Boiling Water Reactor Pressure Vessel Integrity Evaluation by Probabilistic Fracture Mechanics (PVP2010-25195)

2012 ◽  
Vol 135 (1) ◽  
Author(s):  
Bo-Yi Chen ◽  
Chin-Cheng Huang ◽  
Hsoung-Wei Chou ◽  
Ru-Feng Liu ◽  
Hsien-Chou Lin
Keyword(s):  
Fracture Mechanics ◽  
Conditional Probability ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Neutron Fluence ◽  
Probability Of Failure ◽  
Reactor Pressure Vessel ◽  
Circumferential Welds ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

The reactor pressure vessel (RPV) welds unavoidably degrade with the long time operation because of the fast neutron fluence exposure. Thus, the structural integrity of the axial and circumferential welds at the beltline region of reactor vessel must be evaluated carefully. The probabilistic fracture mechanics (PFM) analysis code: Fracture analysis of vessels—Oak Ridge (FAVOR), which has been verified by USNRC, is adopted in this work to calculate the conditional probability of initiation (CPI) and the conditional probability of failure (CPF) for the welds with 32 and 64 effective full power years (EFPY) operation, respectively. The Monte Carlo technique is involved in the simulation. This is the first time that the PFM technique is adopted for evaluating the risk of nuclear power plant components in Taiwan. Actual geometries, material properties, alloying elements, neutron fluence, and operation conditions are used for the plant specific analyses. Moreover, the design basis transients/accidents described in the final safety analysis report are also taken into account. The computed results show that the failure probabilities of welds are less than 10−10 per year. Only the axial weld, W-1001-08 has the probability of failure. The results of this work can be used to evaluate the structural integrity of the welds located at the RPV beltline region and provide the aging analysis results for the life extension and the license renewal applications.

Boiling Water Reactor Pressure Vessel Integrity Evaluation by Probabilistic Fracture Mechanics

2010 ◽  
Author(s):  
Bo-Yi Chen ◽  
Chin-Cheng Huang ◽  
Hsoung-Wei Chou ◽  
Ru-Feng Liu ◽  
Hsien-Chou Lin
Keyword(s):  
Fracture Mechanics ◽  
Conditional Probability ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Neutron Fluence ◽  
Probability Of Failure ◽  
Life Extension ◽  
Water Reactor ◽  
License Renewal ◽  
Reactor Pressure

The Chinshan boiling water reactor (BWR) units 1 and 2, owned by Taiwan Power Company (TPC), started commercial operations in 1978 and 1979, respectively. The reactor pressure vessel (RPV) welds unavoidably degrade with the long time operation because of the fast-neutron fluence exposure. This effect should be considered in the life extension and license renewal application. Thus, the structural integrity of the axial and circumferential welds at the beltline region of reactor vessel must be evaluated carefully. The probabilistic fracture mechanics (PFM) analysis code: Fracture Analysis of Vessels – Oak Ridge (FAVOR), which has been verified by USNRC, is adopted in this work to calculate the conditional probability of initiation (CPI) and the conditional probability of failure (CPF) for the welds with 32 and 64 effective full power years (EFPY) operation, respectively. The Monte Carlo technique is involved in the simulation. This is the first time that the PFM technique is adopted for evaluating the risk of nuclear power plant components in Taiwan. Actual geometries, material properties, chemistry components, neutron fluence and operation conditions are used for the plant specific analyses. Moreover, the design basis transients/accidents described in the final safety analysis report are also taken into account. The computed results show that the failure probabilities of welds are less than 10−10 per year. Only the axial weld, W-1001-08, is found to have the probability of failure. The results of this work can be used to evaluate the structural integrity of the welds located at the RPV beltline region, and provide the aging analysis results for the life extension and the license renewal applications.

Deterministic and Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Benchmark Comparisons

2003 ◽  
Author(s):  
Silvia Turato ◽  
Vincent Venturini ◽  
Eric Meister ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Safety Concern ◽  
Nuclear Regulatory Commission ◽  
Reactor Pressure Vessel ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Reactor Pressure

The structural integrity assessment of a nuclear Reactor Pressure Vessel (RPV) during accidental conditions, such as loss-of-coolant accident (LOCA), is a major safety concern. Besides Conventional deterministic calculations to justify the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Since in the USA the probabilistic fracture mechanics analyses are accepted by the Nuclear Regulatory Commission (NRC), a benchmark has been realized between EDF and Oak Ridge Structural Assessments, Inc. (ORSA) to compare the models and the computational methodologies used in respective deterministic and probabilistic fracture mechanics analyses. Six cases involving two distinct transients imposed on RPVs containing specific flaw configurations (two axial subclad, two circumferential surface-breaking, and two axial surface-braking flaw configurations) were defined for a French vessel. In two separate phases, deterministic and probabilistic, fracture mechanics analyses were performed for these six cases.

The Influence of Chemistry Concentration Uncertainty on the Probabilistic Fracture Mechanics Analysis

2011 ◽  
Author(s):  
Bo-Yi Chen ◽  
Chin-Cheng Huang ◽  
Hsuing-Wei Chou ◽  
Ru-Feng Liu ◽  
Hsien-Chou Lin
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Reactor Pressure Vessel ◽  
Oak Ridge ◽  
Probabilistic Fracture Mechanics ◽  
Full Power ◽  
Reactor Pressure

The chemistry concentration uncertainty of cooper and nickel significantly affects the shift in reference nil-ductility transition temperature (ΔRTNDT). The uncertainty comes from the methods and equipments applied in measurements, the lack of specimen in surveillance capsule, and the non-homogeneous of material. The variations of ΔRTNDT result in the differences of failure probability of reactor pressure vessel. In this study, the structural integrity of Chinshan boiling water reactor RPV shell welds was evaluated by probabilistic fracture mechanics code-Fracture Analysis of Vessel – Oak Ridge (FAVOR). The influence of chemistry concentration uncertainty on the fracture probability of Chinshan nuclear power plant RPV with 32 and 64 effective full power years (EFPY) operation was discussed. The results of this work can be used to evaluate the structural integrity of the welds located at the RPV beltline region, and provide the aging assessment of reactor pressure vessel.

scholarly journals Recent verification activities on probabilistic fracture mechanics analysis code PASCAL4 for reactor pressure vessel

2020 ◽  
Vol 7 (3) ◽  
pp. 19-00573-19-00573
Author(s):  
Kai LU ◽  
Jinya KATSUYAMA ◽  
Yinsheng LI ◽  
Yuhei MIYAMOTO ◽  
Takatoshi HIROTA ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Probabilistic Fracture Mechanics ◽  
Mechanics Analysis ◽  
Reactor Pressure

Fracture Mechanics Assessment of Reactor Pressure Vessel Irradiated Structural Steel for Short Column Type Supports and Neutron Shield Tank

2020 ◽  
Author(s):  
Alexandria M. Carolan ◽  
J. Brian Hall ◽  
Stephen K. Longwell ◽  
F. Arzu Alpan ◽  
Gregory M. Imbrogno ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Structural Steel ◽  
Pressure Vessel ◽  
Radiation Effects ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Radiation Embrittlement ◽  
Short Column ◽  
License Renewal ◽  
Reactor Pressure

Abstract As plants apply for 80 year licensure (subsequent license renewal), the United States Nuclear Regulatory Commission (U.S. NRC) has queried the nuclear power plant industry to investigate the impact of neutron embrittlement (radiation effects) on the reactor pressure vessel (RPV) structural steel supports due to extended plant operation past 60 years. The radiation effects on RPV supports were previously investigated and resolved as part of Generic Safety Issue No. 15 (GSI-15) in NUREG-0933 Revision 3 [1], NUREG-1509 [2] (published in May 1996), and NUREG/CR-5320 [3] (published in January 1989) for design life (40 years) and for first license renewal (20 additional years). The conclusions in NUREG-0933, Revision 3 stated that there were no structural integrity concerns for the RPV support structural steels; even if all the supports were totally removed (i.e. broken), the piping has acceptable margin to carry the load of the vessel. Nevertheless, for plants applying for 80 year life licensure, the U.S. NRC has requested an evaluation to show structural integrity of the RPV supports by accounting for radiation embrittlement (radiation damage) for continued operation into the second license renewal period (i.e. 80 years). The RPV support designs in light water reactors are grouped into one of five categories or types of supports: (1) skirt; (2) long-column; (3) shield-tank; (4) short column; and (5) suspension. In this paper, two of these RPV support configurations (short column supports and neutron shield tank) will be investigated using fracture mechanics to evaluate the effect of radiation embrittlement of the structural steel supports for long term operations (i.e. 80 years). The technical evaluation of other support configurations will be provided in a separate technical publication at a future date.

Development of Probabilistic Fracture Mechanics Analysis Codes for Reactor Pressure Vessels Considering Recent Embrittlement Model and Calculation Method of SIF: Progress of the Work

2016 ◽  
Author(s):  
Jongmin Kim ◽  
Bongsang Lee ◽  
Taehyun Kim ◽  
Yoonsuk Chang
Keyword(s):  
Stress Intensity ◽  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Calculation Method ◽  
Reactor Pressure Vessel ◽  
Radiation Embrittlement ◽  
Calculation Methods ◽  
Probabilistic Fracture Mechanics ◽  
Mechanics Analysis ◽  
Reactor Pressure

It is widely recognized that the state of knowledge and data for the probabilistic calculations which had been proposed in the early 1980s made a conservative treatment of several key factors and models. Recently, applications of some new radiation embrittlement model, material database, calculation method of stress intensity factors and others which can improve fracture mechanics assessment of reactor pressure vessel (RPV) are introduced. This improvement on the accuracy and reliability of the probabilistic fracture mechanics (PFM) analysis necessitated changes in PFM analysis procedures and calculations. Modification and application of newly developed models and calculation methods are the main target of developing a probabilistic fracture mechanics analysis code based on the structure of existing R-PIE and VISA computer code to reflect the latest technical basis. Failure probabilities of reactor pressure vessel under pressurized thermal shock (PTS) conditions were calculated through finite difference method (FDM) and Monte Carlo simulation techniques with user friendly graphic interface. Moreover, various radiation embrittlement models and calculation methods of stress intensity factor at crack tip based on AFCEN code are applied and verified in the present work.

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