Pressurized thermal shock analyses of a reactor pressure vessel using critical crack depth diagrams

Jhung, Myung Jo; Park, Youn Won; Jang, Changheui

doi:10.1016/s0308-0161(99)00063-0

Cited by 9 publications

(2 citation statements)

References 6 publications

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“…It is expected that most failures are observed at small flaw depths and no The ASME code suggests the calculation of stress intensity factor for surface flaw by applying the internal vessel pressure on the surface of the crack and also by considering the plastic zone correction factor. 7) This may cause the difference in stress intensity factors calculated between ASME method and other codes or finite element methods that did not consider them, 10) also generating different probabilities of vessel failures. Therefore, the effect of these two considerations is investigated to calculate the stress intensity factor.…”

Section: Resultsmentioning

confidence: 99%

Structural Integrity of Reactor Pressure Vessel for Small Break Loss of Coolant Accident

Jhung

Choi

JANG

2009

Journal of Nuclear Science and Technology

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Plant-specific analyses of the typical pressurized water reactor in Korea are performed to assure the structural integrity of the reactor pressure vessel during transient, which is expected to initiate pressurized thermal shock event. The deterministic analysis is performed to determine the critical time interval in the transient during which mitigating action can be effective. Also, the failure probability is obtained by performing probabilistic fracture mechanics analysis. In addition, several sensitivity analyses are performed for warm prestressing, plastic zone correction, and simulation option to calculate fracture toughness etc. to assess its effect on the failure probability.

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Section: Resultsmentioning

confidence: 99%

Structural Integrity of Reactor Pressure Vessel for Small Break Loss of Coolant Accident

Jhung

Choi

JANG

2009

Journal of Nuclear Science and Technology

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“…Even though the RT NDT of the material is higher than the upper bound determined by the tangent criterion, the crack will not be initiated due to warm prestressing effect if it is lower than the upper bound determined by the maximum criterion. Therefore, the range of allowable RT NDT is determined by two criteria, tangent criterion and maximum criterion, depending on the warm prestressing effect [7,8].…”

Section: Analysis Methodsmentioning

confidence: 99%

Structural integrity assessment of reactor pressure vessels during pressurized thermal shock

et al. 2008

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A comparative assessment study is performed for the deterministic fracture mechanics approach of the pressurized thermal shock of a reactor pressure vessel. Round robin problems consisting of two transients and two defects are solved. Their results are compared to suggest some recommendations of best practices and to assure an understanding of the key parameters of this type of approach, which will be helpful not only for the benchmark calculations and results comparisons but also as a part of the knowledge management for the future generation. Seven participants from five organizations solved the problem and their results are compiled in this study.

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Analysis of dynamic conditions during thermal transient events for the structural assessment of a nuclear vessel

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Lacalle

Gorrochategui

et al. 2010

Engineering Failure Analysis

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Pressurized thermal shock analyses of a reactor pressure vessel using critical crack depth diagrams

Cited by 9 publications

References 6 publications

Structural Integrity of Reactor Pressure Vessel for Small Break Loss of Coolant Accident

Structural Integrity of Reactor Pressure Vessel for Small Break Loss of Coolant Accident

Structural integrity assessment of reactor pressure vessels during pressurized thermal shock

Analysis of dynamic conditions during thermal transient events for the structural assessment of a nuclear vessel

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