A Simplified Methodology for Extremely Low-Probability Rupture Prediction for PWR LBB Applications

2011 ◽  
Author(s):  
Heqin Xu ◽  
Ashok Nana ◽  
Samer Mahmoud ◽  
Doug Killian
Keyword(s):  
Conditional Probability ◽  
Assessment Tool ◽  
Safety Margin ◽  
Nuclear Industry ◽  
Leak Rate ◽  
Fluid System ◽  
Design Basis ◽  
Safety Margins ◽  
Low Probability ◽  
The U.S

The leak-before-break (LBB) applicability is stated in General Design Criterion 4 (GDC-4) of Title 10 of the Code of Federal Regulation Part 50 (10 CFR 50). GDC-4 requires that analyses reviewed and approved by the U.S. Nuclear Regulatory Commission (NRC) demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping, in order that dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis. Standard review plan 3.6.3 (SRP-3.6.3) further requires a simultaneous safety margin of two and ten on the flaw size and leak rate detectability, respectively, for deterministic analyses, believing that the very conservative and restrictive safety margins would lead to extremely low probability of fluid system piping rupture. The technology advancements of recent years make it possible to numerically quantify the probability of rupture with confidence. Planned for completion within the next six years, a long-term, large-scale assessment tool, xLPR, is currently being developed by the U.S. NRC, in cooperation with the nuclear industry, to assess the extremely low probability of rupture. The tool will include comprehensive evaluations both before and after through-wall cracks are developed in the degraded components. In this study, we are going to utilize a simplified methodology to investigate the probability of piping rupture for a postulated through-wall crack. The conditional probability, when multiplied by the probability of having a through-wall crack during the life time of plant service, produces an overall probability of piping rupture. The major quantifiable uncertainties, such as the uncertainties associated with the material tensile properties and fracture toughness, and flow-path crack morphology parameters will be modeled as correlated random variables in this paper. Efficient Dimension-Reduction methods will be applied to predict this conditional probability and the results will be compared with the Monte Carlo simulation method. As a sample application of the proposed method, the relationship between the magnitude of the conditional probabilities and the required leak rate detection capability will be established.

A Probability-Based Leak-Before-Break Application for Small Modular Reactors

2015 ◽  
Author(s):  
Heqin Xu ◽  
Tamas Liszkai
Keyword(s):  
Dimension Reduction ◽  
Conditional Probability ◽  
Safety Margin ◽  
Fluid System ◽  
Design Basis ◽  
Small Modular Reactors ◽  
Safety Margins ◽  
Piping Systems ◽  
Main Steam ◽  
Leak Before Break

The leak-before-break (LBB) applicability is stated in General Design Criterion 4 (GDC-4) of Title 10 of the Code of Federal Regulation Part 50 (10 CFR 50). GDC-4 requires that analyses reviewed and approved by the U.S. Nuclear Regulatory Commission (NRC) demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping, in order that dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis. Standard review plan 3.6.3 (SRP-3.6.3) further requires a simultaneous safety margin of two and ten on the flaw size and leak rate detectability, respectively, for deterministic analyses, believing that the very conservative and restrictive safety margins would lead to extremely low probability of fluid system piping rupture. The technology advancements of recent years make it possible to numerically quantify the probability of rupture with confidence. In this study, we are going to utilize a simplified methodology based on the Univariate Dimension-Reduction (UDR) method to investigate the probability of piping rupture for postulated circumferential through-wall cracks in piping systems of small modular reactors (SMRs) inside the metal containment vessel, which typically range from 5” to 8” nominal diameter for feedwater and main steam lines. The unique design of the vacuumed metal CNV in an advanced SMR offers improved leak detection capability, which makes the LBB qualification possible even for piping with much smaller diameters than piping systems in the traditional large reactors, which typically range from 18” to 40” for LBB-qualified piping systems such as surge lines, cold legs, hot legs and main steam lines. The conditional probability that a piping with postulated crack would fail by large break, when multiplied by the probability of having a circumferential through-wall crack during the life time of plant service, produces an overall probability of piping rupture. The major quantifiable uncertainties, such as the uncertainties associated with the material properties, and flow-path crack face morphology parameters will be modeled as correlated random variables in this paper. Efficient Dimension-Reduction methods will be applied to predict the conditional probability. As sample applications of the proposed method, the relationship between the conditional probabilities and the loading factor will be established for SMR piping systems.

Best Estimate Plus Uncertainty Analysis of LBLOCA for a Pickering B CANDU Reactor

2010 ◽  
Author(s):  
Larry Blake ◽  
George Gavrus ◽  
Jack Vecchiarelli ◽  
J. Stoklosa
Keyword(s):  
Power Plants ◽  
Safety Margin ◽  
Reactor Power ◽  
Natural Uranium ◽  
Nuclear Industry ◽  
Initial Increase ◽  
Power Pulse ◽  
Pulse Phase ◽  
Best Estimate ◽  
Safety Margins

The Pickering B Nuclear Generating Station consists of four CANDU reactors. These reactors are horizontal pressure tube, heavy water cooled and moderated reactors fuelled with natural uranium. Under a postulated large break loss of coolant accident (LOCA), positive reactivity results from coolant void formation. The transient is terminated by the operation of the safety systems within approximately 2 seconds of the start of the transient. The initial increase in reactor power, terminated by the action of the safety system, is termed the power pulse phase of the accident. In many instances the severity of an LBLOCA can be characterized by the adiabatic energy deposited to the fuel during this phase of the accident. Historically, Limit of Operating Envelope (LOE) calculations have been used to characterize the severity of the accident. LOE analyses are conservative analyses in which the key operational and safety related parameters are set to conservative or limiting values. Limit based analyses of this type result in calculated transient responses that will differ significantly from the actual expected response of the station. As well, while the results of limit calculations are conservative, safety margins and the degree of conservatism is generally not known. As a result of these factors, the use of Best Estimate Plus Uncertainty (BEPU) analyses in safety analyses for nuclear power plants has been increasing. In Canada, the nuclear industry has been pursuing best estimate analysis through the BEAU (Best Estimate Analysis and Uncertainty) methodology in order to obtain better characterization of the safety margins. This approach is generally consistent with those used internationally. Recently, a BEAU analysis of the Pickering B NGS was completed for the power pulse phase of a postulated Large Break LOCA. The analysis comprised identification of relevant phenomena through a Phenomena Identification and Ranking (PIRT) process, assessment of the code input uncertainties, sensitivity studies to quantify the significance of the input parameters, generation of a functional response surface and its validation, and determination of the safety margin. The results of the analysis clearly demonstrate that the Limit of Operating Envelope (LOE) results are significantly conservative relative to realistic analysis even when uncertainties are considered. In addition, the extensive sensitivity analysis performed to supplement the primary result provides insight into the primary contributors to the results.

scholarly journals Technical Maturity Assessment of Risk-Informed Safety Analysis Tools

2020 ◽  
Author(s):  
Yong-Joon Choi
Keyword(s):  
Best Practice ◽  
Safety Margin ◽  
Assessment Method ◽  
Nuclear Industry ◽  
Analysis Tool ◽  
Program Optimization ◽  
Analysis Tools ◽  
Maturity Assessment ◽  
Safety Margins ◽  
Requirement Traceability

Abstract Ensuring maximum safety while enhancing economic benefit is one of most important goal of In the of US Light Water Reactor Sustainability (LWRS) program. Optimization of the safety margins will provide best practice to achieve this goal which can also lead to cost reduction. Under the LWRS framework, the Risk-Informed Systems Analysis (RISA) Pathway has been focusing on the optimization of safety margin and minimization of uncertainties to ensure both safety and economics at the highest level. One of the important activities of the pathway is to deploy risk-informed analysis tools to related nuclear industry to support precise representation of safety margins and factors that contribute to cost and safety. The tools therefore need highest technical maturity so that industry can use immediately with strong credibility. The tools should have a capability to support risk-informed decision making for both probabilistic and deterministic elements of safety. The RISA Pathway, therefore, have been performing a comprehensive assessment of technical maturity and verification and validation (V&V) status of selected tools to improve adaptability to the industry. The technical maturity assessment includes three work scope: (a) define requirements based on risk-informed concept; (b) investigate and review development and V&V status for technical maturity assessment; and (c) identify technical gap and propose improvement to meet RISA toolkit requirements. The Requirement Traceability Matrix (RTM) concept was used to capture the requirements from user and developer of the project and/or software. The importance of each requirements was evaluated by Phenomena Identification and Ranking Technology (PIRT) which systematically gathers information and ranks the importance of the information. Finally, degree of the maturity was measured by Technology Readiness Level (TRL) for estimating the maturity of the technologies during the development and acquisition phase of certain technology. This paper summarizes development of assessment method and technical evaluation of multi-purpose probabilistic risk analysis tool RAVEN.

Extremely Low Probability of Rupture (xLPR) Version 1.0 Code: Pilot Study Problem Results

2011 ◽  
Author(s):  
D. Rudland ◽  
C. Harrington
Keyword(s):  
Pilot Study ◽  
Assessment Tool ◽  
Lessons Learned ◽  
Nuclear Industry ◽  
Piping System ◽  
Primary System ◽  
Degradation Mechanisms ◽  
System Pressure ◽  
Piping Systems ◽  
Low Probability

Nuclear Regulatory Commission (NRC) Standard Review Plan (SRP) 3.6.3 describes Leak-Before-Break (LBB) assessment procedures that can be used to demonstrate 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 nuclear industry through a memorandum of understanding with the Electric Power Research Institute, is developing a new, modular based, comprehensive piping system assessment methodology to directly demonstrate compliance with the regulations. This project, called xLPR (eXtremely Low Probability of Rupture), will model the effects and uncertainties of relevant active degradation mechanisms and the associated mitigation activities. The resulting analytical tool will be comprehensive, 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 will first focus on the development of a viable method and approach to address the effects of PWSCC as well as define the requirements necessary for a modular-based assessment tool. To meet this goal, the first version of this code has been developed as part of a pilot study, which leverages existing fracture mechanics based models and software coupled to both a commercial and an 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 is meant to demonstrate the feasibility of this code and approach and not to determine the absolute values of the probability of rupture. Later development phases will broaden the scope of xLPR to appropriate primary piping systems in pressurized and boiling water reactors (PWR and BWR), using an incremental approach that incorporates the design requirements and lessons learned from previous iterations. This paper specifically examines the xLPR Version 1.0 model, the methods and approach used to couple the deterministic modules within a probabilistic software framework, and the results from the pilot study. A comparison of the results specific to the surge nozzle sample problem is presented. This paper concludes with lessons learned from the pilot study.

scholarly journals Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Uncertainties in the Estimation of Safety Margins

2008 ◽  
Vol 2008 ◽  
pp. 1-9 ◽  
Author(s):  
Enrico Zio ◽  
Francesco Di Maio
Keyword(s):  
Confidence Interval ◽  
Order Statistics ◽  
Nuclear Reactor ◽  
Safety Margin ◽  
Fuel Cladding ◽  
Complete Group ◽  
Bootstrap Technique ◽  
Safety Margins ◽  
Group Distribution

In the present work, the uncertainties affecting the safety margins estimated from thermal-hydraulic code calculations are captured quantitatively by resorting to the order statistics and the bootstrap technique. The proposed framework of analysis is applied to the estimation of the safety margin, with its confidence interval, of the maximum fuel cladding temperature reached during a complete group distribution blockage scenario in a RBMK-1500 nuclear reactor.

Transient Analysis Needs for Generation IV Reactor Concepts

2002 ◽  
Author(s):  
L. J. Siefken ◽  
E. A. Harvego ◽  
E. W. Coryell ◽  
C. B. Davis
Keyword(s):  
Department Of Energy ◽  
Primary Analysis ◽  
Current Generation ◽  
Light Water ◽  
Design Basis ◽  
Water Reactors ◽  
Ultimate Objective ◽  
New Generation ◽  
Different Characteristics ◽  
The U.S

The importance of nuclear energy as a vital and strategic resource in the U. S. and world’s energy supply mix has led to an initiative, termed Generation IV by the U.S. Department of Energy (DOE), to develop and demonstrate new and improved reactor technologies. These new Generation IV reactor concepts are expected to be substantially improved over the current generation of reactors with respect to economics, safety, proliferation resistance and waste characteristics. Although a number of light water reactor concepts have been proposed as Generation IV candidates, the majority of proposed designs have fundamentally different characteristics than the current generation of commercial LWRs operating in the U.S. and other countries. This paper presents the results of a review of these new reactor technologies and defines the transient analyses required to support the evaluation and future development of the Generation IV concepts. The ultimate objective of this work is to identify and develop new capabilities needed by INEEL to support DOE’s Generation IV initiative. In particular, the focus of this study is on needed extensions or enhancements to SCDAP/RELAP5/3D code. This code and the RELAP5-3D code from which it evolved are the primary analysis tools used by the INEEL and others for the analysis of design-basis and beyond-design-basis accidents in current generation light water reactors.

Methodology to Evaluate Nuclear Power Plant Vulnerability to Events in the Transmission Grid

2014 ◽  
Author(s):  
Alberto Del Rosso ◽  
Jean-François Roy ◽  
Frank Rahn ◽  
Alejandro Capara
Keyword(s):  
Risk Assessment ◽  
Nuclear Power ◽  
Power Plants ◽  
Assessment Tool ◽  
Transmission Grid ◽  
Simulation Tools ◽  
Nuclear Plants ◽  
Western Electricity Coordinating Council ◽  
The Impact ◽  
The U.S

This paper presents a general approach to evaluate the risk of trip or Loss of Off-site Power (LOOP) events in nuclear power plants due to contingencies in the power grid. The proposed methodology is based on the Zone of Vulnerability concept for nuclear plants introduced by EPRI in previous work. The proposed methodology is intended to be part of an integrated probabilistic risk assessment tool that is being developed under ongoing EPRI R&D programs. A detailed analysis of many events occurred in actual nuclear plants has been performed in order to identify, classify and characterize the various vulnerability and type of failures that may affect a nuclear plant. Based the outcome of that analysis, a methodology for evaluating the impact of off-site transmission system events on nuclear plants has been outlined. It includes description of the type of contingencies and conditions that need to be included in the analysis, as well as provisions regarding the simulation tools and models that should be used in each case. The methodology is illustrated in a simplified representation of the Western Electricity Coordinating Council (WECC) system in the U.S.

High-Temperature Oxidation Modeling of New Perspective Accident Tolerant Fuel Claddings

2019 ◽  
Author(s):  
Alexander Vasiliev
Keyword(s):  
High Temperature ◽  
High Temperature Oxidation ◽  
Power Plants ◽  
Renewable Energy Sources ◽  
Nuclear Industry ◽  
Safety Level ◽  
Pressurized Water ◽  
Temperature Oxidation ◽  
Design Basis ◽  
New Perspective

Abstract Currently, the comprehension among the specialists and functionaries is getting stronger that the nuclear industry can encounter serious difficulties in development in the case of insufficiently decisive measures to enhance the safety level of nuclear objects. The keen competition with renewable energy sources like wind, solar or geothermal energy takes place presently and is expected to continue in future decades. One of main measures of nuclear safety enhancement could be the drastic renovation of materials used in nuclear industry. The analytical models of high-temperature oxidation of new perspective materials including chromium-nickel-based alloys, zirconium-based cladding with protective chromium coating, FeCrAl alloys and composite claddings on the basis of SiC/SiC in the course of design-basis and beyond-design-basis accidents at nuclear power plants (NPPs) are developed and implemented to severe accident computer running code. The comparison with available experimental data is conducted. The preliminary calculations of nuclear pressurized water reactor loss-of-coolant accidents with new types of claddings demonstrate encouraging results for hydrogen generation rate and integral hydrogen production. It looks optimistic for considerable upgrade of safety level for future generation NPPs using new fuel and cladding materials.

Crack-Opening Displacement and Leak-Rate Calculations for Full Structural Weld Overlays

2009 ◽  
Author(s):  
D.-J. Shim ◽  
E. Kurth ◽  
F. Brust ◽  
G. Wilkowski ◽  
A. Csontos ◽  
...  
Keyword(s):  
Crack Opening ◽  
Nuclear Industry ◽  
Leak Rate ◽  
Opening Displacement ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Weld Overlay ◽  
Water Reactors ◽  
The Impact ◽  
Weld Overlays

Full structural weld overlays have been used in the U.S. nuclear power industry for over twenty years in boiling water reactors (BWRs). Primary water stress corrosion cracking (PWSCC) in nickel-based dissimilar metal welds (DMWs) has been experienced in pressurized water reactors (PWRs) since the early 1990s. As a result, the nuclear industry is implementing full structural weld overlays (FSWOL) as a PWSCC mitigation technique that may be used on primary coolant lines previously approved for Leak-Before-Break (LBB). This work investigates the effect of the FSWOL on the leakage behavior of these lines with postulated defects. In this paper, finite element (FE) based crack-opening displacements (CODs) were developed for pipes with a FSWOL with postulated complex cracks. The COD solutions were then employed in standard leak-rate calculations, where equivalent crack morphology parameters were developed to consider a flow through two different crack morphologies, i.e., PWSCC through the DMW and corrosion fatigue through the weld overlay. The results of the sensitivity study and a discussion on the impact of the weld overlay on the leakage behavior concludes this paper.

License Renewal and Subsequent License Renewal in the U.S.

2014 ◽  
Author(s):  
Garry G. Young
Keyword(s):  
Nuclear Industry ◽  
Significant Progress ◽  
Safe Operation ◽  
Term Operation ◽  
Nuclear Plants ◽  
The Status ◽  
License Renewal ◽  
The U.S ◽  
Operating Plant

As of February 2014, the NRC has renewed the operating licenses for 73 nuclear units, allowing for up to 60 years of safe operation. In addition, the NRC has license renewal applications under review for 18 units and 9 additional units have announced plans to submit applications over the next few years [1]. This brings the total of renewed licenses and plans for renewal to 100% of the operating nuclear units in the U.S. By the end of 2014, there will be 38 nuclear plants that will have operated for more than 40 years and will be eligible to seek a subsequent license renewal (or almost 40% of the nuclear units expected to be operating at the end of 2014). In 2013, nuclear plant owners of 5 units shutdown operation or announced plans to shutdown by the end of 2014. However, most of the remaining operating plant owners are keeping the option open for long term operation beyond 60 years. NRC and the U.S. nuclear industry have made significant progress in preparing the way for subsequent license renewal applications. This paper presents the status of the U.S. license renewal process and issues being addressed for possible applications for subsequent renewals for up to 80 years of operation.

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