During operation of nuclear power reactors reactivity initiated accidents (RIA) can occur, such as a control rod drop. If this occurs, the reactivity increase dramatically and leads to an increase in power, fuel enthalpy and fuel temperature. The fuel and reactor can be damaged. A methodology to simulate these accidents has been developed for Forsmark Nuclear Power Plant in cooperation with Westinghouse, referred to as the POLCA7 methodology. The POLCA7 methodology results in a limit for fuel failure regarding reactivity of the control rod that dropped in pcm/control rod percent. The limit is estimated from simulations in POLCA7, a static and deterministic code and POLCA-T, a dynamic code. The aim of this thesis is to evaluate the methodology and investigate what happens in a reactor if a control rod drops during hot zero power. Hot zero power is a phase during start-up, where the power is low (~2% of installed power) and the reactor have operation pressure and temperature. The POLCA7 methodology was applied on historic cycles in Forsmark. To evaluate the POLCA7 methodology the control rod drop was simulated in S3K, a dynamic software. The results from these cycles indicate that the limit for fuel failure set in the POLCA7 methodology in pcm/control rod percent is very conservative for fuel with low and medium burnup. Even though the limit is exceeded, the dynamic simulation in S3K shows that the fuel is far from failure regarding SSM limits in fuel enthalpy and cladding temperature. In this thesis new limits in POLCA7 has been generated, which is remarkably higher than the original limit from the POLCA7 methodology. To challenge the methodology, an unrealistic fuel design was simulated with fuel with high burnup surrounded by high reactive fuel. With this fuel design, the enthalpy limit from SSM was exceeded for the fuel with high burnup. The limit from the POLCA7 methodology was exceeded which indicate that the POLCA7 methodology meets the goal of detecting severe RIAs. Fuel with high burnup seems to be the most important fuel to investigate at a RIA simulation. Another discovery is that POLCA7 gives the most severe accident at 2% power, but in S3K it is given by 3-4% power. This is a problem with the POLCA7 methodology. Suggestions are made on how to lower the calculation time and improve the methodology. A control rod sequence that gives an even power distribution and a core with the fuel with high burnup in the periphery and only a few fresh fuels is preferred to avoid damage at a RIA. A control rod sequence was designed for the new cycle in Forsmark 1, in order to try to create a cycle without problems due to RIA. The new sequence was a success with no control rods exceeding the limit of 82 pcm/control rod percent, and it shows that conclusions about the impact of the sequence are correct. Conclusion is made that the methodology should be further investigated and there are good chances to develop a good and time efficient analysis in the future. One presented suggestion is to have a dynamic simulation of the incident instead of the axial simulation. The evaluation with SSM’s limits would then be direct.

This report accounts for investigations of ways to reduce the calculation times forsimulations of falling control rods in boiling water reactors done prior to everyreactor startup, known as RIA-calculations. Two methodologies to lower thecalculation times have been proposed, developed and implemented in a set ofmatlab-scripts, which are fully compatible with the previously used methodology.The new methodologies have been applied on 17 authentic power cycles at the threeForsmark reactors, whereby a reduction in calculation times by 70 to 90 % could bedemonstrated while still confidently maintaining the analysis performance. Thesimulations made and the basis of the new methodologies are described in detail inthis report, and possible steps to further lower the calculation times are alsoproposed.

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This study aims to improve the modeling of partial SCRAM in the BISON plant model for the Forsmark 2 nuclear reactor after power uprate. Validation of the BISON model against tests performed from March to May in 2013 have shown that this is one of the areas in which there is room for improvement. After partial SCRAM is performed, the model underestimates the reactor power, recirculation flow and steam flow when compared to the measurement data. In BISON the partial SCRAM is modeled using a relative control rod effect vector (ASC vector). The aim is to replace the old values in this vector to improve the model. The new model was shown to give an improved result for the reactor power, recirculation flow and steam flow. The study gives recommendations on how to apply the new model and what values of the relative control rod effect vector that can be used in the future.

This report investigates the sensitivity of the control rod depletion coefficients, Sg, to different input parameters and how this affects the accumulated 10B depletion, β. Currently the coefficients are generated with PHOENIX4, but the geometries can be more accurately simulated in McScram. McScram is used to calculate Control Rod Worth, which in turn is used to calculate Nuclear End Of Life, and Sg cannot be generated in the current version of McScram. Therefore, it is also analyzed whether the coefficients can be related to CRW and thus be studied indirectly through it. Simulations of the coefficients were done in PHOENIX4, simulations of CRW were done in both PHOENIX4 and McScram and simulations of β were done in POLCA7. All simulations were performed for a CR99 in a BWR reactor. The control rod coefficients were found to be sensitive to the enrichment of the fuel, void fraction of the water and the width of the gap, and these effects were also seen in the results of β. As a result, one of three steps could be taken. First, the parameter values should not be set arbitrarily, instead default values could be chosen such that Sg is calculated more accurately. Second, a set of tables of Sg could be generated for different parameter values so that β can be calculated with Sg from the current conditions, although this would mean that PHOENIX4 needs to be updated. Third, McScram can be updated to be able to calculate Sg directly. It has been concluded that Sg cannot be studied indirectly through CRW since the trends and the sensitivity to the different parameters were not consistent between Sg, CRW calculated with PHOENIX4 and CRW calculated with McScram, where PHOENIX4 was more sensitive than McScram. The results can instead be used to bench-mark the PHOENIX4 results.

In a nuclear power plant, control rods play a central role to control the reactivity ofthe core. In an inspection campaign of three control rods (CR 99) operated in theKKL reactor in Leibstadt, Switzerland, during 6 respectively 7 consecutive cycles,defects were detected in the top part of the control rods due to swelling caused bydepletion of the neutron-absorbing 10B isotope (Boron-10). In order to correlatethese defects to control rod depletion, the 10B depletion has in this study beencalculated in detail for the absorber pins in the top node of the control rods.Today the core simulator PLOCA7 is used for predicting the behavior of the reactorcore, where the retrievable information from the standard control rod follow-up isthe average 10B depletion for clusters of 19 absorber holes i.e. one axial node.However, the local 10B depletion in an absorber pin may be significantly differentfrom the node average depletion that is re-ceived from POLCA7. To learn more, the 10B depletion has been simulated for each absorber hole in the uppermost node usingthe stochastic Monte Carlo 3D simulation code MCNP as well as an MCNP- based2D-depletion code (McScram). It was found that the 10B depletion is significantly higher for the uppermost absorberpins than the node average. Furthermore, the radial depletion in individual absorberpins was found to be much higher than expected. The results are consistent with theexperimental data on control rod defects.

Made available in DSpace on 2014-10-09T12:51:41Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:10:25Z (GMT). No. of bitstreams: 0 / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Made available in DSpace on 2014-10-09T12:51:41Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:10:25Z (GMT). No. of bitstreams: 0 / Um dos principais mecanismos de falha que causam riscos de fratura a reatores de água pressurizada é a corrosão sob tensão de ligas metálicas em água do circuito primário (CSTAP). É causada por uma combinação das tensões de tração, meio ambiente em temperatura e microestruturas metalúrgicas susceptíveis. Ela pode ocorrer, dentre outros locais, nos bocais do mecanismo de acionamento das barras de controle. Essa fratura pode causar acidentes que comprometem a segurança nuclear através do bloqueio das barras de controle e vazamentos de água do circuito primário reduzindo a confiabilidade e a vida útil do reator. O objetivo desta Tese de Doutorado é o estudo de modelos e uma proposta de modelagem para fraturas por corrosão sob tensão em liga 75Ni15Cr9Fe (liga 600), em água de circuito primário de reator de água pressurizada nesses bocais. São superpostos modelos eletroquímicos e de mecânica da fratura e validados com dados obtidos em experimentos e na literatura. Na parte experimental foram utilizados resultados obtidos pelo CDTN no equipamento recém-instalado de ensaio por taxa de deformação lenta. Na literatura está proposto um diagrama que exprime a condição termodinâmica de ocorrerem diversos modos de CSTAP na liga 600: partiu-se de diagramas de potencial x pH (diagramas de Pourbaix), para a liga 600 imersa em água primária à alta temperatura (3000C a 3500C). Sobre ele, determinaram-se os submodos de corrosão, a partir de dados experimentais. Em seguida acrescentou-se uma dimensão adicional ao diagrama, correlacionando uma variável a que se denominou fração de resistência à corrosão sob tensão. No entanto, é possível acrescentar-se outras variáveis que exprimem a cinética de iniciação e/ou crescimento de trinca, provenientes de outras modelagens de CSTAP. A contribuição original deste trabalho se insere nessa fase: partindo-se de uma condição de ensaio de potencial versus pH, foram iniciadas as modelagens de um modelo empírico-comparativo, um semi-empírico-probabilístico, um de tempo de iniciação e um de taxa de deformação, a partir dos ensaios experimentais e superpostas a essa condição. Esses exprimem respectivamente a susceptibilidade à CSTAP, o tempo de falha, e nos dois últimos o tempo de iniciação de falha por corrosão sob tensão. Os resultados foram comparados com os da literatura e se mostraram coerentes. Através desse trabalho, obteve-se uma metodologia de modelagem a partir de dados experimentais. / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Submitted by Pedro Silva Filho (pfsilva@ipen.br) on 2017-11-23T12:33:35Z No. of bitstreams: 0 / Made available in DSpace on 2017-11-23T12:33:35Z (GMT). No. of bitstreams: 0 / O presente trabalho destina-se à apresentar uma contribuição contemporânea no que tange ao debate existente acerca da efetivação do direito à saúde no Brasil face a inoperância do Poder Público no exercício de suas funções. Metodologicamente a pesquisa foi classificada quanto aos fins em exploratória e qualitativa, e quanto aos meios em bibliográfica e documental. Com o intuito de aprimorar a problemática do tema de maneira inovadora, a princípio, foi feita uma abordagem sobre à evolução histórica dos Direitos Fundamentais e Humanos, sendo identificado que o direito à saúde é caracterizado como um direito fundamental social de segunda dimensão. Após, foram apresentadas considerações no contexto do Direito Comparado, referentes à conquista e edificação dos Direitos Fundamentais, merecendo respaldo às contribuições inglesa, americana e francesa. Em sequência, foi trabalhada a ligação existente entre o direito à saúde, a dignidade humana, o mínimo existencial e a teoria da reserva do possível, teoria esta, criticada, por ter vinculação direta quanto aos limites orçamentários públicos dentro dos quais o direito à saúde é exigível. Em continuidade, foi analisada a estrutura normativa que prevê a criação e competência do SUS (Sistema Único de Saúde-SUS), haja vista que a sua ineficácia tem implicação direta no aumento das demandas judicias relativas à ineficácia do direito à saúde. Direcionado à apresentar um benefício inédito no que tange à efetivação da saúde pública pátria, foi desenvolvido um trabalho multidisciplinar, apresentando como relevante objetivo a possibilidade de incidência da Justiça Restaurativa no cenário atual da judicialização do direito à saúde. Tendo como premissa que o direito à saúde é uma garantia constitucional e fundamental, conexa à vida digna, buscou-se assimilar da melhor forma o fenômeno da judicialização do direito à saúde, tendo como ponto de partida seu conceito clássico, em ser um mecanismo onde através da propositura de ações judiciais, uma pessoa ou um grupo de pessoas tutelam ao Poder Judiciário prestações positivas voltadas ao oferecimento de bens e serviços relativos à saúde. A originalidade desta tese consistiu em quebrar paradigmas pré constituídos, no sentido de haver um prático reconhecimento de uma judicialização contemporânea do direito à saúde, com o reconhecimento por parte do Poder Judiciário, seu operador, em ser pluralista, admitindo em seu cotidiano, discussões que venham à contribuir à efetivação do direito à saúde, não sendo necessariamente por meio de uma demanda processual e de cunho judicial. Assim, eis aqui a relevância evolutiva de haver no Brasil uma política judicial moderna, aberta à participação multifacetada e reestruturaste no tocante à efetivação do direito à saúde. / Tese (Doutorado em Tecnologia Nuclear) / IPEN/T / Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP