fissão na vareta, temperatura média do revestimento, volume interno e pressão interna da vareta combustível.
The objective of this work is to verify the results provided by the computer program FRAPCON-3, now in version 5, used in the simulation process of the behavior of fuel rods of pressurized water reactors - PWR permanent, in conditions of high burn. In order to carry out the verification, the FUMEX-III database was used, which provides data on experiments performed with different types of nuclear fuel, under different operating conditions. The results obtained in the simulation of the FRAPCON-3.5 program and its comparison with the experimental data of the FUMEX-III base showed that the program has a good ability to predict the operational behavior of the fuel rod in a steady state at high burn. The work also consists in verifying the correlation between UO2 and UO2-7%Gd2O3 in the analysis of models that simulate the behavior of fuel pellets. The addition of gadolinium oxide (Gd2O3) constitutes the most solidly established and now common technological option in several nuclear power plants. The influence of the properties of UO2 and UO2-7%Gd2O3 on the temperature at the center of the fuel, fission gas release on the rod, average coating temperature, internal volume and pressure were presented and discussed. of the fuel rod.
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Sadeghi, Mohammad Mehdi 1959. "SYMBOLIC MANIPULATION IN REACTOR PHYSICS." Thesis, The University of Arizona, 1986. http://hdl.handle.net/10150/275520.

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HIROMOTO, MARIA Y. K. "PSINCO-um programa para calculo da distribuicao de potencia e supervisao do nucleo de reatores nucleares, utilizando sinais de detetores tipo 'SPD'." reponame:Repositório Institucional do IPEN, 1998. http://repositorio.ipen.br:8080/xmlui/handle/123456789/10706.

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Dissertacao (Mestrado)
IPEN/D
Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP
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Skroch, Eric M. "How to optimally interdict a belligerent project to develop a nuclear weapon." Thesis, Monterey, Calif. : Springfield, Va. : Naval Postgraduate School ; Available from National Technical Information Service, 2004. http://library.nps.navy.mil/uhtbin/hyperion/04Mar%5FSkroch.pdf.

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MUNIZ, RAFAEL O. R. "Análise neutrônica e especificação técnica para o combustível a dispersão UMo-Al com adição de veneno queimável." reponame:Repositório Institucional do IPEN, 2015. http://repositorio.ipen.br:8080/xmlui/handle/123456789/25671.

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Tese (Doutorado em Tecnologia Nuclear)
IPEN/T
Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP
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Lage, Aldo Márcio Fonseca. "Modelagem geométrica computacional das etapas de prensagem e sinterização de pastilhas e de laminação de placas combustíveis em dispersão de microesferas de (Th,25%U) O2 em matriz de aço inoxidável." CNEN - Centro de Desenvolvimento da Tecnologia Nuclear, Belo Horizonte, 2005. http://www.bdtd.cdtn.br//tde_busca/arquivo.php?codArquivo=46.

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Nenhuma
Neste trabalho foi realizada a modelagem geométrica computacional das Cetapas de prensagem e sinterização da pastilha e da laminação da placa de combustível nuclear contendo microesferas de (Th,25%U)O2 dispersas em matriz de aço inoxidável com o objetivo de avaliar a distribuição destas microesferas nas diversas etapas do processamento. As regras de modelagem foram desenvolvidas baseadas nos parâmetros de cada etapa da fabricação da placa combustível. Para isto foram obtidas placas através do processamento por laminação de molduras de chapas de aço inoxidável, contendo pastilha fabricadas com microesferas de (Th,25%U)O2 com carregamentos de 10, 20 e 40% em peso de combustível disperso em matriz de aço inoxidável. Os dados das placas com carregamentos de 30 e 50% foram obtidos por interpolação da curva. As microesferas, obtidas pelo processo sol-gel, foram previamente secas, reduzidas e sinterizadas a 1700oC, durante 2 horas, sob atmosferas de hidrogênio. As microesferas sinterizadas alcançaram uma densidade de cerca de 98% da densidade teórica, e possuem um diâmetro médio de cerca de 300 mm e uma elevada resistência à fratura, de aproximadamente 40 N/microesfera. As regras implementadas neste modelo foram aplicadas nas coordenadas dos centros das esferas virtuais, que simulam as microesferas combustíveis de (Th,25%U)O2, obtendo-se novas coordenadas espaciais para cada uma delas nas etapas de prensagem e sinterização da pastilha e da laminação da placa combustível. Este modelo foi projetado com o uso de técnicas de análise de sistema estruturada, implementado utilizando a linguagem de programação Delphi e os resultados visualizados através do programa AutoCAD. Os resultados do modelo foram validados comparando-se as frações volumétricas experimentais em cada um dos carregamentos estudados com as frações simuladas. Este trabalho será de grande valia para o estudo do carregamento de microesferas na placa combustível, permitindo obter um combustível de elevado desempenho mecânico, térmico e neutrônico mesmo em mais alto carregamento.
The computational geometric modeling of the pressing, sintering and lamination stages for nuclear fuel plates composed by (Th,25%U)O2, microspheres dispersed into stainless steel matrix has been done in order to investigate the microspheres distribution in the various processing stages. The modeling standards were based on the parameters related to each fuel plate manufacturing stage. Accordingly, the plates were obtained through lamination processing of stainless steel plate frames comprising (Th,25%U)O2 microspheres pellets dispersed into stainless steel powder with loading of 10, 20 and 40% of microspheres dispersed into stainless steel matrix. The data for plates with loading of 30 and 50% have been obtained by linear interpolation. The microspheres produced by the sol-gel method were previously reduced and sintered at 1700 0C during 2 hours at hydrogen atmosphere. These sintered microspheres have reached about 98% of the theoretical density, with a mean diameter of 300 mm and a high resistance to fracture, near to 40 N/microsphere. The implemented standards in this model were applied at the virtual spheres center coordinates, which simulate the (Th,25%U)O2 fuel microspheres, and generate the new spatial coordinates to each of them in the pressing, sintering and lamination stages. This model was developed using structured system analysis techniques and it has been implemented using the Delphi programming language. The results were displayed through the AutoCAD program and validated comparing the experimental volumetric fractions in each of the studied loading, with the simulated fractions. The results indicate that this work could be a powerful tool in the investigation of microspheres loading in the fuel plate, allowing the attainment of a high mechanical and neutronic performance fuel, even for higher level loading.
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Bush, Joshua. "Phase Field Modeling of Thermotransport in Multicomponent Systems." Master's thesis, University of Central Florida, 2012. http://digital.library.ucf.edu/cdm/ref/collection/ETD/id/5152.

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Nuclear and gas turbine power plants, computer chips, and other devices and industries are running hotter than ever for longer than ever. With no apparent end to the trend, the potential arises for a phenomenon known as thermotransport to cause undesirable changes in these high temperature materials. The diffuse-interface method known as the phase-field model is a useful tool in the simulation and prediction of thermotransport driven microstructure evolution in materials. The objective of this work is to develop a phase-field model using practical and empirical properties of thermodynamics and kinetics for simulating the interdiffusion behavior and microstructural evolution of single and multiphase binary alloy system under composition and/or temperature gradients. Simulations are carried out using thermodynamics and kinetics of real systems, such as the U-Zr solid metallic fuel, with emphasis on the temperature dependencies of the kinetics governing diffusional interactions in single-phase systems and microstructural evolution in the presence of multiple driving forces in multi-phase systems. A phase field model is developed describing thermotransport in the γ phase of the U-Zr alloy, a candidate for advanced metallic nuclear fuels. The model is derived using thermodynamics extracted from the CALPHAD database and temperature dependent kinetic parameters associated with thermotransport from the literature. Emphasis is placed upon the importance of the heat of transport, Q*, and atomic mobility, β. Temperature dependencies of each term are estimated from empirical data obtained directly from the literature, coupled with the textbook phenomenological formulae of each parameter. A solution is obtained via a finite volume approach with the aid of the FiPy® partial differential equation solver. Results of the simulations are described based on individual flux contributions from the gradients of both composition and temperature, and are found to be remarkably similar to experimental results from the literature. In an additional effort the thermotransport behavior of a binary two-phase alloy is modeled, for the first time, via the phase-field method for a two-phase (γ + β) U-Zr system. The model is similarly built upon CALPHAD thermodynamics describing the γ and β phases of the U-Zr system and thermotransport parameters for the γ phase from literature. A parametric investigation of how the heats of transport for U and Zr in the β phase affect the redistribution is performed, and the interplay between system kinetics and thermodynamics are examined. Importantly, a strict control over the microstructure that is placed into the temperature gradient (at t=0) is used to eliminate the randomness associated with microstructural evolution from an initially unstable state, allowing an examination of exactly how the β phase thermotransport parameters affect the redistribution behavior of the system. Results are compared to a control scenario in which the system evolves only in the presence of thermodynamic driving forces, and the kinetic parameters that are associated with thermotransport are negligible. In contrast to the single-phase simulations, in the presence of a large thermodynamic drive for phase transformation and stability, the constituent redistribution caused by the thermotransport effect is comparatively smaller.
ID: 031001396; System requirements: World Wide Web browser and PDF reader.; Mode of access: World Wide Web.; Title from PDF title page (viewed June 3, 2013).; Thesis (M.S.M.S.E.)--University of Central Florida, 2012.; Includes bibliographical references (p. 50-53).
M.S.M.S.E.
Masters
Materials Science Engineering
Engineering and Computer Science
Materials Science and Engineering
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Liposinovic, Marita. "Usability Principles for (Re)Design of User Interface of Emergency Handling Programs : Case Study on a Tool for Decision Support amidst a Nuclear Emergency: RASTEP." Thesis, KTH, Skolan för elektroteknik och datavetenskap (EECS), 2020. http://urn.kb.se/resolve?urn=urn:nbn:se:kth:diva-277936.

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Highly stressful situations have a prominent influence on the human psyche and it is crucial to take this into account when designing and evaluating software used predominantly in nerve wrecking situations. Lack of principles for design and evaluation of design of emergency handling software programs allows for graphic interface solutions that leave room for mistakes and slips in an environment where making an error can have catastrophic consequences. The study proposes a table of principles for evaluation and (re)design of user interface of programs used in highly stressful situations. The principles are grouped based on the human capability hindered by stress. Said principles were evaluated with the help of a case study on a tool for decision support (RASTEP) when facing a nuclear emergency. Statements on usability issues with RASTEP were collected through open group interviews and participant observation and pattern matched with proposed principles. Result of the pattern matching showed the disposition of usability issues per hindered capability following the same pattern as the one of the principles. Furthermore, all but one collected statement were identified as belonging to a group of hindered capabilities and were identified as violating at least one of the proposed principles. This points to the table of principles being able to serve as a tool in evaluating usability of graphic user interface of an existing emergency handling software and identifying its usability issues.
Mycket stressiga situationer kan ha en betydande påverkan på det mänskliga psyket, och det är avgörande att ta hänsyn till när man skapar och utvärderar mjukvara som främst används i pressade situationer. Brist på principer för utvärdering av design av nödhanteringsprogram möjliggör gränssnittslösningar som kan ge upphov till misstag i en miljö där felsteg kan ha katastrofala konsekvenser. Studien föreslår en tabell med principer för utvärdering och (re)design av användargränssnitt för program som används i mycket stressande situationer. Principerna är grupperade efter mänskliga förmågor som hindras av stress. Nämnda principer utvärderades genom en fallstudie av ett verktyg för beslutsstöd (RASTEP) vid kärnkraftsolycka. Vittnesmål om användbarhetsproblem med RASTEP insamlades genom öppna gruppintervjuer och deltagande observation (participant observation), och patternmatchades med de föreslagna principerna. Resultatet av patternmatchningen visade att fördelningen av användbarhetsproblem per hindrad förmåga följde samma mönster som fördelningen av principerna. Dessutom identifierades alla insamlade vittnesmål utom ett tillhöra en grupp av hindrade förmågor och identifierades bryta mot åtminstone en av de föreslagna principerna. Detta tyder på att tabellen med principerna kan användas som ett verktyg för att utvärdera användbarheten av gränssnitt hos ett befintligt nödhanteringsprogram och identifiera dess användbarhetsproblem.
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Books on the topic "Nuclear fuels – Computer programs"

1

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

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Kikō, Genshiryoku Anzen Kiban. Kakunenryō shisetsu hibaku hyōka shien shisutemu no seibi. [Tokyo]: Genshiryoku Anzen Kiban Kikō, 2005.

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Chan, M. K. User's manual for FIRIN: A computer code to estimate accidental fire and radioactive airborne releases in nuclear fuel cycle facilities. Washington, DC: Division of Fuel Cycle and Material Safety, Office of Nuclear Materials Safety and Safeguards, U.S. Nuclear Regulatory Commission, 1989.

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

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Bilanovic, Z. Neutron-photon energy deposition in CANDU reactor fuel channels: A comparison of modelling techniques using ANISN and MCNP computer codes. Chalk River, Ont: System Chemistry and Corrosion Branch, Chalk River Laboratories, 1994.

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LeNeveu, D. M. Analysis specifications for the CC3 vault model. Pinawa, Man: AECL Research, 1994.

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Commission, Canadian Nuclear Safety. Security programs for category I or II nuclear material or certain nuclear facilities. [Ottawa]: Canadian Nuclear Safety Commission, 2003.

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Azmy, Yousry. Nuclear computational science: A century in review. Dordrecht: Springer, 2010.

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Dougan, Kenneth D. Programmer's guide for the CC3 computer models of the concept for disposal of Canada's nuclear fuel waste. Pinawa, Man: Environmental and Safety Assessment Branch, Whiteshell Laboratories, 1996.

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Dougan, K. D. User's manual for the CC3 computer models of the concept for disposal of Canada's nuclear fuel waste. Pinawa, Man: Environmental and Safety Assessment Branch, Whiteshell Laboratories, 1995.

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Book chapters on the topic "Nuclear fuels – Computer programs"

1

Heyde, Kris L. G. "Some Computer Programs." In The Nuclear Shell Model, 336–76. Berlin, Heidelberg: Springer Berlin Heidelberg, 1994. http://dx.doi.org/10.1007/978-3-642-79052-2_10.

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Heyde, Kris L. G. "Some Computer Programs." In Springer Series in Nuclear and Particle Physics, 298–326. Berlin, Heidelberg: Springer Berlin Heidelberg, 1990. http://dx.doi.org/10.1007/978-3-642-97203-4_10.

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Zerhouni, Fatima Zohra, M’hammed Houari Zerhouni, Mansour Zegrar, and Amine Boudghene Stambouli. "Study of Hybrid Sustainable Energy System Based on PEM Fuel Cells and Photovoltaic-Module Power Generator." In Sustainable ICTs and Management Systems for Green Computing, 399–422. IGI Global, 2012. http://dx.doi.org/10.4018/978-1-4666-1839-8.ch018.

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The computer is the greatest innovation of the 20th century. It has changed our lives. It executes tasks with precision. There is no limit with what we can do with software. Computers are seductive. Companies and students cannot work without them. They help students to perform mathematical computations. It is very important that mathematical ideas are expressed in computer programs in order to have theoretical results and to verify them practically. Nowadays, the development of new and non-polluting energy producing and energy-storage systems is a great challenge for scientists. An alternative to the nuclear and fossil fuel power is renewable energy technologies. Due to ever-increasing energy consumption, rising public awareness of environmental protection, and steady progress in power deregulation, alternative (i.e., renewable and fuel cell based) distributed generation systems have attracted increased interest. There is an accelerating world demand for environmentally friendly power. Among the renewable energy sources, the Photovoltaic (PV) energy is the most promising candidate for research and development for large scale users. Fuel cells have been receiving a lot of attention lately due to their potential of becoming a new energy source with a large range of applications. Fuel cells can be incorporated with other components to create high efficiency industrial power plants. Fuel cells permit clean and efficient energy production. The purpose of the work is to optimize the system’s operation. The main reason to build described system is to supply stand-alone systems using renewable energy sources. Therefore, the power plant has to produce energy independent of any weather fluctuations. Integrating photovoltaic energy sources with fuel cells, as a storage device replacing the conventional lead-acid batteries, leads to a non-polluting reliable energy source. In this chapter, an energy system comprising different energy sources, namely PV and fuel cells, is proposed. Photovoltaic cells coupled with electrolytic devices can be used to produce hydrogen and oxygen in a sustainable manner. With the produced hydrogen from the electrolysis process, it is possible to generate electricity through fuel cells. Photovoltaic panels in particular can provide a good source of producing green electricity. It is autonomous, its operation does not pollute the atmosphere, and it is an inexhaustible and renewable source with great reliability. The simulation program developed also allows the exportation of different configurations. The experimental system described has permitted the validation of the proposed method.
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"COMPUTER PROGRAMS FOR PROBABILISTIC RISK ASSESSMENT." In Risk and Safety Analysis of Nuclear Systems, 179–95. Hoboken, NJ, USA: John Wiley & Sons, Inc., 2011. http://dx.doi.org/10.1002/9781118043462.ch7.

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Blechschmidt, Ingo, and Stratis Vomvoris. "Relevance of underground rock laboratories for deep geological repository programs." In Geological Repository Systems for Safe Disposal of Spent Nuclear Fuels and Radioactive Waste, 113–42. Elsevier, 2017. http://dx.doi.org/10.1016/b978-0-08-100642-9.00005-0.

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De Luca, Domenico, Simone Di Pasquale, Marco Cherubini, Alessandro Petruzzi, and Gianni Bruna. "International Benchmark Activity in the Field of Sodium Fast Reactors." In Recent Advances on Numerical Simulations [Working Title]. IntechOpen, 2021. http://dx.doi.org/10.5772/intechopen.97812.

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Global interest in fast reactors has been growing since their inception in 1960 because they can provide efficient, safe, and sustainable energy. Their closed fuel cycle can support long-term nuclear power development as part of the world’s future energy mix and decrease the burden of nuclear waste. In addition to current fast reactors construction projects, several countries are engaged in intense R&D and innovation programs for the development of innovative, or Generation IV, fast reactor concepts. Within this framework, NINE is very actively participating in various Coordinated Research Projects (CRPs) organized by the IAEA, aimed at improving Member States’ fast reactor analytical simulation capabilities and international qualification through code-to-code comparison, as well as experimental validation on mock-up experiment results of codes currently employed in the field of fast reactors. The first CRP was focused on the benchmark analysis of Experimental Breeder Reactor II (EBR-II) Shutdown Heat Removal Test (SHRT-17), protected loss-of-flow transient, which ended in the 2017 with the publication of the IAEA-TECDOC-1819. In the framework of this project, the NINE Validation Process– developed in the framework of NEMM (NINE Evaluation Model Methodology) – has been proposed and adopted by most of the organizations to support the interpretation of the results calculated by the CRP participants and the understanding of the reasons for differences between the participants’ simulation results and the experimental data. A second project regards the CRP focused on benchmark analysis of one of the unprotected passive safety demonstration tests performed at the Fast Flux Test Facility (FFTF), the Loss of Flow Without Scram (LOFWOS) Test #13, started in 2018. A detailed nodalization has been developed by NINE following its nodalization techniques and the NINE validation procedure has been adopted to validate the Simulation Model (SM) against the experimental data of the selected test. The third activity deals with the neutronics benchmark of China Experimental Fast Reactor (CEFR) Start-Up Tests, a CRP proposed by the China Institute of Atomic Energy (CIAE) launched in 2018 the main objective of which is to improve the understanding of the start-up of a SFR and to validate the fast reactor analysis computer codes against CEFR experimental data. A series of start-up tests have been analyzed in this benchmark and NINE also proposed and organized a further work package focused on the sensitivity and uncertainty analysis of the first criticality test. The present chapter intends to summarize the results achieved using the codes currently employed in the field of fast reactor in the framework of international projects and benchmarks in which NINE was involved and emphasize how the application of developed procedures allows to validate the SM results and validate the computer codes against experimental data.
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Andersson, Kjell. "Methods for social dialogue in the establishment of radioactive waste management programs." In Geological Repository Systems for Safe Disposal of Spent Nuclear Fuels and Radioactive Waste, 741–59. Elsevier, 2017. http://dx.doi.org/10.1016/b978-0-08-100642-9.00026-8.

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Conca, James. "Environmental monitoring programs and public engagement for siting and operation of geological repository systems." In Geological Repository Systems for Safe Disposal of Spent Nuclear Fuels and Radioactive Waste, 667–709. Elsevier, 2017. http://dx.doi.org/10.1016/b978-0-08-100642-9.00024-4.

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Chairat, Aziza, Zhiya Duan, Olivier Fiquet, Carine Ablitzer, Laurent Cassayre, Hugues Vergnes, Pascal Floquet, and Xavier Joulia. "Modeling of a pyrolysis process for the elimination of epoxy resin from embedded nuclear fuels." In Computer Aided Chemical Engineering, 343–48. Elsevier, 2017. http://dx.doi.org/10.1016/b978-0-444-63965-3.50059-3.

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Khojasteh Salkuyeh, Yaser, and Thomas A. Adams. "Combining coal, natural gas, and nuclear heat for liquid fuels production with reduced CO2 emissions." In Computer Aided Chemical Engineering, 247–51. Elsevier, 2012. http://dx.doi.org/10.1016/b978-0-444-59519-5.50050-2.

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Conference papers on the topic "Nuclear fuels – Computer programs"

1

Wren, D. J., N. Popov, and V. G. Snell. "Validation of Physics and Thermalhydraulics Computer Codes for Advanced CANDU Reactor Applications." In 12th International Conference on Nuclear Engineering. ASMEDC, 2004. http://dx.doi.org/10.1115/icone12-49102.

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Atomic Energy of Canada Ltd. (AECL) is developing an Advanced CANDU Reactor (ACR) that is an evolutionary advancement of the currently operating CANDU 6 reactors. The ACR is being designed to produce electrical power for a capital cost and at a unit-energy cost significantly less than that of the current reactor designs. The ACR retains the modular CANDU concept of horizontal fuel channels surrounded by a heavy water moderator. However, ACR uses slightly enriched uranium fuel compared to the natural uranium used in CANDU 6. This achieves the twin goals of improved economics (via large reductions in the heavy water moderator volume and replacement of the heavy water coolant with light water coolant) and improved safety. AECL has developed and implemented a software quality assurance program to ensure that its analytical, scientific and design computer codes meet the required standards for software used in safety analyses. Since the basic design of the ACR is equivalent to that of the CANDU 6, most of the key phenomena associated with the safety analyses of ACR are common, and the CANDU industry standard toolset of safety analysis codes can be applied to the analysis of the ACR. A systematic assessment of computer code applicability addressing the unique features of the ACR design was performed covering the important aspects of the computer code structure, models, constitutive correlations, and validation database. Arising from this assessment, limited additional requirements for code modifications and extensions to the validation databases have been identified. This paper provides an outline of the AECL software quality assurance program process for the validation of computer codes used to perform physics and thermal-hydraulics safety analyses of the ACR. It describes the additional validation work that has been identified for these codes and the planned, and ongoing, experimental programs to extend the code validation as required to address specific ACR design features that have impact on the physics and thermalhydraulics codes.
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Tsujita, Yuichi, Tatsumi Arima, Kazuya Idemitsu, Yoshio Suzuki, and Hideo Kimura. "Building an Application-Specific Grid Computing Environment Using ITBL for Nuclear Material Engineering." In 16th International Conference on Nuclear Engineering. ASMEDC, 2008. http://dx.doi.org/10.1115/icone16-48223.

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Effective use of nuclear fuel is an important issue in nuclear material engineering. Pu recycle is refocused for effective use of nuclear fuel. MOX and inert matrix fuels (e.g., ZrO2-PuO2) are expected for effective burning Pu, however, Pu material is difficult to manage due to its radiotoxicity. As a result, cost of experimental facilities is very expensive. As computer simulation not only cuts cost for experimental researches but also provides deep understandings in atomic behavior, we have performed molecular dynamics (MD) simulations to research its material characteristics. In order to obtain realistic results, many atoms and many time steps are essential, however, such computation takes too much long time. So a parallelized program code is executed on a grid computing environment provided by an Information Technology Based Laboratory (ITBL) project. Its grid computing infrastructure (hereafter an ITBL system) provides users a seamless computing environment and many kinds of software tools such as a file manager, a program execution manager, and a cooperative tool for AVS/Express on a grid computing environment. Furthermore, a client application program interface (API) is provided to build a variety of grid applications on a client terminal PC for accessing the ITBL’s functionalities. As there is a strong request from users in our material simulation research to utilize their native visualization software, we have selected the client API to build an application-specific grid computing environment which cooperates with the visualization software. We have built a customized graphical user interface (GUI) computing environment on a client terminal PC by using the Java client API. It provides a seamless access to ITBL’s computational resources from a user’s terminal PC. It also assists choosing parameters for the computation in parameter survey runs. Moreover, it enables successive processing of computation on remote parallel computers and visualization on a user’s terminal PC in a single operation. The environment also provides a user friendly GUI interface for parameter controls and monitoring of submitted jobs. As a result, this computing environment removes difficulties in manual operations for parallel computations and visualizations in parameter survey runs. So, it prevents users from mistakes in the operations. This environment is expected to accelerate finding procedure for good nuclear fuel.
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Ali, Muhammad Ansab, Tariq Saeed Khan, Ebrahim Al Hajri, and Zahid H. Ayub. "A Computer Program for Working Fluid Selection of Low Temperature Organic Rankine Cycle." In ASME 2015 Power Conference collocated with the ASME 2015 9th International Conference on Energy Sustainability, the ASME 2015 13th International Conference on Fuel Cell Science, Engineering and Technology, and the ASME 2015 Nuclear Forum. American Society of Mechanical Engineers, 2015. http://dx.doi.org/10.1115/power2015-49691.

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Fossil fuels are continuously depleting while the global energy demand is growing at a fast rate. Additionally, fossil fuels based power plants contribute to environmental pollution. Search for alternate energy resources and use of industrial waste heat for power production are attractive topics of interest these days. One way of enhancing power production and decreasing the environmental impact is by recuperating and utilizing low grade thermal energy. In recent years, research on use of organic Rankine cycle (ORC) has gained popularity as a promising technology for conversion of heat into useful work or electricity. Due to simple structure of ORC system, it can be easily integrated with any energy source like geothermal energy, solar energy and waste heat. A computer program has been developed in engineering equation solver (EES) environment that analyzes and selects appropriate working fluid for organic Rankine cycle design based on available heat sources. For a given heat source, the program compares energy and exergy performance of various working fluids. The program also includes recuperator performance analysis and compares its effectiveness on the overall thermal performance of the Rankine cycle. This program can assist in preliminary design of ORC with respect to best performing refrigerant fluid selection for the given low temperature heat source.
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Mesina, George L. "Visualization of Heat Transfer and Core Damage With RGUI 1.5." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22698.

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Graphical User Interfaces (GUI) have become an integral and essential part of computer software. In the ever-changing world of computing, they provide the user with a valuable means to learn, understand, and use the application software while also helping applications adapt to and span different computing paradigms, such as different operating systems. For these reasons, GUI development for nuclear plant analysis programs has been ongoing for a decade and a half and much progress has been made. With the development of codes such as RELAP5-3D [1] and SCDAP/RELAP5 that have multi-dimensional modeling capability, it has become necessary to represent three-dimensional, calculated data. The RELAP5-3D Graphical User Interface (RGUI) [4] was designed specifically for this purpose. It reduces the difficulty of analyzing complex three-dimensional models and enhances the analysts’ ability to recognize plant behavior visually. Previous versions of RGUI [5] focused on visualizing reactor coolant behavior during a simulated transient or accident. Recent work has extended RGUI to display two other phenomena, heat transfer and core damage. Heat transfer is depicted through the visualization of RELAP5-3D heat structures. Core damage is visualized by displaying fuel rods and other core structures in a reactor vessel screen. Conditions within the core are displayed via numerical results and color maps. These new features of RGUI 1.5 are described and illustrated.
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Jeong, Hae-Yong, Kwi-Seok Ha, Kwi-Lim Lee, Young-Min Kwon, Won-Pyo Chang, Su-Dong Suk, and Yeong-Il Kim. "Pre-Test Analysis of Natural Circulation Test of PHENIX End-of-Life With the MARS-LMR Code." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29874.

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PHENIX, a prototype sodium-cooled fast reactor (SFR), has demonstrated a fast breeder reactor technology and also achieved its important role as an irradiation facility for innovative fuels and materials. In 2009 PHENIX reached its final shutdown and the CEA launched a PHENIX end-of-life (EOL) test program, which provided a unique opportunity to validate an SFR system analysis code. The Korea Atomic Energy Research Institute (KAERI) joined this program to evaluate the capability and limitation of the MARS-LMR code, which will be used as a basic tool for the design and analysis of future SFRs in Korea. For this purpose, pre-test analyses of PHENIX EOL natural circulation tests have been performed and one-dimensional thermal-hydraulic behaviors for these tests have been analyzed. The natural circulation test was initiated by the decrease of heat removal through steam generators (SGs). This resulted in the increase of intermediate heat exchanger (IHX) secondary inlet temperature, followed by a manual reactor scram and the decrease of secondary pump speed. After that, the primary flow rate was also controlled by the manual trip of three primary pumps. For the pre-test analysis the Phenix primary system and IHXs were nodalized into several volumes. Total 981 subassemblies in the core were modeled and they were divided into 7 flow channels. The active 4 IHXs were modeled independently to investigate the change of flow into each IHX. The cold pool was modeled by two axial nodes having 5 and 6 sub-volumes, respectively. The reactor vessel cooling system was modeled to match the flow balance in the primary system. The flow path of vessel cooling system was quite complicated. However, it is simplified in the modeling. For a MARS-LMR simulation, the dryout of SGs have been described by the use of the boundary conditions for IHTS as a form of time-to-temperature table. This boundary condition reflects the increase in IHTS temperature by SG dryout during the initial stage of the transient and the increase in heat removal by the opening of the two SG containments at 3 hours after the initiation of the transient. Through the comparison of the pre-analysis results with the prediction by other computer codes, it is found that the MARS-LMR code predicts natural circulation phenomena in a sodium system in a reasonable manner. The final analysis for validation of the code against the test data will be followed with an improved modeling in near future.
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Kawada, Ken-ichi, Ikken Sato, Yoshiharu Tobita, Werner Pfrang, Laurence Buffe, and Emmanuelle Dufour. "Development of PIRT (Phenomena Identification and Ranking Table) for SAS-SFR (SAS4A) Validation." In 2014 22nd International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/icone22-30679.

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SAS-SFR (derived from SAS4A) is presently the most advanced computer code for simulation of the primary phase of the Core Disruptive Accident (CDA) of MOX-fueled Sodium-cooled Fast Reactors (SFR). In the past two decades, intensive model improvement works have been conducted for SAS-SFR utilizing the experimental data from the CABRI programs. The main target of the present work is to confirm validity of these improved models through a systematic and comprehensive set of test analyses to demonstrate that the improved models has a sufficient quality assurance level for applications to reactor conditions. In order to reach these objectives, an approach of PIRT (Phenomena Identification and Ranking Table) on a set of accident scenarios has been applied. Based on the fact that there have been a significant amount of validation studies for decades, development of the code validation matrix concentrated on key issues. Different accident scenarios have been chosen for the PIRT considering typical SFR accident transients that address a large range of phenomena. As the most important and typical Core Disruptive Accident scenarios leading to generalized core melting and to be addressed with SAS-SFR in the present study, ULOF (Unprotected Loss Of Flow), UTOP (Unprotected Transient OverPower) and ULOHS (Unprotected Loss Of Heat Sink) are selected. The PIRT process applied to a given accident scenario consists in an identification of the phenomena involved during the accident, the evaluation of the importance of the phenomena regarding to the evolution and consequences, and the evaluation of the status of knowledge based on the review of available experimental results. The identified phenomena involved in ULOF are explained as follows for the primary phase. Starting from initiating events, a loss of grid power leading to flow coast down without scram is assumed. The scenario up to coolant boiling is the main point within the first part of the ULOF phenomenological chart. Those elements related to reactivity feedback, such as heat up of coolant, fuel and various structures and their deformation due to the thermal transient are picked up. Depending on the time scale before boiling starts, primary, secondary and tertiary loop heat transfer including the DHR (Decay Heat Removal) system response is concerned since it defines the core inlet coolant temperature. Core inlet coolant temperature gives direct impact on the thermal condition of the core. It also affects reactivity through thermal expansion of the grid plate. In the second part of the ULOF phenomenological chart, elements such as coolant boiling, mechanical response of the fuel pin leading to cladding failure, FCI (Fuel-Coolant Interaction) and post-failure material relocation are picked up. This part of the chart is basically common to the ULOHS. Respective identified phenomena are to be simulated in the SAS-SFR code. To validate the function of the models in the code, ten high priority CABRI experiments are selected. Validation studies on these tests are underway. With the present study, important phenomena involved in ULOF, UTOP and ULOHS were identified and an evaluation matrix for the selected CABRI experiments was developed.
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Ohshima, Hiroyuki, and Masahiko Ohtaka. "Development of Computer Program for Whole Core Thermal-Hydraulic Analysis of Fast Reactors." In 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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8

Abutayeh, Mohammad. "Adapting Steady State Solar Power Models to Include Transients." In ASME 2015 9th International Conference on Energy Sustainability collocated with the ASME 2015 Power Conference, the ASME 2015 13th International Conference on Fuel Cell Science, Engineering and Technology, and the ASME 2015 Nuclear Forum. American Society of Mechanical Engineers, 2015. http://dx.doi.org/10.1115/es2015-49021.

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Quite a few computer programs have been developed to model power plant performance. These software codes are geared towards modeling steady state operations which is usually sufficient for conventional power plants. Solar thermal power plants undergo lengthy transient start–up and shut–down operations due to the sporadic nature of solar radiation; therefore, valid modeling of their performance must address those unsteady state operations. A novel scheme has been developed to fine–tune steady state solar power generation models to accurately take account of the impact of those transient operations. The suggested new scheme is implemented by adjusting solar radiation input data and has been shown to significantly improve modeling accuracy by moving modeled results closer to matching real operating data.
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Wren, David J., Patrick Reid, and Len L. Wright. "ACR-1000™ Fuel Design Verification." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-30338.

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The ACR-1000™ design is an evolutionary advancement of the proven CANDU® reactor design that delivers enhanced economic performance, safety, operability and maintainability. The fuel for the ACR-1000 design is based on the well established CANDU fuel bundle design that has over 40 years of demonstrated high performance. Building on its extensive experience in fuel design and analysis, and fuel testing, AECL has designed a CANFLEX-ACR™ fuel bundle that incorporates the latest improvements in CANDU fuel bundle design. The ACR-1000 fuel bundle also includes features that enable the ACR-1000 to achieve higher fuel burn-up and improved reactor core physics characteristics. To verify that the CANFLEX-ACR fuel bundle design will meet and exceed all design requirements, an extensive program of design analysis and testing is being carried out. This program rigorously evaluates the ability of the fuel design to meet all design and performance criteria and particularly those related to fuel failure limits. The design analyses address all of the phenomena that affect the fuel during its residence in the reactor core. Analysis is performed using a suite of computer codes that are used to evaluate the temperatures, deformations, stresses and strains experienced by the fuel bundle during its residence in the reactor core. These analyses take into account the impact of fuel power history and core residence time. Complementing the analyses, testing is performed to demonstrate the compatibility of the fuel with the reactor heat transport system and fuel handling systems, and to demonstrate the ability of the fuel to withstand the mechanical forces that it will experience during its residence in the core. The testing program includes direct measurement of prototype fuel element and fuel bundle properties and performance limits. A number of different test facilities are used including a cold test loop and a hot test loop with a full-scale ACR-1000 fuel channel that operates at reactor coolant temperatures, pressures and flows. This paper summarizes the out-reactor test program and related analysis that provide the basis for verifying that the ACR-1000 fuel design meets its requirements.
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Zhang, Xiaoying, and Zhitao Liang. "Steady-State PWR Core-Wide Analysis Using Subchannel Analysis Method." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-30016.

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A subchannel program was designed to make a steady-state core-wide analysis for PWR core. Equations of continuity, energy, axial and transverse momentum, which are the main elements of the program, were presented in the paper. The spatial forward differentiation method was used to get the thermal hydraulic parameters, such as temperature, pressure, enthalpy of different spatial steps along the axial direction in each channel. In order to validate the program, a physical model of the reactor core, which is based on 900MW PWR nuclear plant, was established and computed in the program. In the program, the core was radially divided into 8 subchannels. Variations of the temperature, pressure and enthalpy in each coolant subchannel were computed. The exterior surfaces’ temperatures of the fuel rods in subchannels were computed as well. The computation shows that outlet pressure of the core is about 15.26 MPa; the average outlet temperature of each channel is about 325.34°. These results are within reasonable range based on comparing with published datum [1].
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Reports on the topic "Nuclear fuels – Computer programs"

1

Yuan, Y. C., S. Y. Chen, D. J. LePoire, and R. Rothman. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel. Office of Scientific and Technical Information (OSTI), February 1993. http://dx.doi.org/10.2172/10140193.

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2

Yuan, Y. C., S. Y. Chen, B. M. Biwer, and D. J. LePoire. RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel. Office of Scientific and Technical Information (OSTI), November 1995. http://dx.doi.org/10.2172/192550.

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3

Biwer, B. M., S. Y. Chen, D. J. LePoire, and Y. C. Yuan. Errata to RISKIND - a computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel. Office of Scientific and Technical Information (OSTI), June 1999. http://dx.doi.org/10.2172/12035.

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4

Opportunities for Russian Nuclear Weapons Institute developing computer-aided design programs for pharmaceutical drug discovery. Final report. Office of Scientific and Technical Information (OSTI), September 1996. http://dx.doi.org/10.2172/505318.

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