Several alternative strategies are being considered as spent nuclear fuel (SNF) management options. Transuranic nuclides (TRU) are responsible for the SNF long-term radiotoxicity beyond the first 500 years. One of the most viable approaches suggests creating new transmutation fuels containing TRUs for use in thermal and fast nuclear reactors. Irradiation of TRUs results in their transmutation and ultimate incineration by fission. The objective of this thesis is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled Very High Temperature Reactor (VHTR) systems. This effort was focused on the prismatic core configuration. The 3D core models were created for use in calculations with the SCALE 5.1 code system. As part of the research effort, basic nuclear characteristics of TRUs were taken into consideration. The potential variations of PWR spent fuel compositions were modeled with the International Atomic Energy Agency (IAEA) Nuclear Fuel Cycle Simulation System, VISTA. The VHTR configurations with varying TRU compositions were analyzed assuming a single-batch core operation. Their performance was compared to the VHTR cases with low enriched uranium (LEU). The analysis shows that TRUs can be effectively utilized in the VHTR systems. The TRU-fueled VHTRs exhibit favorable performance characteristics.

Several alternative strategies are being considered as spent nuclear fuel (SNF) management options. Transuranic nuclides (TRU) are responsible for the SNF long-term radiotoxicity beyond the first 500 years. One of the most viable approaches suggests creating new transmutation fuels containing TRUs for use in thermal and fast nuclear reactors. Irradiation of TRUs results in their transmutation and ultimate incineration by fission. The objective of this thesis is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled Very High Temperature Reactor (VHTR) systems. This effort was focused on the prismatic core configuration. The 3D core models were created for use in calculations with the SCALE 5.1 code system. As part of the research effort, basic nuclear characteristics of TRUs were taken into consideration. The potential variations of PWR spent fuel compositions were modeled with the International Atomic Energy Agency (IAEA) Nuclear Fuel Cycle Simulation System, VISTA. The VHTR configurations with varying TRU compositions were analyzed assuming a single-batch core operation. Their performance was compared to the VHTR cases with low enriched uranium (LEU). The analysis shows that TRUs can be effectively utilized in the VHTR systems. The TRU-fueled VHTRs exhibit favorable performance characteristics.

High Temperature Gas-cooled Reactors (HTGRs) can provide clean electricity,as well as process heat that can be used to produce hydrogen for transportation and other sectors. A prototypic HTGR, the Next Generation Nuclear Plant (NGNP),will be built at Idaho National Laboratory.The need for HTGR analysis tools and methods has led to the addition of gas-cooled reactor (GCR) capabilities to the light water reactor code MELCOR. MELCOR will be used by the Nuclear Regulatory Commission licensing of the NGNP and other HTGRs. In the present study, new input techniques have been developed for MELCOR HTGR analysis. These new techniques include methods for modeling radiation heat transfer between solid surfaces in an HTGR, calculating fuel and cladding geometric parameters for pebble bed and prismatic block-type HTGRs, and selecting appropriate input parameters for the reflector component in MELCOR. The above methods have been applied to input decks for a water-cooled reactor cavity cooling system (RCCS); the 400 MW Pebble Bed Modular Reactor (PBMR), the input for which is based on a code-to-code benchmark activity; and the High Temperature Test Facility (HTTF), which is currently in the design phase at Oregon State University. RCCS results show that MELCOR accurately predicts radiation heat transfer rates from the vessel but may overpredict convective heat transfer rates and RCCS coolant flow rates. PBMR results show that thermal striping from hot jets in the lower plenum during steady-state operations, and in the upper plenum during a pressurized loss of forced cooling accident, may be a major design concern. Hot jets could potentially melt control rod drive mechanisms or cause thermal stresses in plenum structures. For the HTTF, results will provide data to validate MELCOR for HTGR analyses. Validation will be accomplished by comparing results from the MELCOR representation of the HTTF to experimental results from the facility. The validation process can be automated using a modular code written in Python, which is described here.

The design of passive heat removal systems is one of the main concerns for the modular Very High Temperature Gas-Cooled Reactors (VHTR) vessel cavity. The Reactor Cavity Cooling System (RCCS) is an important heat removal system in case of accidents. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The commercial Computational Fluid Dynamics (CFD) STAR-CCM+/ V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. Two models were developed to analyze heat exchange in the RCCS. Both models incorporate a 180 degree section resembling the VHTR RCCS bench table test facility performed at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls temperature below design limits. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The models considered included the first-moment closure one equation Spalart-Allmaras model, the first-moment closure two-equation k-e and k-w models and the second-moment closure Reynolds Stress Transport (RST) model. For the near wall treatments, the low y+ and the all y+ wall treatments were considered. The two-layer model was also used to investigate the effect of near-wall treatment. The comparison of the experimental data with the simulations showed a satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. The tested turbulence models demonstrated that the Realizable k-e model with two-layer all y+ wall treatment performs better than the other k-e models for such a complicated geometry and flow conditions. Results are in satisfactory agreement with the RST simulations and experimental data available. A scaling analysis was developed to address the distortion introduced by the experimental facility and CFD model in simulating the physics inside the RCCS system with respect to the real plant configuration. The scaling analysis demonstrated that both the experimental facility and CFD model give a satisfactory reproduction of the main flow characteristics inside the RCCS cavity region, with convection and radiation heat exchange phenomena being properly scaled from the real plant to the model analyzed.

Minor actinides represent the long-term radiotoxicity of nuclear wastes. As one of their potential incineration options, partitioning and transmutation in fission reactors are seriously considered worldwide. If implemented, these technologies could also be a source of nuclear fuel materials required for sustainability of nuclear energy. The objective of this research was to evaluate performance characteristics of Very High Temperature Reactors (VHTRs) and their variations due to configuration adjustments targeting achievability of spectral variations. The development of realistic whole-core 3D VHTR models and their benchmarking against experimental data was an inherent part of the research effort. Although the performance analysis was primarily focused on prismatic core configurations, 3D pebble-bed core models were also created and analyzed. The whole-core 3D models representing the prismatic block and pebble-bed cores were created for use with the SCALE 5.0 code system. Each of the models required the Dancoff correction factor to be externally calculated. The code system DANCOFF-MCThe whole-core/system 3D models with multi-heterogeneity treatments were validated by the benchmark problems. Obtained results are in agreement with the available High Temperature Test Reactor data. Preliminary analyses of actinide-fueled VHTR configurations have indicated promising performance characteristics. Utilization of minor actinides as a fuel component would facilitate development of new fuel cycles and support sustainability of a fuel source for nuclear energy assuring future operation of Generation IV nuclear energy systems. was utilized to perform the Dancoff factor calculations.

No description available.

A new high-fidelity integrated system method and analysis approach was developed and implemented for consistent and comprehensive evaluations of advanced fuel cycles leading to minimized Transuranic (TRU) inventories. The method has been implemented in a developed code system integrating capabilities of MCNPX for highfidelity fuel cycle component simulations. The impact associated with energy generation and utilization is immeasurable due to the immense, widespread, and myriad effects it has on the world and its inhabitants. The polar extremes are demonstrated on the one hand, by the high quality of life enjoyed by individuals with access to abundant reliable energy sources, and on the other hand by the global-scale environmental degradation attributed to the affects of energy production and use. Thus, nations strive to increase their energy generation, but are faced with the challenge of doing so with a minimal impact on the environment and in a manner that is self-reliant. Consequently, a revival of interest in nuclear energy has followed with much focus placed on technologies for transmuting nuclear spent fuel. In this dissertation, a Nuclear Energy System (NES) configuration was developed to take advantage of used fuel recycling and transmutation capabilities in waste management scenarios leading to minimized TRU waste inventories, long-term activities, and radiotoxicities. The reactor systems and fuel cycle components that make up the NES were selected for their ability to perform in tandem to produce clean, safe, and dependable energy in an environmentally conscious manner. The reactor systems include the AP1000, VHTR, and HEST. The diversity in performance and spectral characteristics for each was used to enhance TRU waste elimination while efficiently utilizing uranium resources and providing an abundant energy source. The High Level Waste (HLW) stream produced by typical nuclear systems was characterized according to the radionuclides that are key contributors to long-term waste management issues. The TRU component of the waste stream becomes the main radiological concern for time periods greater than 300 years. A TRU isotopic assessment was developed and implemented to produce a priority ranking system for the TRU nuclides as related to long-term waste management and their expected characteristics under irradiation in the different reactor systems of the NES. Detailed 3D whole-core models were developed for analysis of the individual reactor systems of the NES. As an inherent part of the process, the models were validated and verified by performing experiment-to-code and/or code-to-code benchmarking procedures, which provided substantiation for obtained data and results. Reactor core physics and material depletion calculations were performed and analyzed. A computational modeling approach was developed for integrating the individual models of the NES. A general approach was utilized allowing for the Integrated System Model (ISM) to be modified in order to provide simulation for other systems with similar attributes. By utilizing this approach, the ISM is capable of performing system evaluations under many different design parameter options. Additionally, the predictive capabilities of the ISM and its computational time efficiency allow for system sensitivity/uncertainty analysis and the implementation of optimization techniques. The NES has demonstrated great potential for providing safe, clean, and secure energy and doing so with foreseen advantages over the LEU once-through fuel cycle option. The main advantages exist due to better utilization of natural resources by recycling the used nuclear fuel, and by reducing the final amount and time span for which the resulting HLW must be isolated from the public and the environment due to radiological hazard. If deployed, the NES can substantially reduce the long-term radiological hazard posed by current HLW, extend uranium resources, and approach the characteristics of an environmentally benign energy system.

Submitted by Isaac Francisco de Souza Dias (isaac.souzadias@ufpe.br) on 2016-02-16T18:00:16Z No. of bitstreams: 2 license_rdf: 1232 bytes, checksum: 66e71c371cc565284e70f40736c94386 (MD5) Tese_completa_portugues_Jesus - Final_B1.pdf: 4458139 bytes, checksum: 1ccef01f0261985c1f30824dd4c94ae6 (MD5) / Made available in DSpace on 2016-02-16T18:00:16Z (GMT). No. of bitstreams: 2 license_rdf: 1232 bytes, checksum: 66e71c371cc565284e70f40736c94386 (MD5) Tese_completa_portugues_Jesus - Final_B1.pdf: 4458139 bytes, checksum: 1ccef01f0261985c1f30824dd4c94ae6 (MD5) Previous issue date: 2015-05-18 / CAPES / A sustentabilidade da energia nuclear dependerá, entre outros fatores, da capacidade de redução dos inventários dos resíduos nucleares de vida longa. Com esse objetivo, desenvolveu-se a nova geração de reatores nucleares, com seis protótipos que se destacam por sua segurança, resistência à proliferação e a gestão dos resíduos. Dentro dessa nova geração de reatores, encontram-se os reatores de temperatura muito alta (VHTR), pela capacidade de produzir energia e a obtenção de altas temperaturas na saída do refrigerante, para seu uso em aplicações de alta temperatura como a produção de hidrogênio. Os ADS (Accelerator Driven Systems), que podem ser projetados como VHTR, são sistemas projetados para a redução dos elementos transurânicos provenientes dos LWRs (Light Water Reactors). O TADSEA (Transmutation Advanced Device for sustainable Energy Applications) é um ADS do tipo leito de bolas, projetado para atingir uma queima profunda dos elementos transurânicos, a produção colateral de energia e a obtenção de altas temperaturas para produzir hidrogênio. O presente trabalho têm como objetivo realizar melhoras no projeto conceitual do TADSEA, através da simulação geométrica detalhada do combustível, para o qual foi desenvolvida e avaliada uma metodologia para a modelagem computacional detalhada da dupla heterogeneidade do combustível em um leito de bolas, usando o código probabilista MCNPX. Foram incluídos novos elementos no projeto como a blindagem, as barras absorvedoras para garantir a segurança do sistema, e foi avaliado o desempenho na redução dos resíduos e sua radiotoxicidade associada, assim como a produção de energia. / The sustainability of nuclear energy will depend of the capability reduction of the inventories of long-lived nuclear waste. With this goal, it was developed the new generation of nuclear reactors with six prototypes, which stand out for their safety, proliferation resistance and the waste management. Within this new reactors generation, there is the very high temperature reactor (VHTR), designed to produce energy and to obtain high temperatures in the coolant, for their use in high temperature applications such as hydrogen production. The ADSs, which can be designed as VHTRs, are systems designed to reduce the mass of transuranic elements coming from the LWRs. The TADSEA is an ADS, pebble bed type, designed to achieve deep burning levels in the fuel, the power generation, and to obtain high temperatures to produce hydrogen. The aim of this study is to make the TADSEA conceptual re-design, by means of a detailed fuel geometric simulation, for which it was developed and evaluated a methodology for the detailed computational simulation of the fuel double heterogeneity in pebble-bed nuclear core, using the probabilistic code MCNPX. New design elements such as the shield and absorbers bars were included, and the performance in the reduction of nuclear waste and their associated radiotoxicity as well as the energy production were evaluated.

Submitted by Rafael Santana (rafael.silvasantana@ufpe.br) on 2018-02-01T18:17:15Z No. of bitstreams: 2 license_rdf: 811 bytes, checksum: e39d27027a6cc9cb039ad269a5db8e34 (MD5) Leorlen_TeseDout_Final.pdf: 6669399 bytes, checksum: 4f46335bd7455730a36a93a7e27b55a5 (MD5) / Made available in DSpace on 2018-02-01T18:17:15Z (GMT). No. of bitstreams: 2 license_rdf: 811 bytes, checksum: e39d27027a6cc9cb039ad269a5db8e34 (MD5) Leorlen_TeseDout_Final.pdf: 6669399 bytes, checksum: 4f46335bd7455730a36a93a7e27b55a5 (MD5) Previous issue date: 2016-10-20 / CAPES / A tecnologia VHTR (do inglês Very High Temperature Reactor, Reator de Temperatura Muito Elevada) representa o próximo estágio na evolução dos reatores HTGR (do inglês High Temperature Gas-Cooled Reactor, Reator de Alta Temperatura Refrigerado a Gás). Moderados a grafite e refrigerados a hélio, os sistemas VHTRs podem ser usados para a cogeração de calor e de eletricidade com temperaturas de saída entre 700 e 950 ºC, e potencialmente com mais de 1.000 ºC no futuro. A temperatura do combustível durante toda a operação do reator é um aspecto muito importante para a segurança dos reatores nucleares, no projeto deseja-se que seja menor que um valor limite para garantir a integridade dos materiais do elemento combustível evitando a liberação de produtos de fissão. O TADSEA (Transmutation Advanced Device for Sustainable Energy Applications) é um VHTR do tipo leito de bolas, projetado para atingir uma queima profunda dos elementos transurânicos, a produção colateral de energia e a obtenção de altas temperaturas para produzir hidrogênio. O presente trabalho tem como objetivo o desenvolvimento de uma metodologia para a análises termoidráulica do núcleo de reatores do tipo leito de bolas de muito alta temperatura, baseada no uso de uma abordagem realística com um código de Dinâmica dos Fluidos Computacional (CFD). Inicialmente, usando o modelo realístico da coluna com altura inteira do reator HTR-10 com células FCC e BCC, foram comparados os resultados obtidos com dados experimentais e de simulação para a primeira tarefa de referência do HTR-10 disponibilizados pela IAEA (2013) para validação do modelo. No reator TADSEA, foram comparados resultados dos projetos inicial e atual do núcleo com uma coluna com a altura completa do reator na região de maior potência. A partir dos resultados o projeto inicial não tem margem de segurança suficiente para casos de perda de refrigerante. Nas simulações do projeto atual do TADSEA as temperaturas máximas atingidas foram muito inferiores ao limite. E os resultados de casos de perda de refrigerante mostram que com 45% do fluxo mássico é atingida uma temperatura apenas 30 K abaixo do limite. / The VHTR (Very High Temperature Reactor) technology represents the next stage in the evolution of reactors HTGR (High Temperature Gas-Cooled Reactor). Moderated by graphite and cooled by helium, VHTRs systems can be used for cogeneration of heat and electricity with outlet temperatures from 700 to 950 ºC, and potentially more than 1.000 ºC in the future. The fuel temperature during all the reactor operation is a very important issue for the safety of nuclear reactors, in the design is desired that it is smaller than a limit value to ensure the integrity in the materials of the fuel element preventing the release of fission products. The TADSEA (Transmutation Advanced Device for Sustainable Energy Applications) is a VHTR pebble bed type. It is designed to achieve deep burning levels in the fuel, the power generation, and to obtain high temperatures to produce hydrogen. The aim of this study is the development of a methodology for the core termohydraulics analysis of pebble bed reactors with very high temperature based in the use of a realistic approach with a code of Computational Fluid Dynamics (CFD). First, using the realistic approach with an entire column height of HTR-10 reactor using FCC and BCC cells, the results obtained were compared with experimental and simulation data of HTR-10 Benchmark available from IAEA (2013) for model validation. In TADSEA reactor were compared the results of initial and current core designs with a column with the full height of the reactor in the higher power region. From the results of the initial design, it does not have sufficient safety margin in case of coolant loss. In the simulations of the current TADSEA design the maximum temperatures were much lower than the limit. And the results of coolant loss cases show that with 45% of the mass flow is achieved temperatures just 30 K below the limit.

Submitted by Fernanda Rodrigues de Lima (fernanda.rlima@ufpe.br) on 2018-07-27T22:51:25Z No. of bitstreams: 2 license_rdf: 811 bytes, checksum: e39d27027a6cc9cb039ad269a5db8e34 (MD5) DISSERTAÇÃO Pedro Paulo Dantas de Souza Paiva.pdf: 5292653 bytes, checksum: 177d836a13787596c993b64d01ede96a (MD5) / Approved for entry into archive by Alice Araujo (alice.caraujo@ufpe.br) on 2018-08-08T17:09:39Z (GMT) No. of bitstreams: 2 license_rdf: 811 bytes, checksum: e39d27027a6cc9cb039ad269a5db8e34 (MD5) DISSERTAÇÃO Pedro Paulo Dantas de Souza Paiva.pdf: 5292653 bytes, checksum: 177d836a13787596c993b64d01ede96a (MD5) / Made available in DSpace on 2018-08-08T17:09:39Z (GMT). No. of bitstreams: 2 license_rdf: 811 bytes, checksum: e39d27027a6cc9cb039ad269a5db8e34 (MD5) DISSERTAÇÃO Pedro Paulo Dantas de Souza Paiva.pdf: 5292653 bytes, checksum: 177d836a13787596c993b64d01ede96a (MD5) Previous issue date: 2017-05-26 / O VHTR é um reator nuclear térmico, moderado a grafite e refrigerado por hélio. Para seu desenvolvimento, há a necessidade de utilização de ferramentas computacionais eficientes para a análise de aspectos de modelagem, operação e segurança. A proposta deste trabalho é estudar o comportamento do VHTR por meio de análise paramétrica, alterando-se modelo de turbulência, perfil de geração de energia nos blocos combustíveis e a influência de modificações na própria geometria. Busca-se também avaliar a implementação de uma metodologia simplificada que reduza o esforço computacional e a duração de uma simulação. Procedeu-se à análise do escoamento do fluido refrigerante através dos canais refrigerantes e canais by-pass em uma seção de 1/12 de uma coluna de blocos combustíveis, utilizando-se diferentes modelos de turbulência. Os resultados obtidos com essas simulações foram comparados àqueles obtidos por meio de correlações do número de Nusselt descritos na literatura. Observou-se que a simulação na qual se utiliza o modelo -ε possibilita a obtenção de resultados que convergem bem com aqueles fornecidos pelas correlações, para ambos os tipos de canais. O modelo -ω proporciona bons resultados para os canais refrigerantes e, o SSG, para o canal by-pass. Utilizou-se geometria contendo canais by-pass de diferentes dimensões, além de uma que possuía apenas os canais refrigerantes, sem canal by-pass. Verificou-se que a existência de um escoamento by-pass induz a um aumento no gradiente de temperatura no bloco combustível. Realizaram-se estudos comparativos entre os resultados obtidos em simulações realizadas com diferentes perfis de geração de energia térmica (uniforme e senoidal) nos canais combustíveis. Verificou-se que, quando há a mesma geração de energia térmica total no bloco combustível, a máxima temperatura constatada em cada um dos materiais é menor para o caso da geração de energia com perfil senoidal. Quando utilizado, no perfil senoidal, um fator radial de pico (1,25), há um aumento considerável na temperatura de todos os materiais, possibilitando a ocorrência de regiões em que a temperatura pode ultrapassar o limite usualmente aceito para o combustível do reator (1250°C) em operação normal. O canal refrigerante localizado no centro do bloco combustível tem diâmetro inferior aos demais canais existentes nesse bloco. Para verificar a hipótese de que a existência de um gradiente de temperatura no bloco combustível, com a temperatura mais elevada ao centro e a temperatura mais baixa estando na periferia desse bloco, deve-se fortemente à menor dimensão desse canal central, realizaram-se simulações computacionais utilizando-se uma geometria com canal central de diâmetro igual ao dos demais. A condição de entrada escolhida para essa nova estrutura foi, primeiramente, o mesmo fluxo mássico total e, depois, a mesma diferença de pressão entre entrada e saída verificados na simulação da geometria padrão. Os resultados obtidos confirmam a hipótese aventada. Realizou-se simulação utilizando uma metodologia simplificada, que consiste em uma análise unidimensional do fluido em um canal refrigerante acoplada à análise tridimensional da condução do calor no grafite e nos canais combustíveis. Os resultados obtidos com a metodologia simplificada apresentaram excelente convergência com aqueles obtidos com a simulação completamente tridimensional, e em um tempo de simulação cerca de 45 vezes menor. / The VHTR is a thermal, graphite moderated and helium cooled reactor. For its development, there is a need to use efficient computational tools to analyze aspects of modeling, operation and safety. The purpose of this work is to study the behavior of the VHTR by means of parametric analysis, altering the turbulence model, the energy generation profile in the fuel blocks and the influence of modifications in the geometry itself. It also seeks to evaluate the implementation of a simplified methodology that reduces the computational effort and duration of a simulation. The coolant flow through the coolant channels and by-pass channels were analyzed in a 1/12 section of a fuel block column using different turbulence models. The results obtained with these simulations were compared to those obtained by means of Nusselt number correlations described in the literature. It was observed that the simulation using the -ε model makes it possible to obtain results that converge well with those provided by the correlations, for both types of channels. The -ω model provides good results for the coolant channels and the SSG for the by-pass channel. Geometry was used with by-pass channels of different dimensions, besides one that had only the cooling channels, without by-pass channel. It has been found that the existence of a by-pass flow induces an increase in the temperature gradient in the fuel block. Comparative studies were performed between the results obtained in simulations carried out with different profiles of thermal energy generation (uniform and sinusoidal) in the fuel channels. It was verified that when there is the same total thermal energy generation in the fuel block, the maximum temperature observed in each of the materials is smaller for the generation of energy with sinusoidal profile. When a peak radial factor (1,25) is used in the sinusoidal profile, there is a considerable increase in the temperature of all materials, allowing the occurrence of regions where the temperature may exceed the limit usually accepted for reactor fuel (1250° C) in normal operation. The coolant channel located in the center of the fuel block has a smaller diameter than the other channels in this block. To verify the hypothesis that the existence of a temperature gradient in the fuel block, with the highest temperature at the center and the lowest temperature being at the periphery of this block, is due to the smaller dimension of this central channel, computer simulations were performed using a geometry with a central channel with the same diameter as the others. The input condition chosen for this new structure was, firstly, the same total mass flow and then the same pressure difference between input and output found in the simulation of the standard geometry. The results confirm the hypothesis. Simulation was performed using a simplified methodology, which consists of a one-dimensional analysis of the coolant flow in a coolant channel coupled to the three-dimensional analysis of heat conduction within the graphite and fuel channels. The results obtained with the simplified methodology were in excellent agreement with those obtained with the fully three-dimensional simulation and the time required to complete the simulation with the simplified methodology was about 45 times less than the time that the fully three-dimensional simulation lasted.