Inspection Optimization Justification for PWR Main Steam and Feedwater Nozzles Using Probabilistic Fracture Mechanics

2019 ◽  
Author(s):  
C. Lohse ◽  
D. J. Shim ◽  
D. Somasundaram ◽  
R. Grizzi ◽  
G. L. Stevens ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Regulatory Commission ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Asme Code ◽  
Sensitivity Studies ◽  
Main Steam ◽  
Regulatory Commission ◽  
Technical Basis

Abstract Pressurized water reactor (PWR) steam generator (SG) main steam and feedwater nozzles are classified as ASME Code, Section XI, Class 2, Category C-B, pressure retaining welds in pressure vessels. Current ASME Code requirements specify that the nozzle-to-shell welds (Item No. C2.21 & C2.32) and nozzle inner radius sections (Item C2.22) are to be examined very 10 years. An evaluation was performed to establish a technical basis for optimized inspection frequencies for these items. The work included a review of inspection history and results, a survey of components in the PWR fleet (which included both U.S. and overseas plants), selection of representative main steam and feedwater nozzle configurations and operating transients for stress analysis, evaluation of potential degradation mechanisms, and flaw tolerance evaluations consisting of probabilistic and deterministic fracture mechanics analyses. The results of multiple inspection scenarios and sensitivity studies were compared to the U.S. Nuclear Regulatory Commission (NRC) safety goal of 10−6 failures per year.

The Inclusion of Inner Surface Breaking Flaws in Probabilistic Fracture Mechanics Analyses of Reactor Vessels Subjected to Planned Normal Cool-Down Transients

2008 ◽  
Author(s):  
Terry Dickson ◽  
Mark EricksonKirk
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Probabilistic Fracture Mechanics ◽  
Inner Surface ◽  
Regulatory Commission ◽  
Technical Basis

The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned reactor startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the ASME Code. The technical basis for these regulations contains many aspects that are now broadly recognized by the technical community as being unnecessarily conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, a goal of current NRC research is to derive a technical basis for a risk-informed revision to the current requirements that reduces the conservatism and also is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). Previous publications have been successful in illustrating potential methods to provide a risk-informed relaxation to the current regulations for normal transients. Thus far, probabilistic fracture mechanics (PFM) analyses have been performed at 60 effective full power years (EFPY) for one of the reactors evaluated as part of the PTS re-evaluation project. In these previous analyses / publications, consistent with the assumptions utilized for this particular reactor in the PTS re-evaluation, all flaws for this reactor were postulated to be embedded. The objective of this paper is to review the analysis results and conclusions from previous publications on this subject and to attempt to modify / generalize these conclusions to include RPVs postulated to contain only inner-surface breaking flaws or a combination of embedded flaws and inner-surface breaking flaws.

Risk-Based Fracture Evaluation of Reactor Vessels Subjected to Cool-Down Transients Associated With Shutdown: An Examination of the Effects of Different Modeling Approaches on Estimated Failure Probabilities

2006 ◽  
Author(s):  
T. L. Dickson ◽  
M. T. EricksonKirk
Keyword(s):  
Nuclear Reactor ◽  
Structural Integrity ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Pressure Vessels ◽  
Asme Code ◽  
Computational Methodology ◽  
Regulatory Commission ◽  
Technical Basis ◽  
Evaluation Project

The current regulations, as set forth by the United States Nuclear Regulatory Commission (USNRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the ASME Code. In 1999, the USNRC initiated the interdisciplinary Pressurized Thermal Shock (PTS) Re-evaluation Project to determine if a technical basis could be established to support a relaxation in the current PTS regulations. The PTS re-evaluation project included the development and application of an updated risk-based computational methodology that incorporates several advancements applicable to modeling the physics of vessel fracture due to thermal hydraulic transients imposed on the RPV inner surface. The results of the PTS re-evaluation project demonstrated that there is a sound technical basis to support a relaxation of the current PTS regulations. The results of the PTS re-evaluation are currently under review by the USNRC. Based on the promising results of the PTS re-evaluation, the USNRC has recently applied the updated computational methodology to fracture evaluations of RPVs subjected to planned cool-down transients, associated with reactor shutdown, derived in accordance with ASME Section XI – Appendix G. The objective of these analyses is to determine if a sound technical basis can be established to provide a relaxation to the current regulations for the derivation of bounding cool-down transients as specified in Appendix G to Section XI of the ASME Code. This paper provides a brief overview of these analyses, results, and the implications of the results.

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.

An Experimental Study on Head Loss of Prototypical Fibrous Debris Beds During Loss-of-Coolant Accident Conditions

2016 ◽  
Vol 2 (3) ◽  
Author(s):  
Amir Ali ◽  
Edward D. Blandford
Keyword(s):  
Safety Issue ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Head Loss ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Debris Accumulation ◽  
Loss Of Coolant Accident ◽  
Accident Conditions ◽  
Regulatory Commission

The United States Nuclear Regulatory Commission (NRC) initiated a generic safety issue (GSI-191) assessing debris accumulation and resultant chemical effects on pressurized water reactor (PWR) sump performance. GSI-191 has been investigated using reduced-scale separate-effects testing and integral-effects testing facilities. These experiments focused on developing a procedure to generate prototypical debris beds that provide stable and reproducible conventional head loss (CHL). These beds also have the ability to filter out chemical precipitates resulting in chemical head loss. The newly developed procedure presented in this paper is used to generate debris beds with different particulate to fiber ratios (η). Results from this experimental investigation show that the prepared beds can provide reproducible CHL for different η in a single and multivertical loops facility within ±7% under the same flow conditions. The measured CHL values are consistent with the predicted values using the NUREG-6224 correlation. Also, the results showed that the prepared debris beds following the proposed procedure are capable of detecting standard aluminum and calcium precipitates, and the head loss increase (chemical head loss) was measured and reported in this paper.

Large Break LOCA Safety Injection Sensitivity for a CE/ABB System 80+ PWR

2002 ◽  
Author(s):  
J. Pottorf ◽  
S. M. Bajorek
Keyword(s):  
Technical Specification ◽  
Nuclear Regulatory Commission ◽  
Injection System ◽  
Engineering System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Core Recovery ◽  
Regulatory Commission ◽  
Saturated Boiling

A WCOBRA/TRAC model of an evolutionary pressurized water reactor with direct vessel injection was constructed using publicly available information and a hypothetical double-ended guillotine break of a cold leg pipe was simulated. The model is an approximation of a ABB/Combustion Engineering System 80+ pressurized water reactor (PWR). WCOBRA/TRAC is the thermal-hydraulics code approved by the U.S. Nuclear Regulatory Commission for use in realistic large break LOCA analyses of Westinghouse 3- and 4-loop PWRs and the AP600 passive design. The AP600 design uses direct vessel injection, and the applicability of WCOBRA/TRAC to such designs is supported by comparisons to appropriate test data. This study extends the application of WCOBRA/TRAC to the investigation of the predicted behavior of direct vessel injection in an evolutionary design. A series of large break LOCA simulations were performed assuming a core power of 3914 MWt, and Technical Specification limits of 2.5 on total peaking factor and 1.7 on hot channel enthalpy rise factor. Two cladding temperature peaks were predicted during reflood, one following bottom of core recovery and a second caused by saturated boiling of water in the downcomer. This behavior is consistent with prior WCOBRA/TRAC calculations for some Westinghouse PWRs. The simulation results are described, and the sensitivity to failure assumptions for the safety injection system is presented.

Uncertainties in Pressurized Thermal Shock Analyses

2019 ◽  
Author(s):  
M. Niffenegger ◽  
O. Costa Garrido ◽  
D. F. Mora ◽  
G. Qian ◽  
R. Mukin ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Ansys Fluent ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Safety Margins ◽  
Pressurized Thermal Shock

Abstract Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.

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.

Fracture mechanics in design and service: ‘living with defects’ - Fracture mechanics as an aid to design and operation of nuclear plant

1981 ◽  
Vol 299 (1446) ◽  
pp. 227-236 ◽  
Keyword(s):  
Fracture Mechanics ◽  
Material Properties ◽  
Pressure Vessels ◽  
Nuclear Plant ◽  
Plant Design ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Plant Safety ◽  
Usual Criterion ◽  
Design Protection

Fracture mechanics analyses are an important part of nuclear plant design, supplementing the conventional design protection against failure to cover the possibility of the presence of crack-like defects. The degree of detail and accuracy required for a particular application depends on the possible consequences of a failure and whether the assessment is concerned with plant safety or with aspects of reliability. In the former case, a conservative approach is necessary and the prevention of initiation is the usual criterion. This approach is typified by the safety assessment applied to pressurized water reactor pressure vessels, which is outlined and discussed in relation to elastic plastic approaches and the importance of plant transient conditions, material properties (especially in weldments) and possible defect distributions. Fracture mechanics can help in defining quality control and quality assurance procedures, including both requirements for mechanical property appraisal and nondestructive testing. The latter aspects extend into operation, in respect of monitoring of plant conditions, surveillance of changes in material properties and the use of periodic inspection and plant condition monitoring techniques. A number of examples are quoted and recommendations made to permit improved fracture mechanics assessments.

Use of Risk Insights to Enhance Safety Focus of Small Modular Reactor Reviews

2011 ◽  
Author(s):  
Stewart L. Magruder
Keyword(s):  
Review Process ◽  
Nuclear Regulatory Commission ◽  
Pressurized Water Reactor ◽  
Design Features ◽  
Pressurized Water ◽  
Analysis And Evaluation ◽  
Small Modular Reactor ◽  
Efficiency And Effectiveness ◽  
Regulatory Commission ◽  
Regulatory Controls

The U.S. Nuclear Regulatory Commission staff plans to apply a more integrated, graded approach to the review of small modular reactor (SMR) pre-application activities and design applications. The concept is to improve the efficiency and effectiveness of the reviews by focusing on safety significant structures, systems, and components (SSCs). The unique design features associated with SMRs and knowledge gained reviewing other passive reactor designs present opportunities to risk-inform the SMR design certification process to a greater extent than previously employed. The review process can be modified for SMR applications by considering the aggregate of regulatory controls pertaining to SSCs as part of the review and determining those regulatory controls which may supplement or replace, as appropriate, part of the technical or engineering analysis and evaluation. Risk insights acquired from staff reviews of passive LWR designs (i.e. AP1000, ESBWR) can also be incorporated into the review process. Further, risk insights associated with integral pressurized water reactor (iPWR) design features (i.e. Underground facilities impact on turbine missiles review) can be incorporated into the review process.

Probabilistic Fracture Mechanics Evaluation of a BWR Feedwater Nozzle

2021 ◽  
Author(s):  
Kevin K. L. Wong ◽  
Garivalde Dominguez ◽  
Do Jun Shim ◽  
Steven K. Richter
Keyword(s):  
Fracture Mechanics ◽  
Element Model ◽  
Nuclear Regulatory Commission ◽  
Boiling Water ◽  
Normal Operation ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Probabilistic Fracture Mechanics ◽  
Asme Code ◽  
Technical Basis

Abstract A probabilistic fracture mechanics (PFM) evaluation was performed for the nozzle blend radius and nozzle-to-shell weld of a boiling water reactor (BWR) feedwater nozzle using the PFM methodology in Electric Power Research Institute (EPRI) Boiling Water Reactor Vessel and Internals Program (BWRVIP) BWRVIP-108-A and BWRVIP-241-A, which are the technical basis for inspection relief in ASME Code Case N-702. Using a finite element model of the feedwater nozzle, stress analysis was performed for plant-specific piping loads and bounding transients, which were grouped by severity and projected cycle counts. Monte Carlo simulations were performed using the VIPER-NOZ (Vessel Inspection Program Evaluation for Reliability, including Nozzle) PFM software to determine probabilities of failure for the reactor pressure vessel (RPV) with an inspection population of 25% of the feedwater nozzles every ten years for sixty years of plant operation. The results show that the probabilities of failure for normal operation and low temperature over pressure (LTOP) event meet the acceptance criteria for RPV failure in NUREG-1806 by the U.S. Nuclear Regulatory Commission (NRC). Thus, there is potential to seek regulatory relief to reduce the inspection population of BWR feedwater nozzles from 100% to 25% every ten years using the technical basis of ASME Code Case N-702.

Sign in / Sign up

Export Citation Format

Share Document