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.

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.

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.

scholarly journals Analysis for the Effect of Spatial Discretization Method on AP1000 Reactor Pressure Vessel Fluence Calculation

2016 ◽  
Vol 2016 ◽  
pp. 1-10 ◽  
Author(s):  
Junxiao Zheng ◽  
Bin Zhang ◽  
Shengchun Shi ◽  
Yixue Chen
Keyword(s):  
Fast Neutron ◽  
Pressure Vessel ◽  
Power Plants ◽  
Structural Integrity ◽  
Neutron Fluence ◽  
Mesh Size ◽  
Reactor Pressure Vessel ◽  
Life Extension ◽  
Discrete Ordinates ◽  
Reactor Pressure

Maintaining the structural integrity of the reactor pressure vessel (RPV) is a critical concern related to the safe operation of nuclear power plants. To estimate the structural integrity over the designed lifetime and to support analyses for a potential plant life extension, an accurate calculation of the fast neutron fluence (E>1.0 MeV orE>0.1 MeV) at the RPV is significant. The discrete ordinates method is one of the main methods to solve such problems. During the calculation process, many factors will affect the results. In this paper, the deviations introduced by different differencing schemes and mesh sizes on the AP1000 RPV fast neutron fluence have been studied, which are based on new discrete ordinates code ARES. The analysis shows that the differencing scheme (diamond difference with or without linear zero fix-up, theta weighted, directional theta weighted, and exponential directional weighted) introduces a deviation within 4%. The coarse mesh (4 × 4 cm meshes inXYplane) leads to approximately 23.7% calculation deviation compared to those of refined mesh (1 × 1 cm meshes inXYplane). Comprehensive study on the deviation introduced by differencing scheme and mesh size has great significance for reasoned evaluation of RPV fast neutron fluence calculation results.

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.

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

2015 ◽  
Vol 2015 ◽  
pp. 1-9
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Low Temperature ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Analysis Model ◽  
Water Reactor ◽  
Integrity Analysis ◽  
Reactor Pressure

The probabilistic structural integrity of a Taiwan domestic boiling water reactor pressure vessel has been evaluated by the probabilistic fracture mechanics analysis. First, the analysis model was built for the beltline region of the reactor pressure vessel considering the plant specific data. Meanwhile, the flaw models which comprehensively simulate all kinds of preexisting flaws along the vessel wall were employed here. The low temperature overpressure transient which has been concluded to be the severest accident for a boiling water reactor pressure vessel was considered as the loading condition. It is indicated that the fracture mostly happens near the fusion-line area of axial welds but with negligible failure risk. The calculated results indicate that the domestic reactor pressure vessel has sufficient structural integrity until doubling of the present end-of-license operation.

Verifying Calculation of Reactor Pressure Vessel Fast Neutron Fluence

2014 ◽  
Vol 986-987 ◽  
pp. 985-989
Author(s):  
Qiao Feng Liu ◽  
Jing Ru Han ◽  
Hai Ying Chen ◽  
Chun Ming Zhang
Keyword(s):  
Standard Deviation ◽  
Fast Neutron ◽  
Radiation Protection ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
Reactor Pressure Vessel ◽  
Light Water Reactor ◽  
Light Water ◽  
Water Reactor ◽  
Reactor Pressure

The reactor pressure vessel is an unchangeable component of the light water reactor. To some extent, the life of the pressure vessel depends on the fast neutron fluence. In addition, the fast neutron fluence is an important parameter for radiation protection. So, the fast neutron fluence is one of the main parameters which should be verifying calculated by the reviewers. The verifying calculation of the fast neutron fluence of one reactor pressure vessel is presented in this paper, and the standard deviation between the verifying and designing calculations is lower than 10%. The reasons for the deviation are discussed.

Probabilistic Structural Integrity Evaluation on a Pressurized Water Reactor Pressure Vessel Under Pressure–Temperature Limit Operations

2015 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Temperature Limit ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

The normal reactor startup (heat-up) and shut-down (cool-down) operation limits are defined by the ASME Code Section XI-Appendix G, to ensure the structural integrity of the embrittled nuclear reactor pressure vessels (RPVs). In the paper, the failure risks of a Taiwan domestic pressurized water reactor (PWR) pressure vessel under various pressure-temperature limit operations are analyzed. Three types of pressure-temperature limit curves established by different methodologies, which are the current operation limits of the domestic RPV based on the KIa fracture toughness curve in 1998 or earlier editions of ASME Section XI-Appendix G, the recently proposed limits according to the KIC fracture toughness curve after the 2001 edition of ASME Section XI-Appendix G, and the risk-informed revision method proposed in MRP-250 report that provides more operational flexibility, are considered. The ORNL’s probabilistic fracture mechanics code, FAVOR, is employed to perform a series of fracture probability analyses for the RPV at multiple levels of embrittlement under such pressure-temperature limit transients. The analysis results indicate that the pressure-temperature operation limits associated with more operational flexibility will result in higher failure risks to the RPV. The shallow inner surface breaking flaw due to the clad fabrication defect is the most critical factor and dominates the failure risk of the RPV under pressure-temperature limit operations. Present work can provide a risk-informed reference for the safe operation and regulation of PWRs in Taiwan.

Comparison of Pressure-Temperature Limits for a Pressurized Water Reactor Pressure Vessel Considering Beltline and Extended Beltline Regions

2018 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Yu-Yu Shen ◽  
Chin-Cheng Huang
Keyword(s):  
Pressure Vessel ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Pressure Vessels ◽  
Reference Temperature ◽  
Regulatory Body ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

To ensure the structural integrity of the embrittled reactor pressure vessels (RPVs) during startup or shutdown operation, the pressure-temperature (P-T) limits are mainly determined by the fracture toughness of beltline region material with the highest level of neutron embrittlement. However, other vessel parts such as nozzles with structural discontinuities may affect the limits due to the higher stress concentration, even though the neutron embrittlement is insignificant. Therefore, not only beltline material with the highest reference temperature, but also other components with structural discontinuities have to be considered for the development of P-T limits of RPV. In the paper, the pressure-temperature operational limits of a Taiwan domestic pressurized water reactor (PWR) pressure vessel considering beltline and extended beltline regions are established per the procedure of ASME Code Section XI-Appendix G. The three-dimensional finite element models of PWR inlet and outlet nozzles above the beltline region are also built to analyze the pressure and thermal stress distributions for P-T limits calculation. The analysis results indicate that the cool-down P-T limit of the domestic PWR vessel is still dominated by the beltline region, but the heat-up limit is partially controlled by the extended beltline region. On the other hand, the relations of reference temperature between nozzles and beltline region on the P-T limits are also discussed. Present work could be a reference for the regulatory body and is also helpful for safe operation of PWRs in Taiwan.

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