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.

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.

Reactor Pressure Vessel Integrity Assessment: French Simplified Analysis Method Applied to Underclad Postulated Defect in Nozzle

2016 ◽  
Author(s):  
Adolfo Arrieta-Ruiz ◽  
Eric Meister ◽  
Stéphane Vidard
Keyword(s):  
Fracture Mechanics ◽  
Fracture Risk ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Core Area ◽  
Integrity Assessment ◽  
Elastic Plastic ◽  
The Core ◽  
Simplified Method ◽  
Reactor Pressure

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main concerns regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Fast fracture risk is the main potential damage considered in the integrity assessment of RPV. In France, deterministic integrity assessment for RPV vis-à-vis the brittle fracture risk is based on the crack initiation stage. As regards the core area in particular, the stability of an under-clad postulated flaw is currently evaluated under a Pressurized Thermal Shock (PTS) through a dedicated fracture mechanics simplified method called “beta method”. However, flaw stability analyses are also carried-out in several other areas of the RPV. Thence-forward performing uniform simplified inservice analyses of flaw stability is a major concern for EDF. In this context, 3D finite element elastic-plastic calculations with flaw modelling in the nozzle have been carried out recently and the corresponding results have been compared to those provided by the beta method, codified in the French RSE-M code for under-clad defects in the core area, in the most severe events. The purpose of this work is to validate the employment of the core area fracture mechanics simplified method as a conservative approach for the under-clad postulated flaw stability assessment in the complex geometry of the nozzle. This paper presents both simplified and 3D modelling flaw stability evaluation methods and the corresponding results obtained by running a PTS event. It shows that the employment of the “beta method” provides conservative results in comparison to those produced by elastic-plastic calculations for the cases here studied.

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

Probabilistic Assessments of the Reactor Pressure Vessel Structural Integrity: Direct Coupling Between Probabilistic and Finite-Element Codes to Model Sensitivity to Key Thermo-Hydraulic Variability

2004 ◽  
Author(s):  
Etienne de Rocquigny ◽  
Yoan Chevalier ◽  
Silvia Turato ◽  
Eric Meister
Keyword(s):  
Finite Element ◽  
Pressure Vessel ◽  
Mechanical Model ◽  
Analytical Approach ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Direct Coupling ◽  
Integrity Assessment ◽  
Key Variables ◽  
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 as a nuclear operator the RPV integrity, Electricite´ de France (EDF) carries out probabilistic analyses. Probabilistic analyses become most interesting when some key variables, albeit conventionally taken at conservative values, can be modelled more accurately through statistical variability. In the context of low failure probabilities, this requires however a specific coupling effort between a specific probabilistic analysis method (e.g. Form-Sorm method) and the thermo-mechanical model to be reasonable in computing time. In this paper, the variability of a key variable — the mid-transient cooling temperature, tied to a climate-dependent tank — has been modelled, in some flaw configurations (axial sub-clad) for a French vessel. In a first step, a simplified analytical approach was carried out to assess its sensitivity upon the thermo-mechanical phenomena; hence, a direct coupling had to be implemented to allow a probabilistic calculation on the finite-element mechanical model, taking also into account a failure event properly defined through minimisation of the instantaneous failure margin during the transient. Comparison with the previous (indirectly-coupled) studies and the simplified analytical approach is drawn, demonstrating the interest of this new modelling effort to understand and order the sensitivity of the probability of crack initiation to the key variables. While being noticeable in the cases studied, sensitivity to the safety injection temperature variability proves to be less than the choice of the toughness model. Finally, regularity of the thermo-mechanical model is evidenced by the coupling exercise, suggesting that a modified response-surface based method could replace direct coupling for further investigation.

NESC-VII: Fracture Mechanics Analyses of WPS Experiments on Large-Scale Cruciform Specimen

2011 ◽  
Author(s):  
Shengjun Yin ◽  
Paul T. Williams ◽  
B. Richard Bass
Keyword(s):  
Fracture Mechanics ◽  
Large Scale ◽  
Structural Integrity ◽  
National Laboratory ◽  
Pressure Vessels ◽  
Cooperative Action ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Feasibility Test ◽  
Reactor Pressure

This paper describes numerical analyses performed to simulate warm pre-stress (WPS) experiments conducted with large-scale cruciform specimens within the Network for Evaluation of Structural Components (NESC-VII) project. NESC-VII is a European cooperative action in support of WPS application in reactor pressure vessel (RPV) integrity assessment. The project aims in evaluation of the influence of WPS when assessing the structural integrity of RPVs. Advanced fracture mechanics models will be developed and performed to validate experiments concerning the effect of different WPS scenarios on RPV components. The Oak Ridge National Laboratory (ORNL), USA contributes to the Work Package-2 (Analyses of WPS experiments) within the NESC-VII network. A series of WPS type experiments on large-scale cruciform specimens have been conducted at CEA Saclay, France, within the framework of NESC VII project. This paper first describes NESC-VII feasibility test analyses conducted at ORNL. Very good agreement was achieved between AREVA NP SAS and ORNL. Further analyses were conducted to evaluate the NESC-VII WPS tests conducted under Load-Cool-Transient-Fracture (LCTF) and Load-Cool-Fracture (LCF) conditions. This objective of this work is to provide a definitive quantification of WPS effects when assessing the structural integrity of reactor pressure vessels. This information will be utilized to further validate, refine, and improve the WPS models that are being used in probabilistic fracture mechanics computer codes now in use by the NRC staff in their effort to develop risk-informed updates to Title 10 of the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G.

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