Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

Chou, Hsoung-Wei; Huang, Chun-Ming

doi:10.1155/2015/785041

Cited by 5 publications

(2 citation statements)

References 7 publications

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“…During long-term operation, the fast neutron fluence (E > 0.5 MeV) causes the ferritic steel of the RPV to become susceptible to brittle fracture, especially shell and weld materials in the beltline region corresponding to the reactor core. The embrittled vessel shell may fracture due to a preexisting fabrication flaw which could lead to a through-wall crack [55]. To experimentally determine the value of ∆T DBT and, therefore, the fracture toughness of irradiated material, surveillance capsules are included in the vessel between the core and the wall.…”

Section: Failure Mode and Effect Analysismentioning

confidence: 99%

Evolution of Standardized Specifications on Materials, Manufacturing and In-Service Inspection of Nuclear Reactor Vessels

et al. 2021

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The cataloguing and revision of reactor pressure vessels (RPV) manufacturing and in-service inspection codes and their standardized material specifications—as a technical heritage—are essential for understanding the historical evolution of criteria and for enabling the comparison of the various national regulations, integrating the most relevant results from the scientific research. The analysis of the development of documents including standardized requirements and the comparison of regulations is crucial to be able to implement learned lessons and comprehend the progress of increasingly stringent safety criteria, contributing to sustainable nuclear power generation in the future. A novel methodology is presented in this work where a thorough review of the regulations and technical codes for the manufacture and in-service inspection of RPVs, considering the implementation of scientific advances, is performed. In addition, an analysis focused on the differences between irradiation embrittlement prediction models and acceptance criteria for detected defects (both during manufacturing and in-service inspection) described by the different technical codes as required by different national regulations such as American, German, French or Russian is performed. The most stringent materials requirements for RPV manufacturing are provided by the American and German codes. The French code is the most stringent with respect to the reference defect size using as a criterion in the in-service inspection.

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Section: Failure Mode and Effect Analysismentioning

confidence: 99%

Evolution of Standardized Specifications on Materials, Manufacturing and In-Service Inspection of Nuclear Reactor Vessels

et al. 2021

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“…During long-term operation, the fast neutron fluence causes the ferritic steel of RPV to be susceptible to brittle fracture, especially in the beltline region corresponding to the reactor core. The embrittled vessel shell may fracture due to a preexisting fabrication flaw and leads to a through-wall crack [12]. Nevertheless, the use of Charpy data from experimental testing is a time-consuming strategy to estimate ductile-to-brittle transition temperature.…”

Section: Introductionmentioning

confidence: 99%

Multicriteria Analytical Model for Mechanical Integrity Prognostics of Reactor Pressure Vessels Manufactured from Forged and Rolled Steels

Rodríguez-Prieto

Callejas²,

Primera

et al. 2022

Mathematics

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The aim of this work is to present a new analytical model to evaluate jointly the mechanical integrity and the fitness-for-service of nuclear reactor pressure-vessels steels. This new methodology integrates a robust and regulated irradiation embrittlement prediction model such as the ASTM E-900 with the ASME Fitness-for-Service code used widely in other demanding industries, such as oil and gas, to evaluate, among others, the risk of experiencing degradation mechanisms such as the brittle fracture (generated, in this case, due to the irradiation embrittlement). This multicriteria analytical model, which is based on a new formulation of the brittle fracture criterion, allows an adequate prediction of the irradiation effect on the fracture toughness of reactor pressure-vessel steels, letting us jointly evaluate the mechanical integrity and the fitness-for-service of the vessel by using standardized limit conditions. This allows making decisions during the design, manufacturing and in-service of reactor pressure vessels. The results obtained by the application of the methodology are coherent with several historical experimental works.

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Probabilistic fracture analysis for boiling water reactor vessels considering seismic loads during decommissioning transition period

Chou

Chang

2022

Annals of Nuclear Energy

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Probabilistic Structural Integrity Analysis of Boiling Water Reactor Pressure Vessel under Low Temperature Overpressure Event

Cited by 5 publications

References 7 publications

Evolution of Standardized Specifications on Materials, Manufacturing and In-Service Inspection of Nuclear Reactor Vessels

Evolution of Standardized Specifications on Materials, Manufacturing and In-Service Inspection of Nuclear Reactor Vessels

Multicriteria Analytical Model for Mechanical Integrity Prognostics of Reactor Pressure Vessels Manufactured from Forged and Rolled Steels

Probabilistic fracture analysis for boiling water reactor vessels considering seismic loads during decommissioning transition period

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