Predicting Pipe Rupture Frequencies Using xLPR

2020 ◽  
Author(s):  
David L. Rudland
Keyword(s):  
Assessment Tool ◽  
Leak Detection ◽  
Nuclear Regulatory Commission ◽  
Pipe Diameter ◽  
Nozzle Geometry ◽  
Pressurized Water ◽  
Version 2.0 ◽  
Low Probability ◽  
Leak Before Break ◽  
Service Inspection

Abstract Over the last several years, the U.S. Nuclear Regulatory Commission (NRC), in cooperation with the Electric Power Research Institute (EPRI), conducted a multi-year project that focused on the development of a viable method and approach to address the effects of primary water stress corrosion cracking (PWSCC) in primary piping systems approved for leak-before-break (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. Version 2.0 of this code has been completed and is currently awaiting public release. Since the focus of xLPR Version 2.0 is investigating the impacts of active piping degradation on the leak-before-break behavior of reactor coolant piping, questions have been raised to whether xLPR can be used to confirm pipe rupture frequencies developed in other efforts, such as NUREG-1829, “Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process.” This paper discusses an initial study focused on whether xLPR can be used to estimate pipe rupture frequencies. A series of analyses were conducted, based on inputs developed by the xLPR program team, focused on the reactor pressure vessel outlet nozzle geometry of a typical pressurized water reactor. Additional analyses were conducted using the same radius-to-thickness ratio but decreasing the pipe diameter. Due to computer memory restrictions, it was difficult calculating low probability events when considering PWSCC initiation, typical residual stresses, leak detection and in-service inspection. Therefore, to bound the problem, an aggressive weld residual stress was assumed with multiple pre-existing defects. By modifying the size and number of these initial defects, results were generated that indicated the conditional probability of rupture was related to the percentage of the inner circumference cracked and the pipe diameter. Using the PWSCC initiation model from xLPR Version 2, the yearly rupture frequency with leak detection and in-service inspection was calculated. The results indicate that the rupture frequencies in NUREG-1829 appear conservative relative to the results from this study. Due to the limited scope of this study, the assumptions used in these analyses were limited or conservative; therefore, additional analyses are needed for a more robust comparison. However, the results suggest that conducting xLPR analyses with pre-existing defects may be useful in bounding LBB applicability with active degradation.

The Treatment of ISI Uncertainty in Extremely Low Probability of Rupture

2011 ◽  
Author(s):  
Patrick G. Heasler ◽  
Scott E. Sanborn ◽  
Steven R. Doctor ◽  
Michael T. Anderson
Keyword(s):  
Epistemic Uncertainty ◽  
Probability Distributions ◽  
Operating Experience ◽  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Piping Systems ◽  
Primary Water ◽  
Low Probability ◽  
Regulatory Commission ◽  
Leak Before Break

The U.S. Nuclear Regulatory Commission (NRC) in cooperation with the nuclear industry is constructing an improved probabilistic fracture model for piping systems that in the past have not been susceptible to known degradation processes that could lead to pipe rupture. Recent operating experience with primary water stress corrosion cracking (PWSCC) has challenged this prior position of leak-before-break and which has now become known as “extremely Low Probability of Rupture” (xLPR). This paper focuses on the xLPR model’s treatment of uncertainty for in-service inspection. In the xLPR model, uncertainty is classified as either aleatory or epistemic, and both types of uncertainty are described with probability distributions. Earlier PFM models included aleatory, but ignored epistemic, uncertainty, or attempted to deal with epistemic uncertainty by use of conservative bounds. Thus, inclusion of both types of uncertainty in xLPR should produce more realistic results than the earlier models. This work shows that by including epistemic uncertainty in the xLPR ISI module, there can be a significant effect on rupture probability; however, this depends upon the specific scenarios being studied. Some simple scenarios are presented to illustrate those where there is no effect and those having a significant effect on the probability of rupture.

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

2012 ◽  
Author(s):  
D. Rudland ◽  
C. Harrington
Keyword(s):  
Pilot Study ◽  
Piping System ◽  
Primary System ◽  
Development Effort ◽  
Degradation Mechanisms ◽  
Pressurized Water ◽  
Initial Development ◽  
System Pressure ◽  
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. Along with a series of existing qualitative steps to assure safety in LBB-approved lines experiencing PWSCC, NRC staff, working cooperatively with the Electric Power Research Institute through a memorandum of understanding, is developing a new, modular based, comprehensive piping system assessment methodology to directly assess compliance with the regulations. This project, called eXtremely Low Probability of Rupture (xLPR), will model the effects and uncertainties of relevant active degradation mechanisms and the associated mitigation activities. The resulting analytical tool will be comprehensive with respect to known and significant materials challenges (PWSCC, etc.), vetted with respect to the technical bases of models and inputs, flexible enough to permit analysis of a variety of in-service situations and adaptable such as to accommodate evolving and improving knowledge. A multi-year project has begun that has been focused on the development of a viable method and approach to address the effects of PWSCC as well as define the requirements necessary for such a modular-based assessment tool. As reported in a previous paper, the first version of this 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. The pilot study focused on PWSCC in pressurizer surge nozzles, and was meant to demonstrate the feasibility of this code and approach and not to determine the absolute values of the probability of rupture. This paper examines the plans for the xLPR Version 2.0 model which will broaden the scope of xLPR to all LBB-approved primary piping in pressurized water reactors (PWR), using an incremental approach that incorporates the design requirements and lessons learned from previous iterations. After a review of the Version 1.0 final results, this paper will document the plans for Version 2.0 including the revised management structure, the technical scope, and the progress in the code development effort to date.

Sensitivity Analysis for XLPR Acceptance Testing

2017 ◽  
Author(s):  
Cédric J. Sallaberry ◽  
Robert E. Kurth ◽  
Frederick W. Brust ◽  
Elizabeth A. Kurth
Keyword(s):  
Sensitivity Analysis ◽  
Nuclear Regulatory Commission ◽  
Acceptance Testing ◽  
Analysis Tool ◽  
The Us ◽  
Perform Sensitivity Analysis ◽  
Regression Techniques ◽  
Version 2.0 ◽  
Regulatory Commission ◽  
Leak Before Break

The US Nuclear Regulatory Commission (US NRC) and the Electric Power Research Institute (EPRI) have jointly funded the development of the probabilistic analysis tool xLPR (extremely Low Probability of Rupture) version 2.0 for calculating the probability of a leak before break for welded pipes [1]. The acceptance testing for the program was completed in July of 2016. This allowed the release of the code to the development team. The released version of xLPR v 2.0 was used to run probabilistic simulations on selected scenarios and then perform uncertainty and sensitivity analysis on selected results. Uncertainty Analysis (UA) informs on how much the uncertainty affect the outputs while Sensitivity Analysis (SA) ranks the uncertain parameters in term of importance (i.e., influence over uncertainty) for selected outputs. In this paper, we present the methodology used to perform Sensitivity Analysis using three regression techniques as well as scatterplots for graphical study. The results generated are consistent both with the experts’ expectations and our understanding of the equations used in the sub-models.

Development of Initial Input Parameters for a Probabilistic Pipe Rupture Assessment Code for Nuclear Reactor Coolant Pressure Boundary Leak-Before-Break Analyses

2010 ◽  
Author(s):  
Eric M. Focht ◽  
Guy DeBoo
Keyword(s):  
Pilot Study ◽  
Nuclear Reactor ◽  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Primary System ◽  
Reactor Coolant ◽  
System Pressure ◽  
Input Parameters ◽  
Low Probability ◽  
Leak Before Break

The U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) entered into a cooperative agreement to develop a modular computer code for determining the probability of rupture of reactor coolant pressure boundary piping. The goal of xLPR project (Extremely Low Probability of Rupture), as it is known, is to develop a flexible modular code that satisfies the assessment methodology in the NRC Standard Review Plan (SRP) 3.6.3 leak-before-break analysis. The NRC SRP 3.6.3 provides guidance on satisfying Title 10 of the Code of Federal Regulations Part 50 (10 CFR Part 50) Appendix A, General Design Criteria 4 that requires primary system pressure piping to exhibit an extremely low probability of rupture. The initial phase of the xLPR project is the Pilot Study where the base structure of the code will be developed and tested on a known case. The Pilot Study version of the xLPR code consists of an alpha version where proxy modules and inputs are used and a beta version where the state-of-the art modules and inputs are used. This paper will discuss the input parameters used for the alpha xLPR code.

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

A Risk-Informed Methodology for ASME Section XI, Appendix G

2012 ◽  
Vol 134 (3) ◽  
Author(s):  
Ronald Gamble ◽  
William Server ◽  
Bruce Bishop ◽  
Nathan Palm ◽  
Carol Heinecke
Keyword(s):  
Nuclear Regulatory Commission ◽  
The United States ◽  
Service Level ◽  
Pressure Vessels ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Leak Testing ◽  
Alternative Methodology ◽  
Water Reactors ◽  
Regulatory Commission

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Appendix G provides a deterministic procedure for defining Service Level A and B pressure–temperature limits for ferritic components in the reactor coolant pressure boundary. An alternative risk-informed methodology has been developed for ASME Section XI, Appendix G. This alternative methodology provides easy to use procedures to define risk-informed pressure–temperature limits for Service Level A and B events, including leak testing and reactor start-up and shut-down. Risk-informed pressure–temperature limits provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials. This work evaluated selected plants spanning the population of pressurized water reactors (PWRs) and boiling water reactors (BWRs). The evaluation included determining appropriate material properties, reviewing operating history and system operational constraints, and performing probabilistic fracture mechanics (PFM) analyses. The analysis results were used to define risk-informed pressure–temperature relationships that comply with safety goals defined by the United States (U.S.) Nuclear Regulatory Commission (NRC). This alternative methodology will provide greater operational flexibility, especially for Service Level A and B events that may adversely affect efficient and safe plant operation, such as low-temperature-over-pressurization for PWRs and system leak testing for BWRs. Overall, application of this methodology can result in increased plant efficiency and increased plant and personnel safety.

Rupture Hardware Minimization in Pressurized Water Reactor Piping

1989 ◽  
Vol 111 (1) ◽  
pp. 64-71 ◽  
Author(s):  
S. K. Mukherjee ◽  
J. J. Szy Slow Ski ◽  
V. Chexal ◽  
D. M. Norris ◽  
N. A. Goldstein ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Ferritic Steel ◽  
High Energy ◽  
Pressurized Water Reactor ◽  
Pipe Failure ◽  
Pressurized Water ◽  
Screening Criteria ◽  
Water Reactor ◽  
Leak Before Break ◽  
Generic Sense

For much of the high-energy piping in light water reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but also improves the overall safety and integrity of the plant since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied at Beaver Valley Power Station—Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferritic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in. (152-mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel line as small as 3-in. (76-mm) diameter (outside containment) can qualify for pipe rupture hardware elimination.

New Needs of Fracture Mechanic Analysis at Design and Operation Level: Status of French Nuclear Mechanical Codes

2018 ◽  
Author(s):  
Claude Faidy
Keyword(s):  
Fracture Mechanic ◽  
Pressure Vessels ◽  
Regulatory Requirements ◽  
Mechanic Analysis ◽  
Term Operation ◽  
Fracture Mechanic Analysis ◽  
Class 1 ◽  
Leak Before Break ◽  
Service Inspection

Based on ASME Boilers and Pressure Vessels Code the major fracture mechanic analysis is limited to protection of class 1 components to brittle fracture. All the Operators of future plants have to enlarge the scope of these analyses to different concepts, at design or operation stage: - brittle and ductile analysis of hypothetical large flaw - leak before break approach - break exclusion concept - incredibility of failure of high integrity components - end of fabrication acceptable defect - in-service inspection performance - acceptable standards in operation - Long Term Operation (LTO) All these requirements needs a procedure, an analysis method with material properties and criteria. After a short overview of each topic, the paper will present how RCC-M, RSE-M French Codes and ASME III and XI take care of all these new modern regulatory requirements.

Effects of Weld Overlays on Leak-Before-Break Margins

2011 ◽  
Author(s):  
G. Angah Miessi ◽  
Peter C. Riccardella ◽  
Peihua Jing
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Operating Conditions ◽  
Intergranular Stress Corrosion ◽  
Pressurized Water ◽  
Weld Overlay ◽  
Water Reactors ◽  
Weld Overlays ◽  
Leak Before Break

Weld overlays have been used to remedy intergranular stress corrosion cracking (IGSCC) in boiling water reactors (BWRs) since the 1980s. Overlays have also been applied in the last few years in pressurized water reactors (PWRs) where primary water stress corrosion cracking (PWSCC) has developed. The weld overlay provides a structural reinforcement with SCC resistant material and favorable residual stresses at the ID of the overlaid component. Leak-before-break (LBB) had been applied to several piping systems in PWRs prior to recognizing the PWSCC susceptibility of Alloy 82/182 welds. The application of the weld overlay changes the geometric configuration of the component and as such, the original LBB evaluation is updated to reflect the new configuration at the susceptible weld. This paper describes a generic leak-before-break (LBB) analysis program which demonstrates that the application of weld overlays always improves LBB margins, relative to un-overlaid, PWSCC susceptible welds when all the other parameters or variables of the analyses (loads, geometry, operating conditions, analysis method, etc…) are kept equal. Analyses are performed using LBB methodology previously approved by the US NRC for weld overlaid components. The analyses are performed for a range of nozzle sizes (from 6″ to 34″) spanning the nominal pipe sizes to which LBB has been commonly applied, using associated representative loads and operating conditions. The analyses are performed for both overlaid and un-overlaid configurations of the same nozzles, and using both fatigue and PWSCC crack morphologies in the leakage rate calculations and the LBB margins are compared to show the benefit of the weld overlays.

Sign in / Sign up

Export Citation Format

Share Document