Fundamentado na mecânica da fratura, o conceito do Leak-Before-Break (LBB) \"Vazamento Antes da Falha\" considera que um vazamento proveniente de uma trinca pode ser detectado antes de alcançar um tamanho crítico que implique na falha da tubulação, ou seja, a análise do LBB demostra através de uma justificativa técnica que a probabilidade de ruptura da tubulação é extremamente baixa. Dentre os aspectos que envolvem a aplicação do LBB, os principais são: a definição das propriedades do material, que são extraídos através de ensaios à tração e à fratura; a análise do vazamento, que determina a taxa de vazamento devido à presença de uma trinca passante; e a análise que verifica se a trinca é estável considerando os modos de falha por rasgamento dúctil e por colapso plástico. Os materiais SA-508 Cl. 3, SA-106 Gr. B e SA-376-TP304 foram avaliados quanto aos seus desempenhos para o LBB. Utilizaram-se dados extraídos de casos da literatura para as propriedades dos materiais, e para a geometria e carregamentos da tubulação, todos correspondentes ao circuito primário de um reator PWR. Após aplicação do LBB, constatou-se que todos os três materiais atenderam os limites do estabelecidos na metodologia. Verificou-se que os materiais SA-508 Cl. 3 e SA-376-TP304 mostraram o melhor desempenho para falha por rasgamento dúctil e falha por colapso plástico, respectivamente, e o material SA-106 Gr. B teve o menor desempenho em ambos. Todos os três materiais apresentaram o colapso plástico como modo de falha mais provável. De uma forma generalizada, o material SA-376-TP304 obteve o melhor desempenho para o LBB dentre os três materiais avaliados neste trabalho. / Based on the fracture mechanics, the Leak-Before-Break (LBB) concept considers that a leakage from a crack can be detected before reaching a critical size that implies the pipe failure, that is, the LBB analysis demonstrates through a technical justification that the probability of pipe rupture is extremely low. Among the aspects that involve the application of LBB, the main ones are: the definition of the material properties, which are obtained through tensile and fracture tests; the leakage analysis, which determines the rate of leakage due to the presence of a through-wall crack; and the analysis that verifies if the crack is stable considering the failure modes by ductile tearing and plastic collapse. The materials SA-508 Cl. 3, SA-106 Gr. B and SA-376-TP304 were evaluated in relation to their performances for LBB. Data obtained from literature cases were used for the materials properties, and for the geometry and loadings of the pipe, all corresponding to the primary circuit of a PWR reactor. After application of the LBB, it was verified that all three materials met the limits established in the methodology. The materials SA-508 Cl. 3 and SA-376-TP304 showed the best performance for ductile tearing failure and plastic collapse failure, respectively, and the material SA-106 Gr. B material had the lowest performance in both. All three materials presented plastic collapse as the most likely failure mode. In general, the material SA-376-TP304 obtained the best performance for the LBB among the three materials evaluated in this work.

Este trabalho apresenta um estudo prático da aplicação do conceito Leak Before Break (LBB), usualmente aplicado em usinas nucleares, em uma tubulação fabricada a partir de aço AISI 316LN soldada com a utilização de eletrodo revestido AISI 316L. O LBB é um critério fundamentado em análises de mecânica da fratura, que considera que um vazamento proveniente de uma trinca, presente em uma tubulação, possa ser detectado por sistemas de detecção de vazamento, antes que esta trinca alcance um tamanho crítico que implique na falha da tubulação. Na tubulação estudada, foram realizados ensaios mecânicos de tração e análises de Ramberg-Osgood, bem como ensaios de tenacidade à fratura para a obtenção da curva de resistência J-R do material. Os ensaios foram realizados considerando o metal base, a solda e a zona termicamente afetada (ZTA), nas temperaturas de operação de uma planta nuclear. Para as propriedades mecânicas encontradas nos ensaios foram realizadas análises de carga limite para se determinar o tamanho da trinca que cause um vazamento detectável e, também, o seu tamanho crítico que cause a falha por colapso plástico. Para o tamanho crítico de trinca encontrado na solda, região que apresentou a menor tenacidade, foram realizadas análises de Integral J e de módulo de rasgamento T, considerando falha por rasgamento dúctil. Os resultados demonstram um comportamento bem definido entre o metal base, a ZTA e a solda, onde o metal base apresenta um comportamento altamente tenaz, a solda um comportamento pouco tenaz e a ZTA apresentou propriedades mecânicas intermediárias entre o metal base e a solda. Utilizando o software PICEP, foram determinadas as curvas de taxa de vazamento versus tamanho de trinca e também o tamanho crítico da trinca, considerando análise por carga limite. Observou-se que, após certo tamanho de trinca, a taxa de vazamento do metal base é muito maior do que para a ZTA e solda, para um mesmo comprimento de trinca. Isso ocorre porque é esperado que a trinca cresça de forma mais arredondada no metal base, devido à sua maior tenacidade. O menor tamanho crítico de trinca foi encontrado para o metal base para trincas circunferenciais. Para as análises de Integral J realizadas na solda, foi demonstrado que a falha por rasgamento dúctil não ocorrerá nas condições consideradas e essa hipótese foi sedimentada pela análise de mecânica da fratura elasto-plástica (MFEL) com o uso do diagrama J/T. Dessa forma, pode-se concluir que a tubulação estudada estaria apta a ser empregada em um circuito primário de um reator que utilizasse o critério de LBB, nas condições de carregamento e geometria consideradas. Adicionalmente, concluiu-se que nessas condições apenas o modo de falha por colapso plástico é esperado. / This work presents a study of application of the Leak Before Break (LBB) concept, usually applied in nuclear power plants, in a pipe made from steel AISI type 316LN welded a coated electrode AISI type 316L. LBB concept is a criterion based on fracture mechanics analysis to show that a crack leak, present in a pipe, can be detected by leak detection systems, before this crack reaches a critical size that results in pipe fail. In the studied pipe, tensile tests and Ramberg-Osgood analyses were performed, as well as fracture toughness tests for obtaining the material resistance curve J-R. The tests were performed considering the base metal, weld and heat affected zone (HAZ), at the same operating temperatures of a nuclear power plant. For the mechanical properties found in these tests, load limit analyses were performed in order to determine the size of a crack which could cause a detectable leakage and the critical crack size, considering failure by plastic collapse. For the critical crack size found in the weld, which is the region that presented the lowest toughness, Integral J and tearing modulus T analyses were performed, considering failure by tearing instability. Results show a well-defined behavior between the base metal, HAZ and weld zones, where the base metal has a high toughness behavior, the weld has a low toughness behavior and the HAZ showed intermediate mechanical properties between the base metal and the weld. Using the PICEP software, the leak rate curves versus crack size and also the critical crack size were determined by considering load limit analysis. It was observed that after a certain crack size, the leak rate in base metal is much higher than for the HAZ and the weld, considering the same crack length. This occurs because in the base metal crack, it is expected that the crack grows in a more rounded form due to its higher toughness. The lowest critical crack size was found for the base metal presenting circumferential cracks. For the Integral J analyses performed in the weld, it was demonstrated that the failure by tearing instability will not occur under the considered conditions and this hypothesis was confirmed by elastic-plastic fracture mechanic (EPFM) analysis using the J/T diagram. Thereby, it can be concluded that it would be possible to apply the investigated pipe in a primary circuit of a reactor that utilizes the criterion LBB under the considered geometry and load conditions. Additionally, it was found that under these conditions, only the failure mode by plastic collapse is expected.

Este trabalho apresenta um estudo prático da aplicação do conceito Leak Before Break (LBB), usualmente aplicado em usinas nucleares, em uma tubulação fabricada a partir de aço AISI 316LN soldada com a utilização de eletrodo revestido AISI 316L. O LBB é um critério fundamentado em análises de mecânica da fratura, que considera que um vazamento proveniente de uma trinca, presente em uma tubulação, possa ser detectado por sistemas de detecção de vazamento, antes que esta trinca alcance um tamanho crítico que implique na falha da tubulação. Na tubulação estudada, foram realizados ensaios mecânicos de tração e análises de Ramberg-Osgood, bem como ensaios de tenacidade à fratura para a obtenção da curva de resistência J-R do material. Os ensaios foram realizados considerando o metal base, a solda e a zona termicamente afetada (ZTA), nas temperaturas de operação de uma planta nuclear. Para as propriedades mecânicas encontradas nos ensaios foram realizadas análises de carga limite para se determinar o tamanho da trinca que cause um vazamento detectável e, também, o seu tamanho crítico que cause a falha por colapso plástico. Para o tamanho crítico de trinca encontrado na solda, região que apresentou a menor tenacidade, foram realizadas análises de Integral J e de módulo de rasgamento T, considerando falha por rasgamento dúctil. Os resultados demonstram um comportamento bem definido entre o metal base, a ZTA e a solda, onde o metal base apresenta um comportamento altamente tenaz, a solda um comportamento pouco tenaz e a ZTA apresentou propriedades mecânicas intermediárias entre o metal base e a solda. Utilizando o software PICEP, foram determinadas as curvas de taxa de vazamento versus tamanho de trinca e também o tamanho crítico da trinca, considerando análise por carga limite. Observou-se que, após certo tamanho de trinca, a taxa de vazamento do metal base é muito maior do que para a ZTA e solda, para um mesmo comprimento de trinca. Isso ocorre porque é esperado que a trinca cresça de forma mais arredondada no metal base, devido à sua maior tenacidade. O menor tamanho crítico de trinca foi encontrado para o metal base para trincas circunferenciais. Para as análises de Integral J realizadas na solda, foi demonstrado que a falha por rasgamento dúctil não ocorrerá nas condições consideradas e essa hipótese foi sedimentada pela análise de mecânica da fratura elasto-plástica (MFEL) com o uso do diagrama J/T. Dessa forma, pode-se concluir que a tubulação estudada estaria apta a ser empregada em um circuito primário de um reator que utilizasse o critério de LBB, nas condições de carregamento e geometria consideradas. Adicionalmente, concluiu-se que nessas condições apenas o modo de falha por colapso plástico é esperado. / This work presents a study of application of the Leak Before Break (LBB) concept, usually applied in nuclear power plants, in a pipe made from steel AISI type 316LN welded a coated electrode AISI type 316L. LBB concept is a criterion based on fracture mechanics analysis to show that a crack leak, present in a pipe, can be detected by leak detection systems, before this crack reaches a critical size that results in pipe fail. In the studied pipe, tensile tests and Ramberg-Osgood analyses were performed, as well as fracture toughness tests for obtaining the material resistance curve J-R. The tests were performed considering the base metal, weld and heat affected zone (HAZ), at the same operating temperatures of a nuclear power plant. For the mechanical properties found in these tests, load limit analyses were performed in order to determine the size of a crack which could cause a detectable leakage and the critical crack size, considering failure by plastic collapse. For the critical crack size found in the weld, which is the region that presented the lowest toughness, Integral J and tearing modulus T analyses were performed, considering failure by tearing instability. Results show a well-defined behavior between the base metal, HAZ and weld zones, where the base metal has a high toughness behavior, the weld has a low toughness behavior and the HAZ showed intermediate mechanical properties between the base metal and the weld. Using the PICEP software, the leak rate curves versus crack size and also the critical crack size were determined by considering load limit analysis. It was observed that after a certain crack size, the leak rate in base metal is much higher than for the HAZ and the weld, considering the same crack length. This occurs because in the base metal crack, it is expected that the crack grows in a more rounded form due to its higher toughness. The lowest critical crack size was found for the base metal presenting circumferential cracks. For the Integral J analyses performed in the weld, it was demonstrated that the failure by tearing instability will not occur under the considered conditions and this hypothesis was confirmed by elastic-plastic fracture mechanic (EPFM) analysis using the J/T diagram. Thereby, it can be concluded that it would be possible to apply the investigated pipe in a primary circuit of a reactor that utilizes the criterion LBB under the considered geometry and load conditions. Additionally, it was found that under these conditions, only the failure mode by plastic collapse is expected.

An investigation into the thermo-mechanical closure effect when a fluid leaks through a crack is presented here. The extended finite element method is the modelling scheme adopted for this, and the application of heat flux and pressure jump conditions along the crack is one of the novel contributions of this work. By modelling the fluid as one dimensional steady state and obtaining a heat transfer coefficient, it has been shown here that coupling the fluid with the structure is possible all within a single element. Convergence studies done with analytical models as a benchmark demonstrate the accuracy of the new method. Simulations are performed with the new element for conditions seen in both gas cooled and water cooled reactors. Significant crack closure is observed when the bulk fluid temperature is 20oC hotter than the structure. It was also found that the amount of closure due to crack wall heating varies depending on the external boundary conditions, this is quantified in the thesis.

Probabilistic assessment and life cycle management of engineering components and systems in a nuclear power plant is intended to ensure safe and efficient operation of energy generation over its entire life. The CANDU reactor core consists of 380-480 pressure tubes, which are like miniature pressure vessels that contain natural uranium fuel. Pressure tubes operate under severe temperature and radiation conditions, which result in degradation with ageing. Presence of flaws in a pressure tube makes it vulnerable to delayed hydride cracking (DHC), which may lead to rupture or break-before-leak situation. Therefore, assessment of flaws in the pressure tubes is considered an integral part of a reactor core assessment program. The main objective of the thesis is to develop advanced probabilistic and mechanical stress field models for the assessment of flaws. The flaw assessment models used by the industries are based on deterministic upper/lower bound values for the variables and they ignore uncertainties associated with system parameters. In this thesis, explicit limit state equations are formulated and first order reliability method is employed for reliability computation, which is more efficient than simulation-based methods. A semi-probabilistic approach is adopted to develop an assessment model, which consists of a mechanics-based condition (or equation) involving partial factors that are calibrated to a specified reliability level. This approach is applied to develop models for DHC initiation and leak-before-break assessments. A novel feature of the proposed method is that it bridges the gap between a simple deterministic analysis and complex simulations, and it is amenable to practical applications. The nuclear power plant systems are not easily accessible for inspection and data collection due to exposure to high radiation. For this reason, small samples of pressure tubes are inspected at periodic intervals and small sample of data so collected are used as input to probabilistic analysis. The pressure tube flaw assessment is therefore confounded by large sampling uncertainties. Therefore, determination of adequate sample size is an important issue. In this thesis, a risk informed approach is proposed to define sample size requirement for flaw assessment. Notch-tip stress field is a key factor in any flaw assessment model. Traditionally, linear elastic fracture mechanics (LEFM) and its extension, serves the basis for determination of notch-tip stress field for elastic and elastic-perfectly-plastic material, respectively. However, the LEFM solution is based on small deformation theory and fixed crack geometry, which leads to singular stress and strain field at the crack-tip. The thesis presents new models for notch and crack induced stress fields based on the deformed geometry. In contrast with the classical solution based on small deformation theory, the proposed model uses the Cauchy's stress definition and boundary conditions which are coupled with the deformed geometry. This formulation also incorporates the rotation near the crack-tip, which leads to blunting and displacement of the crack-tip. The solution obtained based on the final deformed configuration yields a non-singular stress field at the crack-tip and a non-linear variation of stress concentration factor for both elastic and elastic-perfectly-plastic material. The proposed stress field formulation approach is applied to formulate an analytical model for estimating the threshold stress intensity factor (KIH) for DHC initiation. The analytical approach provides a relationship between KIH and temperature that is consistent with experimental results.

Probabilistic assessment and life cycle management of engineering components and systems in a nuclear power plant is intended to ensure safe and efficient operation of energy generation over its entire life. The CANDU reactor core consists of 380-480 pressure tubes, which are like miniature pressure vessels that contain natural uranium fuel. Pressure tubes operate under severe temperature and radiation conditions, which result in degradation with ageing. Presence of flaws in a pressure tube makes it vulnerable to delayed hydride cracking (DHC), which may lead to rupture or break-before-leak situation. Therefore, assessment of flaws in the pressure tubes is considered an integral part of a reactor core assessment program. The main objective of the thesis is to develop advanced probabilistic and mechanical stress field models for the assessment of flaws. The flaw assessment models used by the industries are based on deterministic upper/lower bound values for the variables and they ignore uncertainties associated with system parameters. In this thesis, explicit limit state equations are formulated and first order reliability method is employed for reliability computation, which is more efficient than simulation-based methods. A semi-probabilistic approach is adopted to develop an assessment model, which consists of a mechanics-based condition (or equation) involving partial factors that are calibrated to a specified reliability level. This approach is applied to develop models for DHC initiation and leak-before-break assessments. A novel feature of the proposed method is that it bridges the gap between a simple deterministic analysis and complex simulations, and it is amenable to practical applications. The nuclear power plant systems are not easily accessible for inspection and data collection due to exposure to high radiation. For this reason, small samples of pressure tubes are inspected at periodic intervals and small sample of data so collected are used as input to probabilistic analysis. The pressure tube flaw assessment is therefore confounded by large sampling uncertainties. Therefore, determination of adequate sample size is an important issue. In this thesis, a risk informed approach is proposed to define sample size requirement for flaw assessment. Notch-tip stress field is a key factor in any flaw assessment model. Traditionally, linear elastic fracture mechanics (LEFM) and its extension, serves the basis for determination of notch-tip stress field for elastic and elastic-perfectly-plastic material, respectively. However, the LEFM solution is based on small deformation theory and fixed crack geometry, which leads to singular stress and strain field at the crack-tip. The thesis presents new models for notch and crack induced stress fields based on the deformed geometry. In contrast with the classical solution based on small deformation theory, the proposed model uses the Cauchy's stress definition and boundary conditions which are coupled with the deformed geometry. This formulation also incorporates the rotation near the crack-tip, which leads to blunting and displacement of the crack-tip. The solution obtained based on the final deformed configuration yields a non-singular stress field at the crack-tip and a non-linear variation of stress concentration factor for both elastic and elastic-perfectly-plastic material. The proposed stress field formulation approach is applied to formulate an analytical model for estimating the threshold stress intensity factor (KIH) for DHC initiation. The analytical approach provides a relationship between KIH and temperature that is consistent with experimental results.