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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 2014-10-09T12:45:33Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:08:52Z (GMT). No. of bitstreams: 1 07171.pdf: 7581243 bytes, checksum: 53b2abeaefa7a689061fbc0c51a5c365 (MD5) / Dissertacao (Mestrado) / IPEN/D / 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

McMaster University DOCTOR OF PHILOSOPHY (2022) Hamilton, Ontario (Engineering Physics) TITLE: Development and Evaluation of Polaris CANDU Geometry Modelling and of TRACE_Mac/PARCS_Mac Coupling with RRS for CANDU Analysis AUTHOR: Simon Younan, M.A.Sc. (McMaster University), B.Eng. (McMaster University) SUPERVISOR: Dr. David Novog NUMBER OF PAGES: xiv, 163 / In the field of nuclear safety analysis, as computers have become more powerful, there has been a trend away from low-fidelity models using conservative assumptions, to high-fidelity best-estimate models combined with uncertainty analysis. A number of these tools have been developed in the United States, due to the popularity of light water reactors. These include the SCALE analysis suite developed by ORNL, as well as the PARCS and TRACE tools backed by the USNRC. This work explores adapting the capabilities of these tools to the analysis of CANDU reactors. The Polaris sequence, introduced in SCALE 6.2, was extended in this work to support CANDU geometries and compared to existing SCALE sequences such as TRITON. Emphasis was placed on the Embedded Self-Shielding Method (ESSM), introduced with Polaris. Both Polaris and ESSM were evaluated and found to perform adequately for CANDU geometries. The accuracy of ESSM was found to improve when the precomputed selfshielding factors were updated using a CANDU representation. The PARCS diffusion code and the TRACE system thermalhydraulics code were coupled, using the built-in coupling capability between the two codes. In addition, the Exterior Communications Interface (ECI), used for coupling with TRACE, was utilized. A Python interface to the ECI library was developed in this work and used to couple an RRS model written in Python to the coupled PARCS/TRACE model. A number of code modifications were made to accommodate the required coupling and correct code deficiencies, with the modified versions named PARCS_Mac and TRACE_Mac. The coupled codes were able to simulate multiple transients based on prior studies as well as operational events. The code updates performed in this work may be used for many future studies, particularly for uncertainty propagation through a full set of calculations, from the lattice model to a full coupled system model. / Thesis / Doctor of Philosophy (PhD) / Modern nuclear safety analysis tools offer more accurate predictions for the safety and operation of nuclear reactors, including CANDU reactors. These codes take advantage of modern computer hardware, and also a shift in philosophy from conservative analysis to best estimate plus uncertainty analysis. The goal of this thesis was to adapt a number of modern tools to support CANDU analysis and uncertainty propagation, with a particular emphasis on coupling of multiple interacting models. These tools were then demonstrated, and results analyzed. The simulations performed in this work were successful in producing results comparable to prior studies along with experimental and operational data. This included the simulation of four weeks of reactor operation including “shim mode” operation. Sensitivity and uncertainty analyses were performed over the course of the work to quantify the precision and significance of the results as well as to identify areas of interest for future research.

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Made available in DSpace on 2014-10-09T12:55:30Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T14:07:16Z (GMT). No. of bitstreams: 0 / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP