A Probabilistic Approach to Heatup and Cooldown Operating Limits for a Nuclear Pressure Vessel

2004 ◽  
Author(s):  
Randy G. Lott ◽  
William L. Server
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Probabilistic Approach ◽  
System Pressure ◽  
Operating Limits ◽  
Temperature Operating ◽  
Upper Bound Estimate ◽  
Low Probability ◽  
Material Fracture ◽  
Plant Integrity

During plant heatup and cooldown, the stresses in a nuclear pressure vessel are the membrane stress associated with system pressure and the stress caused by the thermal gradient across the vessel wall. The system pressure and the rate of temperature change (heatup or cooldown) must be controlled in a manner that protects against failure of the vessel. The current basis for setting these pressure-temperature operating limits is a deterministic fracture mechanics analysis that conservatively applies a safety of factor of two to the system pressure and assumes a large flaw corresponding to 1/4 of the thickness of the vessel. This calculation produces an upper bound estimate of the stress intensity at the flaw tip, which is compared to a lower bound estimate of the material fracture toughness to set the pressure-temperature operating limits. This methodology results in restrictive operating limits that correspond to an extremely low probability of vessel failure. The development of state-of-the art probabilistic fracture mechanics for the FAVOR Code has made it possible to determine the risk of vessel failure associated with these restrictive limits. Relaxation of these limits may be possible with negligible impact on the risk of vessel failure. Overall plant risk may be reduced by eliminating other potential challenges to plant integrity that are associated with the current restrictive operating limits.

Uncertainty Sampling of Weld Residual Stress Fields in Probabilistic Analysis: Part II — Examples

2016 ◽  
Author(s):  
Robert E. Kurth ◽  
Cédric J. Sallaberry ◽  
Frederick W. Brust ◽  
Elizabeth A. Kurth ◽  
Michael L. Benson ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Residual Stresses ◽  
Assessment Tool ◽  
Lessons Learned ◽  
Piping System ◽  
Primary System ◽  
Cracking Behavior ◽  
System Pressure ◽  
Piping Systems ◽  
Low Probability

NRC Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to assess compliance with the 10CFR50 Appendix A, GDC-4 requirement that primary system pressure piping exhibit an extremely low probability of rupture. SRP 3.6.3 does not allow for assessment of piping systems with active degradation mechanisms, such as Primary Water Stress Corrosion Cracking (PWSCC) which is currently occurring in systems that have been granted LBB approvals. US NRC staff, working cooperatively with the Electric Power Research Institute through a memorandum of understanding, conducted a multi-year project that focused on the development of a viable method and approach to address the effects of PWSCC in primary piping systems approved for LBB. This project, called eXtremely Low Probability of Rupture (xLPR) [1], defined the requirements necessary for a modular-based probabilistic fracture mechanics assessment tool to directly assess compliance with the regulations. Using the lessons learned from the pilot study, the production version of this code, designated as Version 2.0, focused on those primary piping systems previously approved for LBB. In this version the appropriate fracture mechanics-based models are employed to model the physical cracking behavior and a variety of computational options are provided to characterize, categorize and propagate problem uncertainties. One of the most influential uncertainty on risk in the xLPR code is the one associated with weld residual stresses (WRS). WRS plays a key role in both crack initiation and crack growth. PWSCC is mainly driven by tensile stresses, whose major contributors are the tensile weld residual stresses that develop during fabrication of the piping system. Handling the uncertainty involved with WRS within a probabilistic framework is quite challenging. A companion paper presents the selected approach to represent uncertainty within the framework of the xLPR code while respecting a set of requirements in term of smoothness of profile, efficiency of (potential) importance sampling and (for axial WRS) equilibrium. This paper illustrate with examples the implementation of the described methods into xLPR v2.0.

Development of the Extremely Low Probability of Rupture (xLPR) Version 2.0 Code

2015 ◽  
Author(s):  
D. Rudland ◽  
C. Harrington ◽  
R. Dingreville
Keyword(s):  
Pilot Study ◽  
Fracture Mechanics ◽  
Lessons Learned ◽  
Primary System ◽  
Cracking Behavior ◽  
Code Structure ◽  
System Pressure ◽  
Piping Systems ◽  
Version 2.0 ◽  
Low Probability

NRC Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to assess compliance with the 10CFR50 Appendix A, GDC-4 requirement that primary system pressure piping exhibit an extremely low probability of rupture. SRP 3.6.3 does not allow for assessment of piping systems with active degradation mechanisms, such as Primary Water Stress Corrosion Cracking (PWSCC) which is currently occurring in systems that have been granted LBB approvals. US NRC staff, working cooperatively with the Electric Power Research Institute through a memorandum of understanding, conducted a multi-year project that focused on the development of a viable method and approach to address the effects of PWSCC in primary piping systems approved for LBB. This project, called eXtremely Low Probability of Rupture (xLPR), defined the requirements necessary for a modular-based probabilistic fracture mechanics assessment tool to directly assess compliance with the regulations. As reported in previous technical papers, the first version of the xLPR code was developed as part of a pilot study, which leveraged existing fracture mechanics based models and software coupled to both a commercial and open source code framework to determine the framework and architecture requirements appropriate for building a modular-based code with this complexity. Using the lessons learned from the pilot study, the production version of this code, designated as Version 2.0, focuses on those primary piping systems previously approved for LBB. In this version, the appropriate fracture mechanics-based models are employed to model the physical cracking behavior and a variety of computational options are provided to characterize, categorize and propagate problem uncertainties. This paper examines the xLPR Version 2.0 model by presenting a brief overview of the xLPR scope, the code structure, computational framework and fracture mechanics-based models. As a demonstration of the xLPR Version 2.0 capabilities, an example is presented that focuses on PWSCC in large-bore piping systems. This example exercises some functionalities of the xLPR code to demonstrate its application to assess compliance with 10CFR50 Appendix A, GDC-4. This paper concludes with a brief discussion on the path forward and plans for the control and maintenance of the xLPR code.

Uncertainty Sampling of Weld Residual Stress Fields in Probabilistic Analysis: Part I — Theory

2016 ◽  
Author(s):  
Robert E. Kurth ◽  
Cédric J. Sallaberry ◽  
Frederick W. Brust ◽  
Elizabeth A. Kurth ◽  
Michael L. Benson ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Residual Stresses ◽  
Assessment Tool ◽  
Lessons Learned ◽  
Piping System ◽  
Primary System ◽  
Cracking Behavior ◽  
System Pressure ◽  
Piping Systems ◽  
Low Probability

NRC Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to assess compliance with the 10CFR50 Appendix A, GDC-4 requirement that primary system pressure piping exhibit an extremely low probability of rupture. SRP 3.6.3 does not allow for assessment of piping systems with active degradation mechanisms, such as Primary Water Stress Corrosion Cracking (PWSCC) which is currently occurring in systems that have been granted LBB approvals. US NRC staff, working cooperatively with the Electric Power Research Institute through a memorandum of understanding, conducted a multi-year project that focused on the development of a viable method and approach to address the effects of PWSCC in primary piping systems approved for LBB. This project, called eXtremely Low Probability of Rupture (xLPR), defined the requirements necessary for a modular-based probabilistic fracture mechanics assessment tool to directly assess compliance with the regulations [1]. Using the lessons learned from the pilot study [2] the production version of this code, designated as Version 2.0, focused on those primary piping systems previously approved for LBB [3]. In this version the appropriate fracture mechanics-based models are employed to model the physical cracking behavior and a variety of computational options are provided to characterize, categorize and propagate problem uncertainties. One of the most influential sources of uncertainty on risk in the xLPR code is the one associated with weld residual stresses (WRS). WRS plays a key role in both crack initiation and crack growth. PWSCC is mainly driven by tensile stresses, whose major contributors are the tensile weld residual stresses that develop during fabrication of the piping system. Handling the uncertainty involved with WRS within a probabilistic framework is quite challenging. This paper presents the selected approach to represent uncertainty within the framework of the xLPR code while respecting a set of requirements in term of smoothness of profile, efficiency of (potential) importance sampling and (for axial WRS) equilibrium. The current WRS sampling scheme employs correlation in order to smooth the shape of the WRS fields through the thickness of a dissimilar metal weld. This method presents an enrichment of the Cholesky decomposition on the correlation matrix, in order to satisfy the other two requirements.

Benchmarking PRAISE-CANDU 1.0 With xLPR 1.0

2013 ◽  
Author(s):  
Min Wang ◽  
Xinjian Duan ◽  
Michael J. Kozluk
Keyword(s):  
Quality Assurance ◽  
Pilot Study ◽  
Fracture Mechanics ◽  
Software Quality ◽  
Verification And Validation ◽  
Quality Assurance Program ◽  
Software Quality Assurance ◽  
Full Compliance ◽  
Low Probability ◽  
State Of Art

A probabilistic fracture mechanics code, PRAISE-CANDU 1.0, has been developed under a software quality assurance program in full compliance with CSA N286.7-99, and was initially released in 2012 June. Extensive verification and validation has been performed on PRAISE-CANDU 1.0 for the purpose of software quality assurance. This paper presents the benchmarking performed between PRAISE-CANDU 1.0 and xLPR (eXtremely Low Probability of Rupture) version 1.0 using the cases from the xLPR pilot study. The xLPR code was developed in a configuration management and quality assured manner. Both codes adopted a state-of-art code architecture for the treatment of the uncertainties. Inputs to the PRAISE-CANDU were established as close as possible to those used in corresponding xLPR cases. Excellent agreement has been observed among the results obtained from the two PFM codes in spite of some differences between the codes. This benchmarking is considered to be an important element of the validation of PRAISE-CANDU.

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
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