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.

Benchmarking Probabilistic Code for LBB Analysis for Circumferential Cracks

2017 ◽  
Author(s):  
Robert E. Kurth ◽  
Cédric J. Sallaberry ◽  
Bruce A. Young ◽  
Paul Scott ◽  
Frederick W. Brust ◽  
...  
Keyword(s):  
Probability Of Failure ◽  
Primary System ◽  
Degradation Mechanisms ◽  
Detection Systems ◽  
System Pressure ◽  
Piping Systems ◽  
Primary Water ◽  
Low Probability ◽  
Assessment Procedures ◽  
Leak Before Break

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. There are several codes available for addressing the requirements of GDC-4. This paper addresses three of these codes: (1) xLPR 2.0; (2) PROLOCA; and (3) PROMETHEUS. Each of these codes is described and applied to a representative plant where active degradation mechanisms have been found. Conclusions about the design, results, and interpretation of the results is then provided. In all cases the probability of failure of the pipe is found to be extremely low when the crack inspections and leak detection systems are modeled.

Leak-Before-Break Under Beyond Design Basis Seismic Loading

2013 ◽  
Vol 135 (5) ◽  
Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Seismic Event ◽  
Jet Impingement ◽  
Seismic Loading ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Plant Design ◽  
Piping Systems ◽  
Regulatory Commission ◽  
Leak Before Break

The Atucha II nuclear power plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant design was by Kraftwerk Union (KWU) in the 1970's using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The United States Nuclear Regulatory Commission (US NRC) developed leak-before-break (LBB) procedures in standard review plan (SRP) 3.6.3 Revision 1 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987. The US NRC is currently developing a draft Regulatory Guide on what is called the transition break size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work (Wilkowski et al., “Robust LBB Analysis for Atucha II Nuclear Plant,” 2011 ASME PVP Conference, July 17–21, Baltimore, MD) on the seismic evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1 × 10−6 per year, that a double-ended guillotine break (DEGB) was pragmatically impossible due to the high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94% of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33% around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g's for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1 × 10−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.

Inservice Testing Program Improvements for New Reactors Small Modular Reactors (SMRs)

2014 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Small Modular Reactors ◽  
American Society ◽  
Construction Permit ◽  
Regulatory Commission

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.

TITRAM Fast Transient Licensing Analysis Methodology for Kuosheng BWR/6

2010 ◽  
Author(s):  
Jan-Ru Tang ◽  
Hon-Chin Jien ◽  
Yang-Kai Chiu ◽  
Cheng-Der Wang ◽  
Julian S. C. Chian
Keyword(s):  
Nuclear Power ◽  
Transient Analysis ◽  
Nuclear Regulatory Commission ◽  
Nuclear Industry ◽  
Evaluation Methodology ◽  
Uncertainty Evaluation ◽  
Power Company ◽  
Fast Transients ◽  
Fast Transient ◽  
Regulatory Commission

This paper presents the TITRAM (TPC/INER Transient Analysis Method) methodology for the fast transient analysis of Kuosheng Nuclear Power Station (KSNPS) with two units of General Electric (GE) designed BWR/6 (Boiling Water Reactor). The purpose of this work is to provide a technical basis of Taiwan Power Company (TPC)/Institute of Nuclear Energy Research (INER)’s qualification to perform plant specific licensing safety analyses for the Final Safety Analysis Report (FSAR) basis system fast transients, and related plant operational transient analyses for the Kuosheng plant. The major task of qualifying TITRAM as a licensing method for BWR transient analysis is to adequately quantify its analysis uncertainty. A similar approach as the CSAU (Code Scaling, Applicability, and Uncertainty Evaluation) methodology developed by the USNRC (United States Nuclear Regulatory Commission) was adopted. The CSAU methodology could be characterized as three significant processes, namely code applicability, transient scenario specification and uncertainty evaluation based on Phenomena Identification and Ranking. The applicability of the TITRAM code package primarily using the SIMULATE-3 and RETRAN-3D codes are demonstrated with analyses of integral plant tests such as Peach Bottom Turbine Trip Test and plant startup tests of KSNPS. A Phenomena Identification and Ranking Table (PIRT) with uncertainty values for each identified parameter to cover 95% of possible values are established for the selected KSNPS fast transients. The experience from BWR organizations in the nuclear industry is used as a guide in construction of the PIRT. Sensitivity studies and associated statistical analyses are performed to determine the overall uncertainty of fast transient analysis with TITRAM based on the KSNPS Analysis Nominal Model. Finally, the Licensing Model is established for future licensing applications.

Relating ASME Flaw Evaluation to ASME Task Group on Carbon Fiber Repairs

2017 ◽  
Author(s):  
Tomas Jimenez ◽  
Eric Houston ◽  
Nico Meyer
Keyword(s):  
Carbon Fiber ◽  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Fiber Reinforced Polymers ◽  
Structural Factors ◽  
Operating Life ◽  
The Us ◽  
Piping Systems ◽  
Wave Passage ◽  
Regulatory Commission

As most nuclear power stations in the US have surpassed their initial 40 years of operability, the industry is now challenged with maintaining safe operations and extending the operating life of structures, systems and components. The US Nuclear Regulatory Commission (NRC), Nuclear Energy Institute (NEI), and Electric Power Research Institute (EPRI) have identified safety related buried piping systems as particularly susceptible to degradation. These systems are required to maintain the structural factors of the ASME Construction Codes under pressure and piping loads, which includes seismic wave passage. This paper focuses on evaluation approaches for metallic buried piping that can be used to demonstrate that localized thinning meets the requirements of the Construction Code. The paper then addresses a non-metallic repair option using carbon fiber reinforced polymers (CFRP) as the new pressure boundary.

Regulatory Experience of Leak-Before-Break (LBB) Technology to the High Energy Piping Systems in Korea

2010 ◽  
Author(s):  
Yeon-Ki Chung ◽  
Jin-Su Kim ◽  
Hae-Dong Chung ◽  
Young-Hwan Choi
Keyword(s):  
Material Properties ◽  
Operating Experience ◽  
Leak Detection ◽  
High Energy ◽  
Leak Rate ◽  
Regulatory Review ◽  
Screening Criteria ◽  
Piping Systems ◽  
Crack Stability ◽  
Leak Before Break

The application of the leak-before-break (LBB) technology to the newly constructed pressurized water reactors (PWRs) has been approved for the several high energy piping systems inside containment in Korea. The main purpose of the LBB application for these systems at the design stage is the removal of the dynamic effects associated with the postulated double-ended guillotine break (DEGB) from design basis loads, as well as to the elimination of the pipe whip restraints and jet impingement barriers so as to increase the access the inspections. LBB technology is based on the low probability of pipe ruptures in the candidate piping systems using fracture mechanics and the insights from the state-of-the-art technology including operating experience. The procedures for LBB application is fundamentally based on the Unite States Nuclear Regulatory Commission (US NRC) requirements as detailed in the standard review plan (SRP) 3.6.3. However, a number of the additional requirements and issues are not specified in the review procedure during regulatory review were imposed and addressed during the review process. The regulatory review is focused on the confirmation on the methods for the elements in the screening criteria and several technical concerns on the determination of material properties, the validation of crack evaluation methods and leak rate estimation in the LBB evaluation considering the adequate margin. Although the application of the LBB has been approved by the safety authority for some high energy systems, the validation of LBB is continuously maintained in consideration of operating experience. In this paper, the regulatory positions for LBB application are described for the areas of screening criteria, leak rate estimation including the capability of leak detection system, material properties, load combination, crack stability methods, and margins in the crack stability evaluations. The issues encountered during the regulatory review such as the dynamic fracture test to consider the dynamic strain aging (DSA) of carbon and low alloy steel, thermal stratification and striping in the pressurizer surge line, water/steam hammer in main steam lines, and estimation of the crack opening area at the pipe-to-nozzle interface considering the asymmetry are also introduced. In addition, several regulatory actions to improve the reliability in the capability of leak detection systems and to clarify the screening criteria such as the corrosion resistance is provided.

Derivation of the New Pressurized Thermal Shock Screening Criteria

2011 ◽  
Author(s):  
Terry Dickson ◽  
Shengjun Yin ◽  
Mark Kirk ◽  
Hsuing-Wei Chou
Keyword(s):  
United States ◽  
Fracture Mechanics ◽  
Thermal Shock ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Screening Criteria ◽  
Pressurized Thermal Shock ◽  
Regulatory Commission ◽  
Technical Basis

As a result of a multi-year, multi-disciplinary effort on the part of the United States Nuclear Regulatory Commission (USNRC), its contractors, and the nuclear industry, a technical basis has been established to support a risk-informed revision to pressurized thermal shock (PTS) regulations originally promulgated in the mid-1980s. The revised regulations provide alternative (optional) reference-temperature (RT)-based screening criteria, which is codified in 10 CFR 50.61(a). How the revised screening criteria were determined from the results of the probabilistic fracture mechanics (PFM) analyses will be discussed in this paper.

LBB Under Beyond Design Basis Seismic Loading

2012 ◽  
Author(s):  
Tao Zhang ◽  
Frederick W. Brust ◽  
Gery Wilkowski ◽  
Heqin Xu ◽  
Alfredo A. Betervide ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Seismic Event ◽  
Jet Impingement ◽  
Seismic Loading ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Plant Design ◽  
Piping Systems ◽  
Heavy Water Reactor ◽  
Regulatory Commission

The Atucha II nuclear power plant is a unique pressurized heavy water reactor (PHWR) being constructed in Argentina. The original plant design was by Kraftwerk Union (KWU) in the 1970’s using the German methodology of break preclusion. The plant construction was halted for several decades, but a recent need for power was the driver for restarting the construction. The US NRC developed leak-before-break (LBB) procedures in draft Standard Review Plan (SRP) 3.6.3 for the purpose of eliminating the need to design for dynamic effects that allowed the elimination of pipe whip restraints and jet impingement shields. This SRP was originally written in 1987 with a modest revision in 2005. The United States Nuclear Regulatory Commission (US NRC) is currently developing a draft Regulatory Guide on what is called the Transition Break Size (TBS). However, modeling crack pipe response in large complex primary piping systems under seismic loading is a difficult analysis challenge due to many factors. The initial published work on the seismic evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to the amplitudes for an event with a probability of 1e−6 per year, that a Double-Ended Guillotine Break (DEGB) was pragmatically impossible due to the incredibly high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size in that case was 94-percent of the circumference. This paper discusses further efforts to show how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33 percent around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g’s for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 1e−6 seismic event accelerations, or 240 times higher than the safe shutdown earthquake (SSE) accelerations.

Lessons Learned From the Application and Refinement of Risk-Informed Inservice Inspection Programs

2008 ◽  
Author(s):  
Alan D. Chockie ◽  
M. Robin Graybeal ◽  
Scott D. Kulat
Keyword(s):  
Cost Effective ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Nuclear Industry ◽  
Plant Safety ◽  
Nuclear Plants ◽  
New Information ◽  
Effective Manner ◽  
Regulatory Commission ◽  
Available Information

The risk-informed inservice inspection (RI-ISI) process provides a structured and systematic framework for allocating inspection resources in a cost-effective manner while improving plant safety. It helps focus inspections where failure mechanisms are likely to be and where enhanced inspections are warranted. To date, over eighty-five percent of US nuclear plants and a number of non-US plants have implemented, or are in the process of implementing, RI-ISI programs. Many are already involved in the periodic update of their RI-ISI program. The development of RI-ISI methodologies in the US has been a long and involved process. The risk-informed procedures and rules were developed to take full advantage of PRA data, industry and plant experiences, information on specific damage mechanisms, and other available information. An important feature of the risk-informed methodologies is the requirement to make modifications and improvements to the plant’s RI-ISI application as new information and insights become available. The nuclear industry, ASME Section XI, and the Nuclear Regulatory Commission have all worked together to take advantage of the lessons learned over the years to refine and expand the use of risk-informed methodologies. This paper examines the lessons learned and the benefits received from the application and refinement of risk-informed inservice inspection programs. Also included in the paper is a review of how the information and insights have been used to improve the risk-informed methodologies.

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.

Sign in / Sign up

Export Citation Format

Share Document