Estimating Loss-of-Coolant 

Accident (LOCA) 

Frequencies Through the 

Elicitation Process 

Appendices A through M

Office of Nuclear Regulatory Research 

NUREG-1829 

Vol. 2
AVAILABILITY OF REFERENCE MATERIALS

IN NRC PUBLICATIONS

NRC Reference Material

As of November 1999, you may electronically access

NUREG-series publications and other NRC records at

NRC’s Public Electronic Reading Room at

http://www.nrc.gov/reading-rm.html.

Publicly released records include, to name a few,

NUREG-series publications; Federal Register notices;

applicant, licensee, and vendor documents and

correspondence; NRC correspondence and internal

memoranda; bulletins and information notices;

inspection and investigative reports; licensee event

reports; and Commission papers and their attachments.

NRC publications in the NUREG series, NRC

regulations, and Title 10, Energy, in the Code of

Federal Regulations may also be purchased from one

of these two sources.

1. The Superintendent of Documents

 U.S. Government Printing Office

 Mail Stop SSOP

 Washington, DC 20402–0001

 Internet: bookstore.gpo.gov

 Telephone: 202-512-1800

 Fax: 202-512-2250

2. The National Technical Information Service

 Springfield, VA 22161–0002

 www.ntis.gov 

 1–800–553–6847 or, locally, 703–605–6000

A single copy of each NRC draft report for comment is

available free, to the extent of supply, upon written

request as follows:

Address: Office of Administration, 

 Printing and Mail Services Branch

 U.S. Nuclear Regulatory Commission

 Washington, DC 20555-0001

E-mail: DISTRIBUTION@nrc.gov 

Facsimile: 301–415–2289 

Some publications in the NUREG series that are 

posted at NRC’s Web site address

http://www.nrc.gov/reading-rm/doc-collections/nuregs

are updated periodically and may differ from the last

printed version. Although references to material found

on a Web site bear the date the material was accessed,

the material available on the date cited may

subsequently be removed from the site.

Non-NRC Reference Material

Documents available from public and special technical

libraries include all open literature items, such as

books, journal articles, and transactions, Federal

Register notices, Federal and State legislation, and

congressional reports. Such documents as theses,

dissertations, foreign reports and translations, and

non-NRC conference proceedings may be purchased

from their sponsoring organization.

Copies of industry codes and standards used in a

substantive manner in the NRC regulatory process are

maintained at—

The NRC Technical Library 

Two White Flint North

11545 Rockville Pike

Rockville, MD 20852–2738

These standards are available in the library for

reference use by the public. Codes and standards are

usually copyrighted and may be purchased from the

originating organization or, if they are American

National Standards, from—

American National Standards Institute

11 West 42nd Street

New York, NY 10036–8002

www.ansi.org 

212–642–4900

Legally binding regulatory requirements are stated

only in laws; NRC regulations; licenses, including

technical specifications; or orders, not in 

NUREG-series publications. The views expressed

in contractor-prepared publications in this series are

not necessarily those of the NRC.

The NUREG series comprises (1) technical and

administrative reports and books prepared by the

staff (NUREG–XXXX) or agency contractors

(NUREG/CR–XXXX), (2) proceedings of

conferences (NUREG/CP–XXXX), (3) reports

resulting from international agreements

(NUREG/IA–XXXX), (4) brochures

(NUREG/BR–XXXX), and (5) compilations of legal

decisions and orders of the Commission and Atomic

and Safety Licensing Boards and of Directors’

decisions under Section 2.206 of NRC’s regulations

(NUREG–0750).
Estimating Loss-of-Coolant 

Accident (LOCA) 

Frequencies Through the 

Elicitation Process 

Appendices A through M 

Manuscript Completed: March 2008 

Date Published: April 2008 

Prepared by 

R. Tregoning (NRC), L. Abramson (NRC) 

P. Scott (Battelle-Columbus) 

A. Csontos, NRC Project Manager 

Office of Nuclear Regulatory Research 

NUREG-1829 

Vol. 2 

iii

ABSTRACT 

The NRC is establishing a risk-informed revision of the design-basis pipe break size requirements in 10 

CFR 50.46, Appendix K to Part 50, and GDC 35 which requires estimates of LOCA frequencies as a 

function of break size. Separate BWR and PWR piping and non-piping passive system LOCA frequency 

estimates were developed as a function of effective break size and operating time through the end of the 

plant license-renewal period. The estimates were based on an expert elicitation process which 

consolidated operating experience and insights from probabilistic fracture mechanics studies with 

knowledge of plant design, operation, and material performance. The elicitation required each member of 

an expert panel to qualitatively and quantitatively assess important LOCA contributing factors and 

quantify their uncertainty. The quantitative responses were combined to develop BWR and PWR total 

LOCA frequency estimates for each contributing panelist. The distributions for the six LOCA size 

categories and three time periods evaluated are represented by four parameters (mean, median, 5th and 

95th percentiles). Finally, the individual estimates were aggregated to obtain group estimates, along with 

measures of panel diversity. 

There is general qualitative agreement among the panelists about important technical issues and LOCA 

contributing factors, but the individual quantitative estimates are much more variable. Sensitivity studies 

were conducted to examine the effects on the estimated parameters of distribution shape, correlation 

structure, panelist overconfidence, panel diversity measure, and aggregation method. The group estimates 

are most sensitive to the method used to aggregate the individual estimates. Geometric-mean aggregation 

produces frequency estimates that approximate the medians of the panelists’ estimates and also are 

generally consistent with both operating experience and prior LOCA frequency estimates, except where 

increases are supported by specific material aging-related concerns. However, arithmetic-mean and 

mixture-distribution aggregation are alternative methods that lead to significantly higher mean and 95th

percentile group estimates. Because the results are sensitive to the aggregation method, a particular set of 

LOCA frequency estimates is not generically recommended for all risk-informed applications. 

v

FOREWORD

Estimated frequencies of loss-of-coolant accidents (LOCAs; i.e., pipe ruptures as a function of break size) 

are used in a variety of regulatory applications, including probabilistic risk assessment (PRA) of nuclear 

power plants. Currently, the U.S. Nuclear Regulatory Commission (NRC) is using such information 

to establish a risk-informed alternative to the emergency core cooling system (ECCS) requirements 

in Title 10, Section 50.46, of the Code of Federal Regulations (10 CFR 50.46). Current requirements 

consider pipe breaks in the reactor coolant pressure boundary, up to and including breaks equivalent in size 

to the double-ended rupture of the largest pipe in the reactor coolant system. One aspect of this 

risk-informing activity is to evaluate the technical adequacy of redefining the design-basis break size 

(the largest pipe break to which 10 CFR 50.46 applies) to a smaller size that is consistent with updated 

estimates of pipe break frequencies. 

To provide the technical basis for a risk-informed definition of the design-basis break size, this study 

developed LOCA frequency estimates using an expert elicitation process. This process consolidated 

operating experience and insights from probabilistic fracture mechanics studies with knowledge 

of plant design, operation, and material performance. Expert elicitation is a well-recognized technique 

for quantifying phenomenological knowledge when modeling approaches or data are insufficient. 

The results from the expert elicitation provide LOCA frequency estimates for piping and non-piping 

passive systems, as a function of effective break size and operating time through the end of the plant licenserenewal period, for both boiling- and pressurized-water reactors (BWRs and PWRs, respectively). The 

panelists generally agreed on the important technical issues and LOCA-contributing factors. However, as 

expected, the panelists’ estimates exhibit both significant uncertainty and diversity. The uncertainty is reflected 

in the estimated parameters (mean, median, 5th and 95th percentiles) of the individual LOCA frequency 

distributions, and the diversity is captured by the confidence bounds on the group estimates. In addition, 

this study considered the sensitivity of the results to various analysis approaches. The results are most 

sensitive to the method used to aggregate the individual panelists’ estimates to obtain group estimates. In this 

study, geometric-mean aggregation produces group frequency estimates that approximate the medians of the 

panelists’ estimates and are also generally consistent with both operating experience and prior LOCA 

frequency estimates except where increases are supported by specific material aging-related concerns. 

However, arithmetic-mean and mixture-distribution aggregation are alternative methods that lead to 

significantly higher mean and 95th percentile group estimates. 

Because the alternative aggregation methods can lead to significantly different results, a particular set of 

LOCA frequency estimates is not recommended for all risk-informed applications. The purposes and 

context of the application must be considered when determining the appropriateness of any set of 

elicitation results. In particular, during the selection of the BWR and PWR transition break sizes for the 

proposed 10 CFR50.46a rulemaking, the NRC staff considered the totality of the results from the 

sensitivity studies, rather than only the summary frequency estimates from this study. The NRC anticipates 

that a similar approach will be used in selecting appropriate replacement frequencies for the estimates provided 

in NUREG/CR-5750, “Rates of Initiating Events at U.S. Nuclear Power Plants: 1987 – 1995,” and for other 

applications that require frequencies for break sizes other than those in NUREG/CR-5750. 

Jennifer L. Uhle, Director 

Division of Engineering 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

vii

TABLE OF CONTENTS 

ABSTRACT................................................................................................................................... iii 

FOREWORD ...................................................................................................................................v 

EXECUTIVE SUMMARY............................................................................................................xv 

ACKNOWLEDGMENTS............................................................................................................xxv 

NOMENCLATURE...................................................................................................................xxvii 

1. BACKGROUND...................................................................................................................1-1 

 1.1 Motivation ..........................................................................................................................1-1 

 1.2 Previous Approaches for Estimating LOCA Frequencies..................................................1-2 

 1.3 Modeling Mitigation Effects ..............................................................................................1-8 

 1.4 Other LOCA Initiating Events ...........................................................................................1-9 

 1.5 Expert Elicitation..............................................................................................................1-10 

 1.6 References ........................................................................................................................1-11 

2. OBJECTIVE AND SCOPE...................................................................................................2-1 

 2.1 References ..........................................................................................................................2-3 

3. ELICITATION APPROACH................................................................................................3-1 

 3.1 Pilot Elicitation...................................................................................................................3-1 

 3.2 Panel Selection ...................................................................................................................3-4 

 3.3 Elicitation Training ............................................................................................................3-5 

 3.4 Technical Issue Formulation ..............................................................................................3-8 

 3.5 Base Case Development...................................................................................................3-17 

 3.6 Elicitation Background Information.................................................................................3-25 

 3.7 Flow Rate Correlation ......................................................................................................3-26 

 3.8 Elicitation Question Development ...................................................................................3-27 

 3.9 Individual Elicitations ......................................................................................................3-31 

 3.10 Final Elicitation Responses ............................................................................................3-32 

 3.11 Review and Reporting....................................................................................................3-34 

 3.12 References ......................................................................................................................3-35 

4. BASE CASE RESULTS .......................................................................................................4-1 

 4.1 Piping Base Case Through-Wall Cracking Frequencies ....................................................4-1 

 4.2 Piping Base Case LOCA Frequencies................................................................................4-2 

 4.3 Piping Base Case Sensitivity Studies ...............................................................................4-12 

 4.4 Non-Piping Base Case Results.........................................................................................4-25 

 4.5 References ........................................................................................................................4-31 
viii

 5. ANALYSIS OF ELICITATION RESPONSES...................................................................5-1 

 5.1 Analysis Framework....................................................................................5-1 

 5.2 Elicitation Responses .........................................................................................................5-4 

 5.3 Individual Estimates...........................................................................................................5-6 

 5.4 Group Estimates and Confidence Intervals ......................................................................5-14 

 5.5 Baseline LOCA Frequency Estimates..............................................................................5-16 

 5.6 Sensitivity Analyses .........................................................................................................5-18 

 5.7 References ........................................................................................................................5-26 

6. QUALITATIVE RESULTS AND DISCUSSION................................................................6-1 

 6.1 Base Case Evaluations .......................................................................................................6-1 

 6.2 Safety Culture Effects ........................................................................................................6-2 

 6.3 Passive System Failure Insights .........................................................................................6-5 

 6.4 References ........................................................................................................................6-13 

7. QUANTITATIVE RESULTS...............................................................................................7-1 

 7.1 Safety Culture.....................................................................................................................7-2 

 7.2 Total BWR and PWR LOCA Frequencies (Baseline Estimates).......................................7-5 

 7.3 Comparison of Piping with Non-Piping Contributions to Total 

 LOCA Frequencies...........................................................................................................7-8 

 7.4 Effect of Plant Operating Time on Total LOCA Frequencies..........................................7-10 

 7.5 Individual Uncertainty and Panel Diversity .....................................................................7-12 

 7.6 Sensitivity Analyses .........................................................................................................7-18 

 7.7 Summary Estimates..........................................................................................................7-50 

 7.8 Steam Generator Tube Rupture and Small Break (Category 1) LOCA Frequencies.......7-52 

 7.9 Comparison with Prior Estimates.....................................................................................7-55 

7.10 Comparison with Operating Experience Data.................................................................7-59 

7.11 References.......................................................................................................................7-62 

8. BASIS FOR FUTURE LOCA FREQUENCY UPDATES..................................................8-1 

9. SUMMARY OF RESULTS AND CONCLUSIONS ..........................................................9-1 

LIST OF APPENDICES 

APPENDIX A: PANEL MEMBER QUALIFICATIONS..........................................................A-1 

APPENDIX B: MEETING MINUTES FROM GROUP PANEL MEETINGS.........................B-1 

APPENDIX C: ELICITATION TRAINING EXERCISE RESULTS........................................C-1 

APPENDIX D: PIPING BASE CASE RESULTS OF BENGT LYDELL.................................D-1 
ix

APPENDIX E: PIPING BASE CASE RESULTS OF WILLIAM GALYEAN......................... E-1 

APPENDIX F: PIPING BASE CASE RESULTS OF DAVID HARRIS................................... F-1 

APPENDIX G: PIPING BASE CASE RESULTS OF VIC CHAPMAN...................................G-1 

APPENDIX H: DESCRIPTION OF NON-PIPING DATABASE.............................................H-1 

APPENDIX I: REACTOR VESSEL LOCA PROBABILITY BASE CASE ANALYSES (BWR 

 VESSELS AND PWR TOP HEAD NOZZLES).............................................I-1 

APPENDIX J: ELICITATION QUESTIONS ............................................................................J-1 

APPENDIX K: GENERAL APPROACH AND PHILOSOPHY OF EACH PANEL 

 MEMBER.......................................................................................................K-1 

APPENDIX L: DETAILED RESULTS...................................................................................... L-1 

APPENDIX M: PUBLIC COMMENT RESPONSES...............................................................M-1 

LIST OF FIGURES 

Figure 1 BWR and PWR Error Factor Corrected LOCA Frequency Estimates................ xxiii 

Figure 1.1 Sensitivity of Conditional Failure Probability to Failure Mode Assumption.........1-8 

Figure 1.2 IGSCC Events By Year..........................................................................................1-9 

Figure 3.1 Flowchart of the Overall Elicitation Process..........................................................3-2 

Figure 3.2 Elicitation Structure..............................................................................................3-11 

Figure 4.1 Leak Frequency Comparison Between Operating-Experience-Based....................4-2 

 and PFM-Based Analyses 

Figure 4.2 Category 1 Piping Base Case Frequencies at 25 Years................................4-7 

Figure 4.3 Category 3 Piping Base Case Frequencies at 25 Years..........................................4-8 

Figure 4.4 PWR Piping Base Case Frequencies at 25 Years of Plant Operation...................4-10 

Figure 4.5 BWR Piping Base Case Frequencies at 25 Years of Plant Operation ..................4-10 

Figure 4.6 BWR-1 LOCA Frequencies as a Function of Plant Operating Time ...................4-13 

 for the Case of No ISI with Different Leak Detection Strategies 

Figure 4.7 BWR-1 LOCA Frequencies as a Function of Plant Operating Time ...................4-14 

 for the Case of ISI with 90% POD with Different Leak Detection Strategies 

Figure 4.8 Probability of Detection (POD) Curve Used by Vic Chapman............................4-16 

 in ISI Sensitivity Analysis 

Figure 4.9 Graphical Results from Vic Chapman’s ISI Sensitivity Analysis........................4-16 

Figure 4.10 Through-Wall Residual Axial Stress Distribution................................................4-18 

 Used in Weld Overlay Analyses 

Figure 4.11 Effect of Weld Overlay Repairs on the Cumulative Category 1 ..........................4-18 

 LOCA Frequencies for 12-inch Diameter BWR Recirculation System Piping 

Figure 4.12 Pre and Post-Mitigation Recirculation LOCA Frequencies .................................4-19 
x

Figure 4.13 PRAISE Analysis Showing Effect of Normal Operating.....................................4-21 

 Stress on Various Size Precursor Leak and LOCA 

 Frequencies for a Single Weld Joint in 12-inch Diameter 

 Section of Recirculation System 

Figure 4.14 Effect of a Hydro Test on the Cumulative LOCA Probabilities...........................4-23 

 for the Hot Leg with a SSE Load Included 

Figure 4.15 Effect of a SSE Load on the Cumulative LOCA Probabilities.............................4-23

 for the Hot Leg Without a Hydro Test 

Figure 4.16 Effect of Material Properties (Fracture Toughness) on the Cumulative...............4-25 

 LOCA Probabilities for the PWR-1 Base Case 

Figure 4.17 Steam Generator Tube Leaks as a Function of Event Date..................................4-26 

Figure 4.18 PWR CRDM LOCA Frequencies at 25, 40, and 60 years....................................4-30 

Figure 4.19 BWR Combined RPV LOCA Frequencies at 25, 40, and 60 years .....................4-30 

Figure 5.1 Flowchart of the Analysis Framework ...................................................................5-3 

Figure 5.2 Flowchart of Required Calculations.......................................................................5-7 

Figure 7.1 Effect of Industry Safety Culture on Category 1 LOCAs.......................................7-3 

Figure 7.2 Effect of Regulatory Safety Culture on Category 1 LOCAs ..................................7-3 

Figure 7.3 Effect of Industry Safety Culture on Category 4 LOCAs.......................................7-4 

Figure 7.4 Total BWR and PWR LOCA Frequencies ............................................................7-6 

 (Means and 95th Percentile Values) as a Function of LOCA 

Category at 25 Years of Plant Operations 

Figure 7.5 Total BWR and PWR LOCA Frequencies.............................................................7-7 

 (Means and 95th Percentile Values) as a Function of the Threshold 

Break Diameter at 25 Years of Plant Operations 

Figure 7.6 BWR Piping and Non-Piping LOCA Frequencies (Current-day Estimates)..........7-8 

Figure 7.7 PWR Piping and Non-Piping LOCA Frequencies (Current-day Estimates)..........7-9 

Figure 7.8 BWR and PWR Mean Total LOCA Frequencies at 25 and 60 Years..................7-12 

Figure 7.9 Individual Estimates of BWR...............................................................................7-13 

 Category 1 LOCA Frequency Percentiles 

Figure 7.10 Individual Estimates of BWR Category 3 ............................................................7-14 

 LOCA Frequency Percentiles 

Figure 7.11 Individual Estimates of PWR Category 1.............................................................7-14 

 LOCA Frequency Percentiles 

Figure 7.12 Individual Estimates of PWR Category 6.............................................................7-15 

 LOCA Frequency Percentiles 

Figure 7.13 95% Confidence Intervals for Total BWR LOCA Frequencies ..........................7-17 

 at 25 Years of Plant Operations 
xi

Figure 7.14 95% Confidence Intervals for Total PWR LOCA Frequencies............................7-18 

at 25 Years of Plant Operations 

Figure 7.15 Mean Values for Selected Distributions by UEF < 100 .......................................7-21 

 (Median = 1 and LEF = 1000) 

 (Baseline Distribution is Truncated Split at 99.9%) 

Figure 7.16 Mean Values for Selected Distributions by UEF < 1200 .....................................7-22 

 (Median = 1 and LEF = 1000) 

 (Baseline Distribution is Truncated Split at 99.9%) 

Figure 7.17 Blanket 1 and Targeted 1 Coverage Interval Adjustments...................................7-24 

 For the Mean and 95th Percentile 

Figure 7.18 Targeted 3 Coverage Interval Adjustments..........................................................7-25 

 For the Mean, Median, and 95th Percentile 

Figure 7.19 BWR LOCA Frequencies with T3 Adjustment....................................................7-26 

Figure 7.20 PWR LOCA Frequencies with T3 Adjustment ....................................................7-27 

Figure 7.21 BWR LOCA Frequencies with Error-Factor Adjustment ....................................7-30 

Figure 7.22 PWR LOCA Frequencies with Error-Factor Adjustment.....................................7-31

Figure 7.23 Comparison of Baseline Results and Error-Factor Adjustment ...........................7-31 

Figure 7.24 Ratio of Monte Carlo to Baseline Results Assuming...........................................7-34 

 a Perfect Rank Correlation Structure 

Figure 7.25 Ratio of PRC to IC Structure Results...................................................................7-35 

Figure 7.26 Group Estimates of Mean PWR LOCA Frequencies at 25 Years........................7-37 

Figure 7.27 Group Estimates of Mean BWR LOCA Frequencies at 25 Years........................7-37 

Figure 7.28 Arithmetic-Mean and Geometric-Mean Group Estimates of ...............................7-38 

 Mean BWR LOCA Frequencies at 25 Years 

Figure 7.29 Arithmetic-Mean and Geometric-Mean Group Estimates of ...............................7-39 

 Mean PWR LOCA Frequencies at 25 Years 

Figure 7.30 Effect of Aggregation Point on PWR LOCA Frequency Estimates.....................7-43 

 at 25 Years of Operation 

Figure 7.31 Comparisons of Baseline and MEF Aggregation Parameters for.........................7-44 

 PWR LOCA Frequencies at 25 Years

Figure 7.32 BWR LOCA Frequencies Determined by Mixture-Distribution......................... 7-46 

 Aggregation: Current-day Estimates 

Figure 7.33 PWR LOCA Frequencies Determined by Mixture-Distribution......................... 7-46 

 Aggregation: Current-day Estimates 

Figure 7.34 Measures of Panel Diversity for BWR LOCA Frequencies at 25 Years .........7-49 

Figure 7.35 Measures of Panel Diversity for PWR LOCA Frequencies at 25 Years .........7-49 
xii

Figure 7.36 PWR LOCA Frequencies with Error-Factor Adjustment (reproduced from...... 7-51 

 Figure 7.21) 

Figure 7.37 PWR LOCA Frequencies with Error-Factor Adjustment.....................................7-52

 (reproduced from Figure 7.22) 

Figure 7.38 Comparison of Elicitation Estimates for PWR.....................................................7-57 

 Medium Break LOCAs at 25 Years with Estimates from Prior Studies 

Figure 7.39 Comparison of Elicitation Estimates for BWR ....................................................7-58 

 Large Break LOCAs at 25 Years with Estimates from Prior Studies 

LIST OF TABLES 

Table 1 Total BWR and PWR LOCA Frequencies.............................................................xxi 

 (After Overconfidence Adjustment using Error Factor Scheme) 

Table 1.1 Comparison of Mean Results from Previous Studies ............................................1-6 

 (per Calendar Years) 

Table 1.2 DEGB Frequencies for PWRs................................................................................1-7 

Table 3.1 LOCA Frequency Expert Panel..............................................................................3-5 

Table 3.2 LOCA Category Definitions...................................................................................3-9 

Table 3.3 Material Aging Degradation Mechanisms............................................................3-13 

Table 3.4 BWR LOCA-Sensitive Piping Systems ...............................................................3-14 

Table 3.5 PWR LOCA-Sensitive Piping Systems................................................................3-15 

Table 3.6 Reactor Pressure Vessel (RPV) Failure Modes....................................................3-16 

Table 3.7 Piping Base Case Definitions ...............................................................................3-18 

Table 3.8 Break Size to Leak Rate Correlation....................................................................3-27 

Table 4.1 Piping Base Case Frequency Results by Participant ..............................................4-6 

Table 4.2 Tabular Results from Vic Chapman’s ISI Sensitivity Analysis ...........................4-17 

Table 4.3 Average and Average Plus and Minus Two .........................................................4-27 

 Standard Deviations of the Adjusted Through-Wall Crack (TWC) 

 Frequency from the PTS Re-Evaluation Project 

Table 4.4 BWR RPV LOCA Frequencies Broken Down by Major Subcomponent............4-29 

 (Beltline and Feedwater Nozzles) 

Table 4.5 PWR RPV (CRDM) LOCA Frequencies .............................................................4-29 

Table 5.1 Elicitation Response Adjustment Schemes ..........................................................5-21 

Table 5.2 Summary of Monte Carlo Trials...........................................................................5-23 

Table 6.1 Major Piping and Non-Piping Contributors ..........................................................6-6 

 to the Various Size LOCA Categories 

Table 7.1 Total BWR and PWR LOCA Frequencies.............................................................7-6 
xiii

Table 7.2 Ratio of BWR and PWR Non-Piping.....................................................................7-9 

 Contribution to Piping Contribution 

Table 7.3 Sensitivity Analysis Summary Matrix..................................................................7-19 

Table 7.4 Overconfidence Adjustment Schemes..................................................................7-23 

Table 7.5 Targeted (T3) Overconfidence Adjustment..........................................................7-27 

 Compared to Unadjusted Results (Baseline) 

Table 7.6 Geometric Means of the Panelists’ Error Factors.................................................7-28 

 Current-day Estimates 

Table 7.7 Total BWR and PWR LOCA Frequencies...........................................................7-29 

(After Overconfidence Adjustment using Error-Factor Scheme) 

Table 7.8 Current-Day Total BWR and PWR LOCA Frequencies.....................................7-29 

 and Upper Confidence Bounds (After Overconfidence Adjustment 

 using Error-Factor Scheme) 

Table 7.9 Ratio of Adjusted Error Factor to Baseline Results .............................................7-30 

Table 7.10 Summary of Monte Carlo Trials...........................................................................7-33 

Table 7.11 Total BWR and PWR LOCA Frequencies ..........................................................7-40 

 (Using Arithmetic-Mean Aggregation 

Table 7.12 Ratio of AM to GM (Baseline) Group Estimates.................................................7-40 

Table 7.13 Total BWR and PWR LOCA Frequencies ..........................................................7-41 

 (Using Arithmetic-Mean Aggregation with Error-Factor 

 Overconfidence Adjustment) 

Table 7.14 Current-day Total BWR and PWR LOCA Frequencies.......................................7-45 

(Using Mixture Distribution for Aggregation) 

Table 7.15 Ratios of Mixture-Distribution to Geometric-Mean.............................................7-47 

Aggregated Results (not adjusted for overconfidence) 

Table 7.16 Current-day Total BWR and PWR LOCA Frequencies.......................................7-48 

(Using Mixture-Distribution Aggregation with Error-Factor 

Overconfidence Adjustment) 

Table 7.17 Total BWR and PWR LOCA Frequencies...........................................................7-51 

(After Overconfidence Adjustment using Error-Factor Scheme) 

(reproduced from Table 7.7 

Table 7.18 PWR Steam Generator Tube Rupture Frequencies (Geometric Mean Value ......7-53 

After Overconfidence Adjustment) 

Table 7.19 Total PWR LOCA Frequencies without SGTR Contributions.............................7-55 

(LOCA Categories 3 – 6 Reproduced from Table 7.17) 

Table 7.20 Comparison of Current Elicitation and Selected Prior Mean Estimates...............7-56 

xv

EXECUTIVE SUMMARY 

The emergency core cooling system (ECCS) requirements are contained in 10 CFR 50.46, Appendix K to 

Part 50, and GDC 35. Specifically, ECCS design, reliability, and operating requirements exist to ensure 

that the system can successfully mitigate postulated loss-of-coolant accidents (LOCAs). Consideration of 

an instantaneous break with a flow rate equivalent to a double-ended guillotine break (DEGB) of the 

largest primary piping system in the plant generally provides the limiting condition in the required 10 

CFR Part 50, Appendix K analysis. However, the DEGB is widely recognized as an extremely unlikely 

event. Therefore, the staff is establishing a risk-informed revision of the design-basis break size 

requirements for operating commercial nuclear power plants. 

A central consideration in selecting a risk-informed design basis break size is an understanding of the 

LOCA frequency as a function of break size. The most recent NRC-sponsored study of pipe break failure

frequencies is contained in NUREG/CR-5750. Unfortunately, these estimates are not sufficient for design 

basis break size selection because they do not address all current passive-system degradation concerns

and they do not discriminate among breaks having effective diameters greater than 6 inch. There have 

been two approaches traditionally used to assess LOCA frequencies and their relationship to pipe size: (i) 

estimates based on statistical analysis of operating experience and (ii) probabilistic fracture mechanics 

(PFM) analysis of specific postulated failure mechanisms. Neither approach is particularly suited to 

evaluate LOCA event frequencies due to the rareness of these events and the modeling complexity. In 

this study, LOCA frequency estimates have been calculated using an expert elicitation process to 

consolidate operating experience and insights from PFM studies with knowledge of plant design, 

operation, and material performance. This process is well-recognized for quantifying phenomenological

knowledge when data or modeling approaches are insufficient. 

The principal objective of this study was to develop separate boiling water reactor (BWR) and pressurized 

water reactor (PWR) piping and non-piping passive system LOCA frequency estimates as a function of 

effective break size at three distinct time periods: current-day (25 years fleet average), end-of-plantlicense (40 years fleet average), and end-of-plant-license-renewal (60 years fleet average). These 

estimates are based on the responses from an expert panel and one aim of this study was to obtain 

estimates that represent a type of group consensus. Additionally, another objective was to reflect both the 

uncertainty in each panelist’s estimates as well as the diversity among the individual estimates. 

The elicitation focused on developing generic, or average, estimates for the commercial fleet and the 

uncertainty bounds on these generic estimates rather than bounding values associated with one or two 

plants. This approach is consistent with prior studies that did not consider plant-specific differences in 

developing LOCA frequencies for use in probabilistic risk assessment (PRA) modeling. Consequently, 

the elicitation panelists were instructed to consider broad differences among plants related to important 

variables (i.e., plant system, material, geometry, degradation mechanism, loading, 

mitigation/maintenance) in determining both the generic LOCA frequencies and especially the estimated 

uncertainty bounds. It is the broad differences in these important variables that contribute most to passive 

system failure and there is generally sufficient commonality among plants to make such a generic 

assessment valuable. 

The elicitation was solely focused on determining LOCA frequencies that initiate by unisolable primary

system side failures that can be exacerbated by material degradation with age. Therefore, active system 
xvi

failures (e.g., stuck open valve, pump seals, interfacing system LOCAs) and consequential primary 

pressure boundary failures due to either secondary side failures or failures of other plant structures (e.g., 

crane drops) were not considered. Active system frequency contributions should be combined with the 

passive system LOCA frequencies to estimate the total system risk. The effects of safety culture on 

primary side failure (i.e., LOCA) frequencies were also considered. 

This study developed LOCA frequency estimates consistent with historical small break (SB), medium 

break (MB), and large break (LB) flow rate definitions. Additionally, three larger LOCA categories were 

defined within the classical LB LOCA regime to examine trends with increasing break size, up to and 

including, a DEGB of the largest piping system in the plant. Contrary to earlier studies, the six LOCA 

categories are defined in terms of cumulative thresholds rather than break intervals because this definition 

was more conducive to the elicitation structure. However, the differences between the interval and 

cumulative estimates are typically much smaller than the estimates’ uncertainties. It is therefore 

recommended that the cumulative threshold estimates from any set of tabular results be used (e.g., Table 

1) if interval-defined LOCA frequencies are desired. 

Because the LOCA frequency estimates were intended to be both generic and consistent with historical 

internal-event PRAs, the elicitation primarily considered normal plant operational cycles and loading 

histories. The loads include representative constant stresses (e.g. pressure, thermal, residual) and 

expected transient stresses (e.g. thermal striping, heat-up/cool-down, pressure transients) that occur over 

the license-renewal period. The elicitation implicitly considered all modes of operation based on the

loading or operational experience associated with each piping system or non-piping component. Rare 

event loading from seismic, severe water hammer, and other sources was also not considered in this 

generic evaluation because of their strong dependency on plant-specific factors. However, a separate 

research study was conducted to assess the potential impact of seismic loading on the break frequency 

versus break size relationship. As part of that study, both unflawed and flawed seismic piping 

contributions were considered. The results of that seismic LOCA analysis are summarized in Section 7.2. 

Several important assumptions were made to guide the elicitation process. One such assumption is that

plant construction and operation comply with all applicable codes and standards required by the 

regulations and the technical specifications. The specific impact on the LOCA frequencies of 

purposefully violating these requirements was not considered. However, it was assumed that regulatory

oversight policies and procedures will continue to be used to identify and mitigate risk associated with 

plants having deficient safety practices. While deviations from these requirements do represent some 

percentage of the events included in the passive-system failure data, extrapolation of this data implicitly 

assumes that similar future deviations will continue to occur with similar frequency. 

Another important assumption is that current regulatory oversight practices will continue to evaluate 

aging management and mitigation strategies in order to reasonably assure that future plant operation and 

maintenance has equivalent or decreased risk. A related assumption inherent in this elicitation is that all 

future plant operating characteristics will be essentially consistent with past operating practice. The 

effects of operating profile changes were not considered because of the large uncertainty surrounding 

possible operational changes and the potentially wide-ranging ramifications of significant changes on the 

underlying LOCA frequencies. 

The expert elicitation process employed in this study is an adaptation of the formal expert judgment 

processes used to evaluate reactor risk (NUREG-1150), to develop seismic hazard curves (NUREG/CR6372), and to assess the performance of radioactive waste repositories (NUREG/CR-5411). The process 

consisted of a number of steps. To begin, the project staff identified many of the issues to be evaluated 

through a pilot elicitation. The panel members for this pilot elicitation were all NRC staff. A group of 

twelve panel members was then selected for the formal elicitation. The staff gathered background 
xvii

material and prepared an initial formulation of the technical issues which was provided to the panel. At 

its initial meeting, the panel discussed the issues and, using the staff formulation as a starting point, 

developed a final formulation for the elicitation structure. This structure included the decomposition of 

the complex technical issues which impact LOCA frequencies into fundamental elements so that these 

important contributing factors could be more readily assessed. Piping and non-piping base cases were 

also defined for use in anchoring the quantitative elicitation responses. The base cases represent a set of 

well-defined conditions which could cause a LOCA. A subset of the panel was created to develop 

quantitative LOCA frequencies estimates associated with the base case conditions. At this initial meeting, 

the panel was also trained in subjective elicitation of numerical values through exercises and discussion of 

potential biases. 

After this initial meeting, the staff prepared a draft elicitation questionnaire and iterated with the panel to 

develop the final questionnaire. The panelists quantifying the base case conditions also developed their 

initial estimates. A second meeting was then held with the entire panel to review the base case results, 

review the elicitation questions, and finalize the formulation of remaining technical issues. At their home 

institutions, the individual panel members performed analyses and computations to develop answers to 

the elicitation questionnaire. 

The elicitation questionnaire required panelists to assess the following technical areas: the base case 

evaluation effort, utility and regulatory safety culture effects on LOCA frequencies, piping system LOCA 

frequencies, and non-piping system LOCA frequencies. The utility and safety culture questions required 

the panelists to compare future safety culture with the existing culture and predict the effect on LOCA 

frequencies. These effects were considered separately from passive system degradation because the 

panelist judged safety culture and degradation to be independent. The base case evaluation required the 

panelists to assess the accuracy and uncertainty in the base case analyses, and to also choose a particular 

base case approach for anchoring their elicitation responses. The piping and non-piping LOCA frequency 

questions required each panelist to first identify important LOCA contributing factors (i.e., piping 

systems, materials, degradation mechanisms, etc.) and select appropriate base case conditions for 

comparison. The panelists were then required to provide the relative ratios between their important 

contributing factors and the base case conditions based on their knowledge of passive system component

failure. Each relative comparison required mid value, upper bound, and lower bound values. The mid 

value is defined such that, in the panelist’s judgment, there is a 50% chance that the unknown true answer 

lies above the mid value. The upper and lower bounds are defined such that there is a 5% chance that the 

true answer lies above the upper bound or below the lower bound, respectively. Each panelist was also

required to provide their qualitative rationale supporting their quantitative values. 

A facilitation team consisting of individuals knowledgeable about the technical issues (substantive 

members), a staff member with extensive experience conducting expert elicitations (normative member) 

and two recorders met separately with each panel member in day-long individual elicitation sessions. At 

these sessions, each panel member provided answers to the elicitation questionnaire along with their 

supporting technical rationales. The panelists were asked to self-select, based on their expertise, the 

questions that they addressed. Consequently, several panelists only provided responses for either BWR

or PWR plants. After the elicitation sessions, the panel members returned to their home institutions 

where they refined their responses based on feedback obtain during their session. Upon receipt of the

updated responses, the project staff compiled the panel’s responses and developed preliminary estimates 

of the LOCA frequencies. Along with the rationales, these preliminary estimates were presented to the

panel at a wrap-up meeting. Panel members were invited to fill in gaps in their questionnaire responses 

and, if desired, to modify any of their responses based on group discussion of important technical issues 
xviii

considered during the individual elicitations1

. Final individual estimates of the LOCA frequencies were 

then calculated and provided to the panel members for final review and quality assurance. 

The qualitative insights provided by the panel are reasonably consistent. Panelists identified several 

advantages and disadvantages with determining the base case LOCA frequencies through operatingexperience assessment or PFM analyses. However, there was a general consensus that operating 

experience provides the best basis for evaluating current-day, small-break LOCA estimates. Hence, many 

panelists used the operating-experience base cases to anchor their elicitation responses. The panel 

members also generally expressed the opinions that the future safety culture will not differ dramatically 

from the current culture, and that the utility and regulatory safety cultures are highly correlated. Many 

panelists do believe that safety culture can significantly affect LOCA frequencies at specific plants, but 

there is also an expectation that regulatory actions using existing enforcement measures will diminish

both the possibility and impact of deficient safety culture at particular plants. However because it was 

thought that these plant-specific issues do not affect the generic averages, no specific adjustment to the 

LOCA frequency estimates was applied to explicitly account for safety culture effects. This decision was 

endorsed by the elicitation panelists. 

There were several technical insights that were consistently identified. Many participants believe that the 

number of precursor events (e.g., cracks and leaks) is generally a good barometer of the LOCA 

susceptibility for the associated degradation mechanism. Welds are almost universally recognized as 

likely failure locations because they can have relatively high residual stress, are preferentially-attacked by 

many degradation mechanisms, and are most likely to have preexisting fabrication defects. Most 

panelists also agreed that a complete break of a smaller pipe, or non-piping component, is generally more 

likely than an equivalent size opening in a larger pipe, or component, because of the increased severity of 

fabrication or service cracking. Therefore, the biggest frequency contributors for each LOCA size tend to 

be systems having the smallest pipes, or component, which can lead to that size LOCA. The exception to 

this general rule is the BWR recirculation system, which is important at all LOCA sizes due to IGSCC. 

Many panelists thought that aging may have the greatest effect on intermediate diameter (i.e., 6 to 14-inch 

nominal diameter) piping systems due to the large number of components within this size range and the 

fact that this piping generally receives less attention than the larger diameter piping and is harder to 

replace than the more degradation-prone smaller diameter piping. 

The participants generally identified thermal fatigue, stress corrosion cracking (SCC), flow accelerated 

corrosion (FAC), and mechanical fatigue as the degradation mechanisms that most significantly 

contribute to LOCA frequencies in BWR plants. Generally, the most important BWR degradation 

mechanism is intergranular SCC (IGSCC), although the panelist’s recognize that mitigation has greatly 

reduced the susceptibility of BWR plants to this mechanism over the past 20 years. With the exception of 

FAC, similar degradation mechanisms and concerns were also deemed to be important in PWR plants. 

Specifically, primary water SCC (PWSCC) is a principal concern. Many panelists believe that PWSCC 

will be mitigated in PWR plants over the next 15 years, but that effective mitigation has yet to be 

developed and implemented. 

The panelists generally agreed on the important technical issues and LOCA-contributing factors. 

However, the individual quantitative responses are much more uncertain and there are relatively large 

differences among the panelists’ responses. This is to be expected given the underlying scientific 

uncertainty. The analysis of these responses was structured to account for the uncertainty in the 

 

1

Each panelist’s quantitative elicitation responses can be found through the “Electronic Reading Room” link on the NRC’s 

public website (http://www.nrc.gov/) using the Agencywide Documents Access and Management System (ADAMS). The 

document is found in ADAMS using the following accession number: ML080560005. 
xix

individual estimates and the diversity among the panelists. The quantitative responses were analyzed 

separately for each panel member to develop individual BWR and PWR total LOCA frequency estimates. 

A unified analysis format was developed to ensure consistency in processing the panelists’ inputs. The 

panelists’ mid-value, upper bound and lower bound responses were assumed to represent the median, 95th, 

and 5th percentiles, respectively, of their subjective uncertainty distributions for each elicitation response. 

The analysis structure was based on the assumption that all the responses correspond to percentiles of

lognormal distributions. These distributions were then combined using a lognormal framework. The 

final outputs for each panelist are estimates for the means, medians and 5th and 95th percentiles of the total 

BWR and PWR LOCA frequencies. The panelists’ estimates were then aggregated to obtain group 

LOCA frequency estimates, along with measures of panel diversity. 

The individual and group estimates for the means, medians, 5th and 95th percentiles of the LOCA 

frequency distributions were calculated using the following principal assumptions and choices: 

(i) The mid-value, upper bound, and lower bound supplied by each panelist for each elicitation 

question were assumed to correspond to the median, 95th percentile and 5th percentile, 

respectively, of a split lognormal distribution, with the mean calculated assuming that the 

upper tail is truncated at the 99.9th percentile. 

(ii) Only those panelists whose uncertainty ranges were relatively small were adjusted using an 

error-factor adjustment scheme to account for possible overconfidence (Section 7.6.2.2). 

(iii) Split lognormal distributions were summed by assuming perfect rank correlation among the 

individual terms. 

(iv) The individual estimates are the total LOCA frequency parameters (i.e., mean, median, 5th

percentile, and 95th percentile) determined for each panelist. 

(v) The group estimates of the total LOCA frequency parameters were determined using the 

geometric means of the individual estimates. 

(vi) Panel diversity was characterized with two-sided 95% confidence intervals based on an 

assumed lognormal model for the individual estimates. 

In the report, these six assumptions and choices define what are termed the summary estimates. The 

report also calculates and discusses baseline estimates. These baseline estimates are calculated using all 

the above assumptions, except no overconfidence adjustment (as described in (ii) above) is applied. 

Because it is well-established that experts tend to be overconfident, the overconfidence adjustment is

deemed to result in improved LOCA frequency estimates 

The resultant individual and group summary estimates are consistent with the elicitation objectives and 

structure and are reasonably representative of the panelists’ quantitative judgments. In particular, they are 

not dominated by extreme results, either on the high or low end, and the geometric means of the 

individual estimates approximate the medians of these estimates. The median is often used to represent 

group opinion in elicitations, especially when the individual estimates differ by several orders of 

magnitude, as they do in this study. 

The LOCA frequency summary estimates for the current-day (25 years) and end-of-plant-license (40 

years) periods are provided in Table 1 for both BWR and PWR plant types. The aggregated group 

estimates for the median, mean, 5th and 95th percentiles are presented. Frequency estimates are not 

expected to change dramatically over the next 15 years for any size LOCA, or even over the next 35 years 

for BWR LOCA Categories 1- 5 and PWR LOCA Categories 1 - 2 (see results in Appendix L). While 

order of magnitude increases in BWR LOCA Category 6 and PWR LOCA Categories 3 - 6 over the next 

35 years are expected, these increases are largely due to uncertainty about the future and the concern that 

new degradation mechanisms could arise in the operating fleet. However, while aging will continue, the 

panelists’ consensus is that mitigation procedures are in place, or will be implemented in a timely manner, 
xx

to alleviate significant increases in future LOCA frequencies for existing degradation mechanisms. 

Because of the predicted stability in these estimates over the near-term, the current-day (25 year) results 

can be used to represent the LOCA frequencies over the next 15 years of fleet operation.

The current-day median, mean, and 95th percentile estimates are graphically presented in Figure 1. The 

95% confidence intervals calculated for these parameters are also illustrated in this figure. The LOCA 

frequencies as a function of threshold break diameter were estimated only for the six specified LOCA 

categories in the elicitation. The plotted points are connected with straight lines in the figure for visual 

clarity, but this should not be construed as a recommended interpolation scheme. Interpolation of 

frequencies between category sizes can be done at the user’s discretion depending on the conservatism 

required by the application. Some common interpolation schemes are linear, multi-point nonlinear and 

cubic spline. A step-wise or stair-step interpolation between two categories where the frequency for the 

lower category size is used for all flow rates or corresponding break sizes between the two categories

provides the most conservative interpolation scheme. Note that any interpolation scheme does not reflect 

the uncertainty in the interpolated frequencies. 

A measure of the individual uncertainties in Table 1 and Figure 1 is given by the differences between the 

medians and the corresponding 5th or 95th percentile estimates. Panel diversity is reflected in the 

confidence bounds in Figure 1. The large widths of these confidence bounds (as much as 3 orders of 

magnitude) reflects the significant diversity of the individual estimates in this study As the LOCA size 

increased, the panel members generally expressed greater uncertainty in their predictions, and the 

variability among individual panelists’ estimates increased. This is to be expected because of the 

increased extrapolation required from available passive-system failure data for larger LOCA sizes. 

While it is acknowledged that operating experience-based estimates do not necessarily reflect the current 

state, it is informative to compare such estimates with the elicitation frequencies for the smallest LOCAs 

(Category 1). This comparison requires the least extrapolation of passive-system failure data since no 

larger LOCAs have occurred. The BWR and PWR Category 1 LOCA frequencies (including the steam 

generator tube rupture (SGTR) frequency for PWRs) were estimated up through December 2006. For 

BWR plants, the average SB LOCA frequency based solely on the number of reported events was 

estimated to be 5.5E-04 per calendar year. The mean elicitation BWR SB LOCA estimate is 6.5E-04 per 

calendar year. The BWR elicitation estimate is less than 20 percent higher than the operating-experience 

estimate which, given the uncertainty of these estimates, is not statistically significant. 

For PWR plants, the average SB LOCA frequency estimated analogously is 3.6E-03 per calendar year. 

The mean elicitation PWR SB LOCA estimate is 7.3E-03 per calendar year. The PWR elicitation 

estimate is about 100 percent (or a factor of two) higher than the operating-experience-based estimate. 

Additional insight into this difference can be gained by partitioning the PWR passive-system failure data 

into frequencies for SGTRs and for all other passive-system SB LOCAs. 

Based on reported failures, the mean SGTR LOCA frequency is 3.2E-03 per calendar year. This result is

almost identical to the current-day elicitation estimate of 3.7E-03 per calendar year. The frequency of all 

other Category 1 PWR passive-system failures is 4.0E-04 per calendar year based on operating 

experience. The corresponding elicitation estimate is 1.9E-03 per calendar year. While this value is 5 

times greater than the operating-experience-based estimate, this difference is explained by the elicitation 

panelists’ estimation of the effect of PWSCC on small diameter component failures. 

There are several prior studies that also estimated LOCA frequencies. Some care is needed when 

comparing the elicitation LOCA frequency estimates with these earlier studies because the LOCA 

categories are defined differently. Specifically, the current-day LOCA Category 1 and 2 estimates in 

Table 1 are comparable to total system SB, MB, and LB LOCA frequencies, respectively, reported in 
xxi

NUREG/CR-5750. Additionally, current-day SGTR frequencies (Table 7.18) and PWR LOCA 

frequencies for all other passive system failures (i.e., frequencies for breaks greater than 100 gpm (380 

lpm) in Table 7.19) are comparable to NUREG/CR-5750 SGTR and PWR SB LOCA frequencies when 

these failure modes are analyzed separately. The NUREG/CR-5750 LB LOCA frequency estimates are 

best compared to the elicitation LOCA Category 4 frequency estimates because the pipe break sizes are 

most similar. 

After accounting for these differences, the elicitation LOCA frequency estimates are generally much 

lower than the WASH-1400 estimates and more consistent with the NUREG/CR-5750 estimates. The SB 

LOCA PWR elicitation estimates after subtracting the SGTR frequencies are approximately 3 times 

greater than the NUREG/CR-5750 estimates, due to the aforementioned PWSCC concerns. However, the 

total BWR and PWR SB LOCA frequency estimates are similar once the SGTR frequencies are added to 

the NUREG/CR-5750 PWR results. The elicitation MB LOCA estimates are higher than the 

NUREG/CR-5750 estimates by factors of approximately 4 and 20 for BWR and PWR plant types, 

respectively. These increases are partly due to concerns about PWSCC of piping and non-piping (e.g., 

CRDM) components as well as general aging concerns with piping in this size range. The NUREG/CR5750 LB LOCA frequency estimates are slightly higher (less than a factor of 3) than the current elicitation 

results for both PWR and BWR plants. The generally good agreement between the NUREG/CR-5750 

and current elicitation estimates is somewhat surprising given the markedly different methodologies used 

to arrive at these results. 

 Table 1 Total BWR and PWR LOCA Frequencies 

(After Overconfidence Adjustment using Error-Factor Scheme) 

Current-day Estimate (per cal. yr) End-of-Plant-License Estimate (per cal. yr) 

(25 yr fleet average operation) (40 yr fleet average operation) 

Plant 

Type 

LOCA 

Size 

(gpm) 

Eff. 

Break 

Size 

(inch) 5

th Per. Median Mean 95th Per. 5th Per. Median Mean 95th Per. 

>100 1⁄2 3.3E-05 3.0E-04 6.5E-04 2.3E-03 2.8E-05 2.6E-04 6.2E-04 2.2E-03 

>1,500 1 7/8 3.0E-06 5.0E-05 1.3E-04 4.8E-04 2.5E-06 4.5E-05 1.2E-04 4.8E-04 

>5,000 3 1⁄4 6.0E-07 9.7E-06 2.9E-05 1.1E-04 5.4E-07 9.8E-06 3.2E-05 1.3E-04 

>25K 7 8.6E-08 2.2E-06 7.3E-06 2.9E-05 7.8E-08 2.3E-06 9.4E-06 3.7E-05 

>100K 18 7.7E-09 2.9E-07 1.5E-06 5.9E-06 6.8E-09 3.1E-07 2.1E-06 7.9E-06 

BWR 

>500K 41 6.3E-12 2.9E-10 6.3E-09 1.8E-08 7.5E-12 4.0E-10 1.0E-08 2.8E-08 

>100 1⁄2 6.9E-04 3.9E-03 7.3E-03 2.3E-02 4.0E-04 2.6E-03 5.2E-03 1.8E-02 

>1,500 1 5/8 7.6E-06 1.4E-04 6.4E-04 2.4E-03 8.3E-06 1.6E-04 7.8E-04 2.9E-03 

>5,000 3 2.1E-07 3.4E-06 1.6E-05 6.1E-05 4.8E-07 7.6E-06 3.6E-05 1.4E-04 

>25K 7 1.4E-08 3.1E-07 1.6E-06 6.1E-06 2.8E-08 6.6E-07 3.6E-06 1.4E-05 

>100K 14 4.1E-10 1.2E-08 2.0E-07 5.8E-07 1.0E-09 2.8E-08 4.8E-07 1.4E-06 

 

PWR 

>500K 31 3.5E-11 1.2E-09 2.9E-08 8.1E-08 8.7E-11 2.9E-09 7.5E-08 2.1E-07 

Sensitivity analyses were also conducted to examine the robustness of the quantitative results to the 

analysis procedure used to develop the summary estimates. These sensitivity analyses investigated the

effect of distribution shape on the means as well as the effects of correlation structure, panelist 

overconfidence, panel diversity measure, and aggregation method on the estimated parameters. The mean

calculation in the analysis procedure used a split lognormal distribution truncated at the 99.9th percentile 

to obtain reasonably conservative values compared with other possible choices. The correlation structure 

in the analysis procedure assumed maximal correlation, which is reasonably representative of the 

elicitation structure. The structure provides conservative 95th percentile estimates. However, based on 

selected Monte Carlo simulations, an independent correlation structure leads to larger median and 5th

percentile estimates. The means are unaffected by the choice of the correlation structure. The analysis 

procedure also used confidence intervals for the aggregated estimates to measure panel diversity. An 

alternative approach used quartiles of the individual estimates, leading to comparable, but narrower 

intervals. 
xxii

The analysis procedure also adjusted those panelists’ responses that had relatively narrow uncertainty

bands using an error-factor scheme to account for a known tendency for people, including experts, to be 

overconfident when making subjective judgments. Sensitivity analyses examined the effects of other 

overconfidence adjustments of the nominal subjective confidence levels supplied by the panelists. No 

overconfidence adjustment was also investigated. While blanket overconfidence adjustments can result in 

large, unsupportable increases in the mean and 95th percentile frequency estimates, no adjustment results 

in modest decreases in these estimates. Therefore, the error-factor scheme, which adjusts only those 

panelists who are most overconfident, is deemed to the most appropriate. 

Finally, the largest sensitivity is associated with the method used to aggregate the individual panelist 

estimates to obtain group estimates. The baseline and summary estimates were developed using 

geometric-mean aggregation. In this study, the geometric-mean aggregation produces frequency 

estimates that approximate the median of the panelists’ estimates and therefore effectively leads to 

consensus-type results. Therefore, the summary estimates in Table 1 are believed to be a reasonable 

representation of the expert panel’s current state of knowledge regarding LOCA frequencies. 

The sensitivity analyses evaluated the effect of using alternative aggregation methods to calculate the 

group estimates. Specifically, mixture-distribution and arithmetic-mean aggregation were evaluated. For 

the panelists’ responses in this study, these alternative aggregation methods can lead to significantly 

higher mean and 95th percentile estimates than those obtained using geometric-mean aggregation. 

Alternative LOCA frequency estimates that are higher than the summary estimates (Table 1) can be 

derived by using either the summary estimates with 95% confidence bounds (Tables 7.8), the arithmeticmean aggregated results (Table 7.13), or the mixture-distribution results (Table 7.16). These estimates 

also incorporate the same overconfidence adjustment as the summary estimates. 

Because alternative aggregation methods can lead to significantly different results, a particular set of 

LOCA frequency estimates is not generically recommended for all risk-informed applications. The 

purposes and context of the application must be considered when determining the appropriateness of any

set of elicitation results. In particular, during the selection of the BWR and PWR transition break sizes 

for the proposed 10CFR50.46a rule making, the NRC staff considered the totality of the results from the 

sensitivity studies rather than only the summary estimates from this study The NRC anticipates that a

similar approach will be used in selecting appropriate replacement frequencies for NUREG/CR-5750 

estimates and for other applications where frequencies for break sizes other than those in NUREG/CR5750 are required. While the lack of clear application guidance places an additional burden on the users 

of the study results, those users are in the best position to judge which study results are most appropriate 

to consider for their particular applications. 
xxiii

Threshold Break Diameter (in)

0.1 1 10 100

LOCA Frequency (CY-1

)

10-10

10-9

10-8

10-7

10-6

10-5

10-4

10-3

10-2

10-1

Median

Mean

95th Percentile

BWR: Error Factor Adjustment

median and 95th percentile results offset slightly for clarity

Threshold Break Diameter (in)

0.1 1 10 100

Frequency (CY-1

)

10-10

10-9

10-8

10-7

10-6

10-5

10-4

10-3

10-2

10-1

Median

Mean

95th Percentile

PWR: Error Factor Adjustment

median and 95th percentile results offset for clarity

Figure 1 BWR and PWR Error-Factor Adjusted LOCA Frequency Estimates 

xxv

ACKNOWLEDGMENTS

This work was supported by the United States Nuclear Regulatory Commission (NRC) through the 

Component Integrity Branch of the Division of Engineering of the Office of Nuclear Regulatory Research

under several different contracts. Battelle Memorial Institute support was provided under NRC Contract 

NRC-04-02-074 (NRC Job Code Y6538). Mr. William Galyean of Idaho National Engineering 

Environmental Laboratory was supported under NRC Job Code Y6332. Dr. Cory Atwood was supported 

under NRC Job Code Y6492. Dr. Alan Brothers was supported under NRC Job Code Y6604. 

The authors would first like to thank the members of the expert elicitation panel whose insights formed 

the basis for the results and conclusions reached in this report. The elicitation panel members also 

provided valuable editorial and technical comments that have been incorporated into this summary 

document. The panel members included: 

• Mr. Bruce Bishop Westinghouse Electric Co. LLC 

• Dr. Vic Chapman OJV Consultancy Limited 

• Mr. Guy DeBoo Exelon Nuclear 

• Mr. William Galyean Idaho National Engineering Environmental Laboratory 

• Dr. Karen Gott Swedish Nuclear Power Inspectorate

• Dr. David Harris Engineering Mechanics Technology, Inc. 

• Mr. Bengt Lydell ERIN® Engineering and Research, Inc. 

• Dr. Peter Riccardella Structural Integrity Associates, Inc. 

• Mr. Helmut Schulz Gesellschaft für Reaktorsicherheit (GRS) mbh 

• Dr. Sampath Ranganath Formerly GE Nuclear Energy/Now XGEN Engineering 

• Dr. Fredric Simonen Pacific Northwest National Laboratory 

• Dr. Gery Wilkowski Engineering Mechanics Corporation of Columbus 

A special debt of gratitude is expressed to the base case team members (Dr. David Harris, Mr. Bengt 

Lydell, Mr. William Galyean, and Dr. Peter Riccardella) for the extra effort they provided to conduct the 

base case and sensitivity analyses discussed in Sections 3 and 4 of this report. In addition, Dr. Riccardella 

provided base case frequencies for a series of non-piping components and Mssrs. Galyean and Lydell 

developed the non-piping precursor database. Summaries of the base case team members’ analyses are 

contained in Appendices D – I. Each appendix has been written by the responsible base case team 

member and these contributions are much appreciated and are vital to the technical basis of this NUREG. 

Mr. Lydell and Dr. Harris also tirelessly discussed and corrected the authors’ characterization of their 

efforts for accuracy and clarity. 

The authors would also like to recognize the facilitation team members who assisted with the 

development of the technical issues and associated elicitation questions and who also participated in the 

individual elicitations sessions. Along with the authors, Dr. Ken Jacquay of Casco Services, Inc. was a 

principal member of the facilitation team 

Other members included: 

• Ms. Bennett Brady NRC 

• Mr. Frank Cherny NRC

• Mr. Alan Kuritzky NRC 

• Mr. Arthur Salomon NRC 

The authors are also appreciative of Dr. Corwin Atwood of Statwood Consulting and Dr. Alan Brothers 

of Pacific Northwest National Laboratory, who provided an external peer review of the elicitation 
xxvi

approach and analysis of results. They suggested a number of sensitivity studies to further validate and 

clarify the conclusions reached in this report. They also provided many helpful comments that have been 

used to clarify this report. Dr. Atwood, in particular, provided many valuable insights which were used to 

develop and refine the techniques used to analyze the elicitation results. 

Many NRC staff also provided valuable comments and contributions that were used to revise the draft 

report, including Dr. Arthur Buslik, Mr. Stephen Dinsmore, Mr. Gary Hammer, Mr. Glenn Kelly, and Mr. 

Arthur Salomon,. In particular, the authors would like to thank Dr. Arthur Buslik for his contribution to 

Section 5.6.4.4 and Mr. Arthur Salomon for his reviewing and editing assistance. The authors are also

grateful to Dr. David Rudland of Engineering Mechanics Corporation of Columbus for assisting with 

selected Monte Carlo calculations to evaluate the analysis of individual panelist responses. 

Finally, the authors would like to thank Dr. Al Csontos of the NRC, who served as the program manager 

over the final months of this program, and to Ms. Charlotte Matthews and Ms. Patricia Zaluski, both of

Battelle-Columbus, for their invaluable assistance in the preparation of this report. 
xxvii

 NOMENCLATURE 

1. Symbols

a Flaw depth 

a

*

 Crack depth having 50% chance of being detected 

b Value of the upper bound supplied by a panelist

b’ Value of the lower bound supplied by a panelist 

bp(Y) pth percentile of distribution Y 

C Parameter in fatigue crack growth relationship 

C1, C2 Coefficients used in probability of detection curve definition 

D Diameter 

DN Nominal pipe diameter 

E Young’s modulus 

E(Y) Expected value of distribution Y 

EF(Y) Error factor of distribution Y 

f Fraction of welds inspected for cracks or flaws

f Inspection coverage/scope 

g Median of the lognormal distribution U 

g

*

 Group estimate of g 

J J-integral fracture parameter 

JD Deformation J 

JIc Plane strain J at crack initiation by ASTM813 

JM Modified J 

J-R J-resistance 

ka path percentile of the standard lognormal distribution 

kp pth percentile of the standard lognormal distribution 

K Stress intensity factor 

KI Mode I stress intensity factor 

KIc Plane strain stress intensity factor at crack initiation 

fPOD Probability of detection function 

I ISI effectiveness factor 

L Length 

LBC Likelihood of a leak due to any degradation mechanism (base case) 

LPL Likelihood of a perceptible leak 

LTSL Likelihood of a technical specification leak 

L50 Likelihood of a crack 50% through wall deep 

m Median value 

m Parameter in fatigue crack growth relationship 

m(Y) Mean of distribution Y 

n Number of panelists 

nc

 Number of cracks or flaws 

nf Number of failure events 

N Normal operating stress 

N Number of components 

N Number of stress cycles 

p Percentile 

P Pressure 

pa

 Value of the percentile in the adjusted distribution corresponding to the original 

error factor 

PBC Conditional failure probability for the chosen seismic piping base case 
xxviii

PFD Probability inspected welds will find existing flaw

PND Probability of not detecting a crack 

PPL Conditional failure probability of a crack that has just formed a perceptible leak 

PTSL Conditional failure probability of a crack leaking at the technical specification 

limit 

PTSL@SLB Conditional failure probability of a crack leaking at the technical specification 

limit assuming a Service Level B load 

PTSL@SLD Conditional failure probability of a crack leaking at the technical specification 

limit assuming a Service Level D load 

P50 Conditional failure probability of a crack with a maximum depth of 50% of the 

wall thickness 

pL/F Conditional failure probability 

p’ Value of the percentile of the panelist’s assumed lognormal distribution 

corresponding to b’ 

r Error factor 

rp Error factor of the panelist’s adjusted lognormal distribution 

R Stress ratio 

RC/F Number of non-through-wall cracks per leak event

S Sum of cyclic stress 

S Seismic 

S

2

 Sample variance 

SD(Y) Standard deviation of distribution Y 

t Wall thickness 

T Thermal 

T Tearing modulus 

T Total time 

V(Y) Variance of distribution Y 

Z1 Anchoring factor 

Z2 Adjustment ratios 

ε Strain 

ε Probability of not detecting a crack regardless of depth 

λ Pipe failure frequency (through-wall crack) 

μ Mean 

ν Parameter controlling slope of PND curve 

φ Total frequency (cracks and leaks) 

σ Stress 

σ Standard deviation 

σ

2

 Variance 

σDW Dead weight stress 

σNO Normal operating stress 

σte Thermal expansion stress

2. Acronyms and Initialisms

ACRS Advisory Committee on Reactor Safeguards 

ADAMS Agencywide Document Access and Management System 

AM Arithmetic mean 

ANL Argonne National Laboratories 

ASME American Society of Mechanical Engineers 

ASTM American Society for Testing and Materials 

BINP Battelle Integrity of Nuclear Piping 
xxix

BWR Boiling water reactor 

B&W Babcock and Wilcox 

CBP Conditional break probability 

CD Compact disk 

CDF Cumulative distribution function 

CE Combustion Engineering 

CFR Code of Federal Regulations 

CFP Conditional failure probability 

COD Crack opening displacement 

CRD Control Rod Drive 

CRDM Control Rod Drive Mechanism 

CS Carbon steel 

CV Correct value 

CVCS Chemical Volume and Control System 

CY Calendar year 

DEGB Double ended guillotine break 

DVI Direct Volume Injection 

DW Dead weight 

ECCS Emergency Core Cooling System 

ECSCC External chloride stress corrosion cracking

EDY Effective degradation years 

EFa

 Error factor after overconfidence adjustment 

EFi

 Error factor of distribution Zi

EF0 Error factor before overconfidence adjustment 

EF(Y) Error factor of Y

EFPY Effective full power years 

Emc2 Engineering Mechanics Corporation of Columbus 

EMT Engineering Mechanics Technology 

EPRI Electric Power Research Institute 

EQ Elicitation question 

FAC Flow accelerated corrosion 

FAD Failure assessment diagram 

FDR Fabrication defect and repair 

FS Flow sensitive 

FW Feed water 

GALL Generic aging lessons learned 

GC General corrosion 

GDC General Design Criterion 

GE General Electric 

GL Generic letter 

GM Geometric mean 

GRS Gesellschaft für Reactorsicherheit 

HAZ Heat affected zone 

HPCS High Pressure Core Spray 

HPI/MU High Pressure Injection/Make-up 

HWC Hydrogen water chemistry 

IAEA International Atomic Energy Agency 

IC Independent correlation 

ICI In-core Instrumentation 

ID Inside diameter 

IGSCC Intergranular stress corrosion cracking 
xxx

IHSI Induction heat stress improvements 

INEEL Idaho National Engineering and Environmental Laboratory 

IPE Individual plant evaluation 

IPIRG International Piping Integrity Research Group 

IQR Interquartile range 

IRS Incident Reporting System 

ISI In-service inspection 

IS LOCA Interfacing system loss of coolant accident 

LAS Low alloy steel 

LB Large break 

LB Lower bound 

LBB Leak before break 

LC Localized corrosion 

LEF Lower error factor 

LER Licensee Event Report 

LERF Large early release frequency 

LIV Loop Isolation Valve 

LLNL Lawrence Livermore National Laboratory 

LOCA Loss of coolant accident 

LOOP Loss of offsite power 

LPCS Low Pressure Core Spray 

LPHSW Last pass heat sink welding 

LQ Lower quartile 

LTOP Low temperature over pressurization 

MA Material aging 

MB Medium break 

MERIT Maximizing Enhancements in Risk Informed Technology program 

MF Mechanical fatigue 

MITI Ministry of International Trade and Industry (Japan) 

MRP Materials Reliability Program 

MSIP Mechanical Stress Improvement Process 

MSIV Main Steam Isolation Valve 

MV Mid value 

NB Nickel-based weld 

NDE Non-destructive examination 

NG Nuclear grade 

NPP Nuclear power plant 

NPS Nominal pipe size 

NRC Nuclear Regulatory Commission 

NSSS Nuclear steam supply system 

NUPEC Nuclear Power Engineering Test Center (Japan) 

NWC Normal water chemistry 

OBE Operational basis earthquake 

OD Outside diameter 

OECD Organization for Economic Co-operation and Development 

OPDE OECD Piping Data Exchange 

ORNL Oak Ridge National Laboratories 

PFM Probabilistic fracture mechanics 

PIFRAC Pipe fracture mechanics material property database 

PIV Pressurizer isolation valve 

PNNL Pacific Northwest National Laboratories 
xxxi

POD Probability of detection 

PORV Power operated relief valve 

PRA Probabilistic risk assessment 

PRC Perfect rank correlation 

PSA Probabilistic safety assessment 

PSI Pre-service inspection 

PSL Pressurizer Spray Line 

PTS Pressurized thermal shock 

PVP Pressure Vessel and Piping 

PWHT Post weld heat treatment 

PWR Pressurized water reactor 

PWSCC Primary water stress corrosion cracking 

QA Quality assurance 

QC Quality control 

RCIC Reactor Core Isolation Cooling 

RCP Reactor Cooling Piping 

RCPB Reactor Coolant Primary Boundary 

RCS Reactor cooling system 

RES Office of Nuclear Regulatory Research 

RH Reactor head 

RHR Residual Heat Removal 

RI-ISI Risk informed in-service inspection 

RPV Reactor Pressure Vessel 

RR Rolls Royce 

RS Residual stress 

RV Random variable 

RWCU Reactor Water Cleanup 

SAM Seismic anchor motion 

SB Small break 

SCC Stress corrosion cracking 

SCSS Sequence Coding and Search System 

SG Steam generator 

SGTR Steam generator tube rupture 

SI Stress improvement 

SIS Safety Injection System 

SKI Swedish Nuclear Inspectorate 

SLB Service Level B 

SLC Standby Liquid Control 

SLD Service Level D 

SRM Staff requirements memorandum 

SRV Safety relief valve 

SQUIRT Seepage Quantification of Upsets in Reactor Tubes 

SS Stainless steel 

SSE Safe shutdown earthquake 

TBS Transition break size 

TF Thermal fatigue 

TGM Trimmed geometric mean 

TGSCC Transgranular stress corrosion cracking 

TMI Three Mile Island 

TS Thermal stratification 

TSL Technical specification leakage 
xxxii

TWC Through-wall crack 

UA Unanticipated mechanism 

UB Upper bound 

UEF Upper error factor 

UQ Upper quartile 

US United States 

USNRC United States Nuclear Regulatory Commission 

VTC Video Teleconference 

WH Water hammer 

WO Weld overlay 

WOR Weld overlay repair 

WOG Westinghouse Owners Group 
APPENDIX A 

PANEL MEMBER QUALIFICATIONS 

A-1

APPENDIX A 

PANEL MEMBER QUALIFICATIONS

This appendix contains short resumes/biographical sketches for each of the panel members demonstrating

their qualifications and credentials for this expert elicitation. 
A-2

BRUCE BISHOP 

PRINCIPAL ENGINEER 

RELIABILITY AND RISK ASSESSMENT GROUP 

WESTINGHOUSE ELECTRIC COMPANY’S NUCLEAR SERVICE DIVISION 

PITTSBURGH, PENNSYLVANIA

Mr. Bishop has been at Westinghouse for almost 35 years, working primarily in the area of structural 

reliability analysis, initially on breeder reactor core components and then on light water RPVs, piping and 

other primary loop components. During this time, Mr. Bishop developed and applied the structural 

reliability and risk assessment (SRRA) models, methods and software PFM analyses supporting a number 

of risk informed inspection initiatives. Included are the SRRA applications for most plant piping, reactor 

internals and reactor coolant pump components, and the irradiated belt line region, head penetration 

nozzles and nozzle inner radius region of the RPV. Mr. Bishop has been the recipient of six George 

Westinghouse Signature Awards for Engineering Excellence and two Special Recognition Awards from 

the ASME Pressure Vessel and Piping Division for conference tutorials on Application of Probabilistic 

Structural Mechanics. He has participated in the development and presentation of numerous 

Westinghouse technical reports and more than 35 publications on application of structural reliability to 

risk informed decisions, including a chapter on pressure vessel and piping applications in the 

Probabilistic Structural Mechanics Handbook. Mr. Bishop, a member of the editorial boards for 

Reliability Engineering and System Safety and Nuclear Engineering and Design, is currently a member of 

the ASME Safety Engineering and Risk Analysis Division and the ASME Codes and Standards Working 

Groups on Implementation of Risk-Based Inspection and Operating Plant Criteria (for RPV integrity 

issues). He also actively participates in the PWR Materials Reliability Program (MRP) Issue Task 

Groups on Reactor Vessel Integrity and Alloy-600 Issues, including the PFM analyses of the RPV during 

postulated PTS events and the Alloy 82/182 butt welds that are subject to PWSCC. 
A-3

VIC CHAPMAN 

O J V CONSULTANCY, LTD. 

DERBY, ENGLAND

Mr. Chapman has a Diploma in Engineering and an Honors Degree in mathematics. He has worked in the 

pressure system design and maintenance field for over thirty years. Over this period he has gained 

experience in the design and fracture analysis of pressure vessels, the study of material properties and the 

statistical interpretation of data. Over the past twenty-five years he has concentrated on the area of PFM, 

risk based decision-making, risk based inspection and inspection qualification. He was responsible for 

the introduction of the Risk-Based ISI program into the Royal Naval Nuclear Fleet. 

Mr Chapman retired from Rolls Royce (Naval Nuclear Division) four years ago and became an 

independent consultant, forming his own company, O J V Consultancy Ltd. In addition to his 

consultancy work, which is funded by many organizations and international bodies, he has played a 

leading role on several international committees: 

He is currently Chairmen of ENIQ-TGR (European Network for Inspection & Qualification-Task 

Group on Risk). 

A member of the European program NURBIM (Nuclear Risk-Based Inspection Methodology). 

A member of the ASME Research Committee on Risk Technology. 

He chaired the European program EURIS (European Network of RI-ISI). 

Was a member of the ASME Research task force on Risk Based Inspection – Developed of 

Guidelines. 

 

Was a member of the UK Technical Advisory Group on Structural Integrity (TAGSI) sub-committee 

on defects in welds. 
A-4

GUY DEBOO 

SENIOR STAFF ENGINEER 

EXELON NUCLEAR 

CHICAGO, ILLINOIS

Mr. Deboo has 28 years experience working in the nuclear power generation field. Mr. DeBoo’s recent 

experience includes fatigue, crack growth and flaw stability analyses necessary to demonstrate operability 

for most power plant components. These evaluations would include root cause and remaining life 

determinations. He has extensive experience with IGSCC and other material degradation issues. He also 

has performed and supervised functionality and operability evaluations of systems and components to 

address unanticipated operating events or conditions, which do not meet inspection or test requirements. 

During his 28 years in nuclear power generation, Mr. DeBoo has worked on three major nuclear projects 

(six units) including all design, inspection and testing phases leading to commercial operation. This

experience included system and component seismic qualification, component fatigue qualification, 

licensing and design review for fuel load, special analytical assessments of the safety significance of 

installation discrepancies, and system/equipment/component functionality/operability evaluations 

resulting from startup and operating test programs.

Mr. DeBoo has extensive experience in the evaluation of fatigue-related problems, material degradation

issues and the assessment of remaining life in vessels, piping and supports for nuclear applications. He 

has performed safety evaluations for unanticipated operating events, and has developed plant-unique 

acceptance criteria to permit continued operation. Those events include fluid transients, thermally 

stratified flows, and flow-induced vibration. He supervised the AMSE Class 1 piping fatigue analysis on 

two BWR units and two PWR units. 

Mr. Deboo has a B.S. in Mechanical Engineering from Northwestern University and a M.S. in 

Mechanical Engineering from the University of Illinois. He is a member of the ASME. Currently he is 

serving on Section XI of the ASME Boiler and Pressure Vessel Code as a member of the Working Group 

on Pipe Flaw Evaluations and as the Secretary of the Working Group on Flaw Evaluation. 
A-5

WILLIAM GALYEAN 

SENIOR PRA ANALYST 

IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LABORATORY 

IDAHO FALLS, IDAHO

Mr. Galyean has over 25 years experience in performing PRA on commercial nuclear power plants. After 

earning his Bachelor of Science degree in Physics at Millersville State College in 1976, he went to the 

University of New Mexico where he obtained his Master of Science degree in Nuclear Engineering in 

1978. 

Mr. Galyean is a senior PRA analyst at the Idaho National Engineering and Environmental Laboratory 

(INEEL), which he joined in 1986. In recent years, Mr. Galyean was the Principal Investigator of two 

large NRC-sponsored PRA-related programs. The latest entailed performing detailed statistical analyses 

on nuclear power plant reliability data collected from operating experience. This data was then used to 

estimate both the historical system reliability for operational missions actually performed and the 

expected reliability for postulated risk-significant missions. The earlier program was a research program 

aimed at supporting the resolution of USNRC Generic Issue 105, “Interfacing System Loss-of-Coolant 

Accidents at Light Water Reactors.” This program integrated many disciplines in the area of risk 

analysis, including: human reliability, thermal-hydraulics, consequences, external events, and stress 

analysis. In addition a number of innovative techniques were developed for generating human error 

probabilities and fluid system rupture probabilities. Mr. Galyean has also developed and teaches 1-week 

courses on PRA modeling techniques and on Level-2 PRA, and developed low-power and shutdown 

(LP/SD) specific PRAs as part of the Standardized Plant Analysis Risk (SPAR) model program. Other 

activities include serving on an NRC-sponsored expert panel that was formed to produce updated 

estimates of loss of coolant accident frequencies, supporting the independent validation of a PFM 

computer code (FAVOR), participating as the INEEL project manager on a multi-company, multi-lab 

project funded through the DOE NERI program performing a risk-informed assessment of new reactor 

design requirements. He has also provided significant contributions to a number of other risk/reliability 

related programs, including NUCLARR (Nuclear Computerized Library for Assessing Reactor 

Reliability), a PC-based databank of reliability data for both hardware and human actions; a reliability 

analysis of the INEEL site power distribution system; and an analysis of the risk significance of possible 

operator actions for managing severe accidents at a commercial nuclear power plant. 

Prior to joining INEEL Mr. Galyean worked for Falcon Research and Development Company, Science 

Application International Corporation, and the NUS Corporation. 

Mr. Galyean is a member of the American Nuclear Society, and has served on an International Atomic 

Energy Agency (IAEA) review teams (International Peer Review Service – IPERS) that reviewed a 

Level-1 PRA on the Borssele NPP (Dutch) and a Level-2 PRA on the Krsko NPP (Slovenian). 
A-6

KAREN GOTT 

SWEDISH NUCLEAR POWER INSPECTORATE 

Dr. Gott studied metallurgy and materials science at Imperial College, London. 

During the more than 20 years she has worked in Studsvik Dr. Gott studied many aspects of the 

environmental effects on structural materials in nuclear power plants, both through contract research 

projects and failure analysis. She has held a number of different types of position whilst at Studsvik

including project manager, marketing manager and manger of the reactor chemistry group. She was also 

on periodic loan to a US subsidiary in Richland, WA, to help them establish laboratory support for their 

decontamination services. 

The main areas of her research activities were 

• Creep crack formation in stainless steels (mechanical testing, electron and light optical 

metallography) 

• Fracture mechanics (corrosion fatigue, residual stress measurement, non-destructive testing) 

• Reactor chemistry (PWR and BWR chemistry, activity build-up including field measurements, 

decontamination) 

• Reactor materials (surveillance testing, failure analysis, metallography of Inconel 182) 

In her current position at the Swedish Nuclear Power Inspectorate she has continued to work in the field 

of environmental degradation of nuclear power plant structural materials. The work covers both the 

regulatory and the research aspects. On the regulatory side she is involved in the development of 

regulations, inspection and safety evaluations that form the basis for decisions based on Swedish law and 

regulations. One of her responsibilities includes the management of the materials and chemistry research 

area for the Inspectorate. In addition she has built a database covering operationally induced failures and 

damage to mechanical components in the Swedish nuclear fleet and is responsible for its maintenance and 

the associated analysis of failure cases. In 2003 she was on a six month job rotation to the Materials

Engineering Branch of the NRC’s Office of Nuclear Reactor Regulation working amongst other things on 

PWSCC problems. 

She is a member of the international conference committee which arranges the regular water chemistry 

conferences in the nuclear field, and has also acted on the international committee for the Fontevraud

conference in France. She served as chairperson of the steering committees of two large international

projects concerning irradiation assisted SCC and the establishment of a pipe failure database. 
A-7

DAVID HARRIS 

PRINCIPAL ENGINEER AND VICE-PRESIDENT 

ENGINEERING MECHANICS TECHNOLOGY 

SAN JOSE, CALIFORNIA

Dr. Harris is a principal engineer at Engineering Mechanics Technology (EMT), Inc. and has some 30 

years of experience in fracture mechanics and solid mechanics analysis and applications. His background 

is in mechanical engineering, and he has extensive experience in probabilistic structural mechanics, 

especially as related to fracture mechanics. 

Dr. Harris began his career as a mechanical engineer at Lawrence Radiation Laboratory (LRL) in 

Livermore, California. After several years at LRL, Dr. Harris joined one of the earliest vendors of 

acoustic emission instrumentation, Dunegan Corporation, as Director of Research. After four years at 

Dunegan Corporation, Dr. Harris joined Science Application, Inc. (SAI, now known as SAIC) in their 

Palo Alto office. During his seven years at SAI, Dr. Harris’ efforts included performing some of the 

earliest applications of PFM to nuclear reactor piping. He was the principal developer of the PRAISE 

code, which was developed for the USNRC. The PRAISE code is based on PFM and is one of the most 

widely applied tools for evaluation of the reliability of weldments in nuclear reactor piping. 

Dr. Harris worked at Failure Analysis Associates for over ten years. During this time he developed and 

applied fracture mechanics to a wide variety of problems, ranging from railroad wheels to rocket ship 

engines. These efforts included both deterministic and probabilistic aspects, and involved both computer 

software development and applications to industrial problems. He was the manager of the Fracture 

Mechanics section, which included some five engineers involved in fracture mechanics and related finite 

element stress analysis. He was the principal developer of the NASCRAC code, which is a general 

purpose code for deterministic analysis of crack growth that was developed for NASA. 

Dr. Harris is currently a vice-president and principal engineer at EMT a company that he was involved in 

founding some seven years ago. EMT is an engineering consulting firm that specializes in fracture 

mechanics, life prediction and related software – both deterministic and probabilistic. Efforts at EMT 

include development of the PRAISE code in Windows (WinPRAISE), including enhancements to make 

the software easier to use in routine applications, and expansion of PRAISE to include crack initiation 

due to cyclic loading in air and water environments. He was also involved in the development of 

commercial fracture mechanics software – including linear and nonlinear SmartCrack. BLESS is a code 

for analysis of reliability of headers and piping in fossil-fired power plants that was developed with

support of the Electric Power Research Institute (EPRI). BLESS is a physics-based model that considers 

both crack initiation and growth due to creep and cyclic loading. 

Dr. Harris has been involved in ASME activities related to reliability considerations in design and 

inspection of nuclear reactor piping. He was an original member of the ASME Research Task Force on 

Risk-Based Inspection Guidelines, and was the editor of Volume 3 of a series of reports published by this 

committee. Volume 3 was on applications to fossil fired power plants. He is currently vice chairman of 

the Risk Technology Committee of the ASME. Dr. Harris is a member of ASTM as well as ASME. He 

has nearly 100 publications in the open literature, primarily in the areas of acoustic emission and fracture 

mechanics. He received a B.S. and M.S. in mechanical engineering from the University of Washington 

and a Ph.D. in applied mechanics from Stanford University. 
A-8

BENGT LYDELL 

SUPERVISOR 

ERIN ENGINEERING AND RESEARCH, INC. 

WALNUT CREEK, CALIFORNIA

Mr. Lydell has 30 years of risk and reliability analysis experience. Prior to joining ERIN, he held 

positions with the Swedish Nuclear Power Inspectorate (SKI), Pickard, Lowe and Garrick, Inc., and NUS 

Corporation. Mr. Lydell has extensive, practical experience with applied quantitative risk assessment. In 

various capacities (systems analyst, human reliability analyst, independent reviewer), he has supported 

numerous domestic and foreign PSA projects (Level 1 and 2, and internal flooding). As an independent 

contractor, during the period 1993-99 he performed R&D in piping reliability analysis for the oil and gas 

and nuclear industries. This work explored field experience data and its role in quantitative piping 

reliability analysis, including the interfaces between PSA requirements and PFM. The SKI pipe failure

database resulted from this work. Under contract to SKI and BKAB (a Swedish utility), during 1998-99 

he performed a pilot LOCA-frequency study; a summary report is published as SKI Report 98:30 (May 

1999). This particular study was commissioned to address the feasibility of applying BWR pipe 

operational experience data to the estimation of plant-specific LOCA frequencies. The SKI pipe failure 

database formed the basis for the OECD Nuclear Energy Agency’s “OECD Pipe Failure Data Exchange” 

Project (OPDE), an international forum for the exchange of pipe failure information. Managed by Mr. 

Lydell, a clearinghouse is operating the OPDE database and provides the quality assurance function. 
A-9

SAM RANGANATH 

XGEN ENGINEERING 

SAN JOSE, CALIFORNIA

Dr. Ranganath has spent 30 years working with BWRs. He spent over 28 years working on BWRs at 

General Electric (GE) before moving to set up a consulting company - XGEN Engineering that provides 

fracture mechanics, materials and stress analysis services to the power industry. His last position at GE 

Nuclear Energy was Engineering Fellow and Manager, Hardware Design. He has also taught graduate 

courses in structural mechanics and materials at Santa Clara University and San Jose State University for 

over 10 years 

Dr. Ranganath has a Ph. D in Engineering from Brown University and a Masters degree in Business 

Administration from Santa Clara University. He is a Fellow of the ASME. He has also been an 

Engineering Fellow at GE Nuclear Energy and was elected to the Engineering Hall of Fame at GE 

Nuclear Energy. 

Dr. Ranganath has been active in the development of the ASME Code for over 20 years. He led the effort 

on developing flaw acceptance rules for austenitic piping in the ASME Code. He was also played a major 

role in developing improved rules for seismic design of nuclear power plant piping. He has also been the 

principal investigator on several materials research programs at the Electric Power Research Institute. He 

has also been active in the BWR Vessel and Internals Program (BWRVIP) and has been the lead author of 

several Inspection and Evaluation documents for BWR internal components. His expertise in BWR 

issues such as IGSCC, corrosion fatigue, fracture mechanics, ASME Section XI and Section III Codes, 

repair hardware design and BWR design is important in assuring that the LOCA frequency conclusions 

reflect BWR field experience. 
A-10

PETE RICCARDELLA 

SENIOR ASSOCIATE 

STRUCTURAL INTEGRITY ASSOCIATES 

GREENWOOD VILLAGE, COLORADO 

Pete Riccardella received his PhD. from Carnegie Mellon University in 1973 and is an expert in the area 

of structural integrity of nuclear power plant components. He co-founded Structural Integrity Associates 

(SIA) in 1983, and has contributed to the diagnosis and correction of several critical industry problems, 

including: 

• Feedwater nozzle cracking in BWRs 

• Stress corrosion cracking in BWR piping and internals 

• Irradiation embrittlement of nuclear reactor vessels 

• Primary water stress corrosion cracking in PWRs 

• Turbine-generator cracking and failures. 

Dr. Riccardella has been principal investigator for a number of EPRI projects that led to advancements

and cost savings for the industry. These include the FatiguePro fatigue monitoring system, the 

RRingLife software for turbine-generator retaining ring evaluation, RI-ISI methodology for nuclear 

power plants, and several PFM applications to plant cracking issues. He has led major failure analysis 

efforts on electric utility equipment ranging from transmission towers to turbine-generator components

and has testified as an expert witness in litigation related to such failures. 

He has also been a prime mover on the ASME Nuclear ISI Code in the development of evaluation 

procedures and acceptance standards for flaws detected during inspections. In 2002 he became an 

honorary member of the ASME Section XI Subcommittee on ISI, after serving for over twenty years as a 

member of that committee. 

In 2003, Dr. Riccardella was elected a Fellow of ASME International. 
A-11

HELMUT SCHULZ 

DEPARTMENT HEAD – COMPONENT INTEGRITY 

GESELLSCHAFT FÜR ANLAGEN- UND REAKTORSICHERHEIT (GRS) 

KÖLN, GERMANY

Mr. Schulz has over 35 years of experience in nuclear engineering, structural and fracture mechanics, 

materials, and nuclear safety. At GRS he is Head of the Department of Components Integrity and was/is a 

member of various national and international advisory bodies regarding nuclear safety, component 

integrity, and codes and standards. In this role, he is responsible for the safety assessment of nuclear 

components and structures, as well as related research and verification of fracture mechanics codes for 

safety applications. 

Prior to joining GRS, he worked for Gesellschaft für Reaktorsicherheit where he held various staff 

positions and project management responsibilities for PWR safety assessment work. Prior to that he was 

with United Nuclear Corporation where he did work with fuel element design and inspection of reference

elements in US Nuclear Power Plants. He was also on staff with AEG Research Center for which he 

worked in the area of fuel element design, qualification of fabrication processes, testing programs on fuel 

elements and inspection of reference elements in nuclear pilot plants. Mr. Schulz holds a B.S. /M.S. in 

mechanical and nuclear engineering and has served on the engineering faculty at Essen. 

 
A-12

FRED SIMONEN 

LABORATORY FELLOW 

PACIFIC NORTHWEST NATIONAL LABORATORY 

RICHLAND, WASHINGTON

Since joining the Pacific Northwest National Laboratory (PNNL) in 1976, and before that at the Battelle 

Columbus Division beginning in 1966, Dr. Simonen has worked in the areas of fracture mechanics and 

structural integrity. His research has addressed the safety and reliability of nuclear pressure vessels and 

piping as well as other industrial and aerospace structures and components. 

During the 1990’s Dr. Simonen was a leader on the behalf of NRC and the ASME in the implementation 

of risk-informed methods for the inspection of nuclear piping. Dr. Simonen supported NRC staff by 

writing all chapters on piping reliability that are now part of DG-1063 Regulatory Guide on RI-ISI of 

Nuclear Power Plant Piping. These chapters provided the first formal set of guidelines to industry for 

probabilistic structural mechanics calculations for estimating piping failure probabilities. His 

recommendations impacted the selection of critical piping components that are given high priority for 

nondestructive examinations. On behalf of NRC and ASME Research, Dr. Simonen led a national effort 

during 1996-97 to benchmark PFM computer codes. The exercise concluded with PNNL performing the 

first ever statistically based calculations to quantify the uncertainties in calculated piping failure

probabilities. 

Since the early 1980’s he has led several studies for the USNRC on the effects of PTS on the failure 

probability of RPVs. This work has advanced the technology of PFM and methods for estimating the 

number and sizes of flaws in vessel welds. Dr. Simonen's research has corrected longstanding 

deficiencies in traditional methods used to estimate the number and sizes of the welding flaws that govern 

the structural reliability of high-energy reactor piping and vessels. 

Dr. Simonen was invited during 1995 and 1998 by the IAEA to meetings in Russia and Sweden to 

participate with a group of experts who evaluated the application of the LBB concept to RBMK reactors. 

The Central Research Institute of the Electric Power Industry invited Dr. Simonen to Japan during 1998

to present lectures on the reliability of reactor piping and methods to quantify the benefits of ISI 

programs. 

Dr. Simonen has published over 200 papers, articles and reports in the open literature. He is a 

member/fellow with the ASME and serves on numerous ASME committees and codes and standards 

bodies, and has been awarded a number of prestigious awards from ASME. 

Dr. Simonen holds a PhD. and Masters Degree in Engineering Mechanics from Stanford University and a 

B.S. in Mechanical Engineering from Michigan Technological University. 
A-13

GERY WILKOWSKI 

PRESIDENT 

ENGINEERING MECHANICS CORPORATION OF COLUMBUS 

COLUMBUS, OHIO 

Dr. Wilkowski is an internationally recognized expert on the fracture behavior of piping in the nuclear 

as well as oil and gas industries. His areas of expertise include: full-scale pipe and pressure vessel 

fracture testing, nondestructive examination, JR-curve testing, high-rate toughness testing, experimental 

design and instrumentation, elastic-plastic estimation scheme analysis, impact testing, ASME Section 

XI flaw analyses, LBB analyses, and pipe system fracture behavior under seismic loading. 

He was heavily involved in the development and verification of the fracture mechanics analyses for 

circumferential cracks in nuclear pipe for ASME Section XI. He was also a member of the following 

review committees: 

(1) NRC Pipe Crack Task Group member that developed the NRC LBB procedure, 

(2) NRC Peer Review Committee for proposed new seismic design rules for nuclear piping, 

(4) NRC CRDM cracking review team member, 

(5) NRC Davis-Besse clad integrity review team member, 

(6) Consultant to AECB on CANDU pressure tube guillotine break phenomena, and 

(7) Member of DOE's Peer Review Groups for: Savannah River plant, New Production Reactor plant, 

Advanced Neutron Reactor, and uranium hexafluoride storage cylinders. 

Dr. Wilkowski is a fellow of ASME. Currently he is a member of the following ASME Boiler and 

Pressure Vessel Code Section XI groups: Plant Operating Criteria Special Working Group, Flaw 

Evaluation Working Group, and Secretary of the Pipe Flaw Evaluation Working Group. He is the past 

chairman of the ASME Materials Fabrication Committee, and past chairman of the Pipe and Support 

Subcommittee of the ASME Operations, Applications, and Components Committee, all of which are 

part of the ASME Pressure Vessel and Piping Division. He was a coordinator for the 14th, 16th, and 17th

Structural Mechanics in Reactor Technology (SMiRT) Conferences. He is a registered professional 

engineer in the State of Ohio since 1979. 

Dr. Wilkowski has more than 200 technical publications, most on piping fracture. He is currently on 

the Editorial Board of the International Journal of Pressure Vessels and Piping. He is a past Associate 

Technical Editor of the ASME Journal of Pressure Vessel Technology, and guest editor of the Nuclear 

Engineering and Design journal. He was editor or co-editor of eleven ASME special technical 

publications. He was co-editor of four NRC Conference Proceeding Reports on LBB. 

Dr. Wilkowski has both a B.S. and M.S. degree in Mechanical Engineering from the University of 

Michigan and a PhD in Nuclear Engineering from the University of Tokyo. 

APPENDIX B 

MEETING MINUTES FROM GROUP PANEL MEETINGS

B-1

APPENDIX B 

MEETING MINUTES FROM GROUP PANEL MEETINGS 

In this appendix the meeting minutes from the three group meetings of the expert panel are 

presented. First the meeting minutes from the kick-off meeting are presented, followed by the 

meeting minutes from the base case review meeting. Lastly, the meeting minutes from the wrapup meeting of the elicitation panel are presented. 
B-2

MEETING NOTES FROM US NRC LOCA ELICITATION KICK-OFF

MEETING 

DOUBLETREE HOTEL, ROCKVILLE, MD 

FEBRUARY 4 – 6, 2003 

Day 1 – Tuesday, February 4, 2003

Welcoming Remarks, Agenda Review, and General Information 

Rob Tregoning of the USNRC began the meeting with a review of the agenda and general 

announcements. In addition, the individuals present were asked to introduce themselves with a short 

background of their experience related to the issue of LOCA frequency estimations. 

Mike Mayfield of the USNRC welcomed the group and offered his perspective on the subject. Hossein 

Hamzehee of the USNRC also stressed the importance of the LOCA frequency determination for the 

continuing effort to explore risk-informed revision of 10 CFR 50.46, which govern ECCS requirements. 

The meeting attendance list was provided to all of the meeting participants. 

Presentation: Importance of LOCA Distributions to 50.46 

The first presentation on the agenda was made by Alan Kuritzky of the USNRC. Alan laid out the 

importance of LOCA frequency estimates with respect to the 50.46 revision effort. Some of the key 

points from his presentation and subsequent discussion are outlined below: 

• The NRC staff proposed a plan for risk-informing the technical requirements of 10 CFR Part 50 

(Option 3) in SECY 99-264. 

• Stakeholder input was considered in the recommendation to focus revision on the ECCS 

requirements. 

• These requirements are covered in three regulations: 10 CFR 50.46, Appendix K to 10 CFR 

50.46, and General Design Criterion (GDC) 35. 

• Potential changes to 10 CFR 50.46 fall in one of the following areas 

• ECCS reliability (one of the focuses of elicitation – due to the simultaneous Loss of Offsite 

Power [LOOP] requirement) 

• ECCS acceptance criteria 

• ECCS evaluation model 

• ECCS LOCA size definition (another focus of the elicitation) 

• The elicitation results will impact changes to the ECCS reliability areas and the ECCS LOCA 

size definition. 

• ECCS reliability is primarily impacted because of the effort to eliminate the simultaneous LOCALOOP requirement. This has two pieces: 

• LOCA initiation frequencies and 

• Conditional probability of LOOP given a LOCA. 

The focus of the current expert elicitation effort will be to obtain robust LOCA frequencies for use in the 

LOCA-LOOP evaluation. Interim LOCA frequencies were developed last spring as part of an internal 
B-3

NRC elicitation effort and these have been used to demonstrate the technical feasibility of a change in this 

requirement. The results of this interim NRC effort will not be made available to this panel until after the 

wrap-up meeting to ensure that the panel results are independent. The objective is to have this project 

completed by December 2003. More detail on this topic is provided in Rob Tregoning’s subsequent 

presentation. 

For the LOCA size definition effort, a computational code is being developed to incorporate LB LOCA 

contributions from pipe breaks and other component failures. The results of this panel will be used to 

normalize the analytical code results and also provide distributions for important input variables. The 

targeted completion date for this technical feasibility study is July 2004. More detail on this topic is also 

provided subsequently. 

Rob Tregoning indicated that the LOCA frequencies would also be used within the 10 CFR 50.61 riskinformed revision effort (PTS Rule) to ensure that current calculations are acceptable. 

Presentation: Current LOCA Frequencies and Failure Mechanisms 

The next presentation was made by Bill Galyean of INEEL in which he reviewed Appendix J of 

NUREG/CR-5750. Some of the key points from his presentation and subsequent discussion include: 

• There are a number of varieties of LOCA initiating events, including: 

• Traditional pipe break LOCAs 

• Stuck open PORVs and SRVs 

• Steam generator tube ruptures 

• Reactor coolant pump seal failures 

• Interfacing system LOCAs (ISLOCAs) – where primary system coolant is inadvertently 

introduced into the secondary side piping and a secondary pipe fails creating a leak path of 

primary coolant outside containment 

• Reactor vessel rupture 

• While failure data exists for some of those categories, data for pipe break LOCAs and other 

similar events simply does not exist because it has never occurred. 

• There is methodology for estimating the frequency of an event that has never occurred. A 

Bayesian update of a non-informative prior can be employed. This assumes that the mean value 

for the distribution is 1⁄2 of a failure over the service life. This can be result in a very conservative 

estimate because the assumed failure frequency in the prior is so high (pf = 0.5). If the failure 

rate is not constant over time, one also needs to account for time dependency and this 

methodology is not equipped for this. 

• A primary Appendix J assumption is that you needed a leak before you can get a break. A 

conditional pipe break probability given a leak was based on the Beliczey-Schulz correlation. 

• There was also a presentation of passive LOCA failures that can occur in non-piping systems as 

well as a list of possible data sources for this information. 

Discussion: The elicitation panel discussed the validity of this assumption for degradation mechanisms 

that result in long surface flaws which are not as likely to leak prior to failure. Also, the expectation is 

that leaking flaws will be fixed after they are discovered during a plant walkdown or through other leak 

detection methods. 

Discussion: Bruce Bishop indicated the need for very clear definitions of what constitutes a large, 

medium, small, and very small break LOCA. The concern is that the system response to a DEGB where 

the flow rates can reach 860,000 gpm (3,250,000 lpm) (according to Westinghouse calculations) is very 
B-4

different from a 5,000 gpm (19,000 lpm) leak which is also often characterized as a large break LOCA. 

Rob Tregoning indicated that clear definitions will be developed as part of this exercise. 

Discussion: Gery Wilkowski relayed information provided by Helmut Schulz that the Beliczey and 

Schulz correlation of conditional probability of a rupture given a leak was developed for cyclic fatigue 

crack growth. 

Discussion: Rob Tregoning emphasized that Bill Galyean’s presentation was provided to recap the last 

NRC-sponsored work in this area. This NUREG/CR-5750, Appendix J approach is not endorsed for the 

expert elicitation process; however, it represents one manner in which LOCA frequencies have been 

developed. Tregoning also emphasized that because substantial LOCAs have not occurred, past operating

experience data needs to be augmented by information from other areas. If information was available 

simply from operating experience, there would be no need for the elicitation. 

Discussion: The point was also raised that the panel needs to consider LOCA sources other than 

traditional pipe LOCAs. 

Presentation: LOCA Frequency Determination Using Expert Elicitation 

Rob Tregoning then made a presentation on LOCA frequency determination using expert elicitation. The 

objectives of this presentation are to motivate the expert elicitation effort, discuss the limitations of 

relying solely on past operating experience; present ongoing NRC-sponsored research in this area; define 

the objective of the expert elicitation; outline the approach; and discuss the structure of the kick-off 

meeting. Some of the specific key points from his presentation and subsequent discussion are outlined

below: 

 

• The LB LOCA design basis size will be determined by considering all relevant LOCA sources. 

A PFM-based model is under development for predicting LOCA contributions as a function of 

break size. This code will also account for LOCAs from non-piping sources, and will include 

contributions from future unknown failure mechanisms. Expert elicitation input from this panel 

will be used throughout program development. 

• The objectives of the elicitation are to 

• Develop future SB, MB, and LB LOCA frequency estimates extending up through the end of 

the plant-license-renewal period (approximately 35 years). 

• Develop benchmark problems and standardized inputs for conducting PFM simulations of LB 

LOCA events in important BWR and PWR systems. 

• The elicitation approach will construct base cases. Quantitative LOCA estimates as a function of 

break size will be developed for these base cases. Then, important variables and issues will be 

discussed within the relative framework of the base cases. 

Discussion: It was stressed that the base cases will just provide a reference point. Adjustments to the 

base cases will account for the impact of those issues which contribute significantly to the LOCA 

frequencies. These adjustments must consider their effect on current LOCA frequencies and their time 

dependence up through the end of the plant license-renewal period (≈ 35 years). 

• The programmatic approach for the elicitation was presented. It consists of the following 

important areas. 
B-5

• Conduct the kick-off meeting 

• Develop elicitation questions 

• Allow individual study to develop answers for elicitation questions 

• Conduct the individual elicitations 

• Analyze results (facilitation team) 

• Conduct a wrap-up meeting 

Discussion: It was conveyed that when the facilitation team queries each elicitation panel member, best 

estimate answers will be sought as well as the uncertainty in the estimates. It was also stressed that each 

panel member need not answer all questions, but only those that they feel that they are qualified and 

comfortable with answering. However, people will be encouraged to answer all questions, even in areas

outside of their specific expertise. Uncertain knowledge should be reflected in the uncertainty estimates. 

Rob Tregoning indicated that one job of the facilitation team will be to filter out responses which are not 

well-founded, and exclude them from the final analysis. 

• The principal components of the kick-off meeting were reviewed again. There are several major 

objectives which need to be accomplished during this kick-off meeting: 

• Present the elicitation objectives and define fundamental terms to ensure common 

understanding. 

• Undergo elicitation training to understand the process and approach. 

• Construct methodology for developing baseline LOCA estimates. Develop a classification 

scheme and approach for issues which could affect the baseline LOCA estimates. 

• Identify and classify issues for consideration. Discuss issues as necessary for clarification. 

• Agree on significant issues to include in the elicitation. 

• Determine the structure of the elicitation questions. 

Presentation: Expert Elicitation Process 

Lee Abramson of the US NRC spoke on the expert elicitation process that will be followed in this 

exercise. Some of the specific key points from his presentation and subsequent discussion are outlined 

below: 

• Key word is “formal” use of expert judgment. Engineers practice informal expert judgment every 

day. 

• It was emphasized that elicitation is a structured process and that the process requires experienced 

practitioners to conduct the exercise. This is not a “do it yourself” activity. 

Discussion: A question was raised if the results of this elicitation or past elicitations could be used as a 

baseline for future efforts, in much the same way that Bayesian analysis is performed. Lee Abramson 

indicated that there is no natural means of updating results from prior elicitations based on recent 

experience or new data. However, it may be appropriate to use the results of a prior elicitation as starting 

point for future elicitation. 

• The need for comprehensive documentation was also stressed to ensure that the process approach, 

issues, analysis techniques, results and uncertainties are clear. Additionally, follow-on work to 

refine the results requires comprehensive documentation in order to understand the basis of the 

initial study. 

• The need for an expert panel with a broad range of expertise and experiences was expressed. 

Also all of the stakeholders (both utilities and regulators) must be represented. 
B-6

• There are two methods of elicitation: group and individual. The problem with group sessions 

(versus individual sessions) is that often group dynamics lead to domination of one or two 

individual opinions. The results then no longer represent everyone’s input. 

• Elicitation team for this exercise consists of 

• Normative expert – Lee Abramson 

• Substantive experts – Alan Kuritzky, Ken Jaquay, Rob Tregoning, others? 

• Recorder – Paul Scott 

• Documenter – Paul Scott (could be same as recorder) 

• Panel members need to provide rationale for answers so others can see why certain panelists 

came up with certain answers. In that way other panelists have the option of changing their 

answers based on feedback from the group. The panel will largely be provided this feedback at 

the wrap-up meeting. Panel members can revise answers to any question at any time. 

Discussion: It was asked if the response will be weighted in any way to account for expertise in a given 

area. Lee Abramson replied that the analysis will use unweighted responses so that everyone’s response 

is judged equally. With this size of panel, weighting should not substantially affect the final results. The 

elicitation will also query the panel member’s uncertainty for each answer. If inordinate uncertainty

exists, then the response may be downgraded. Also, the rationale provided by each panelist will help 

determine if responses need to be weighted. 

• Types of biases present in elicitation processes: 

• Motivational biases (i.e., social pressure or group pressure to make a certain decision). These 

need to be recognized and avoided at all costs. 

• Cognitive biases --- biases can occur when people have developed an initial answer and more 

data becomes available which require the initial answer to be modified. Typically people 

underestimate the impact of the new data. This bias is referred to as anchoring. The 

elicitation structure will be developed in an attempt to minimize these biases. For instance, 

initial estimates of the total LOCA frequencies will not be asked. 

• Background biases (i.e., what an individual might see as reasonable, or would expect, based 

on his background.) For example, an experimentalist might see a high probability of failure 

of a piping based on the number of experiments he has run in which he saw a failure, but 

typically the test conditions were such that similar conditions in the field are highly unlikely 

to ever occur. This bias is natural, but it is important to get each individual to consider all 

variables which affect the result and break them down into meaningful pieces. 

• People are more than likely to underestimate the true uncertainty, by a factor of 1/2. 

• People are more likely to anchor on median value, not on the extremes. 

• Goal is to make the questions as unambiguous as possible (very precise) and to focus questions 

on the major issues affecting the LOCA analysis. 

• The uncertainty range will be queried during the elicitation by asking for the “number” such that 

there is 5% chance that the true response is less than this number. A separate number will be 

provided for the UB such that there is also a 5% change that the true response is higher than this 

number. This corresponds to the 90% coverage interval of the variable. 

• Purpose of elicitation panel members is to come up with individual answers, not a consensus. 

Discussion: There was quite a bit of discussion and confusion about the definition of the coverage 

interval. Lee Abramson said that the uncertainty range (difference between higher and lower response to 

a given question) should cover the true number for that variable 90% of the time. The true value should 

fall below the lower response 5% of the time and the true value should land above the higher response 5% 

of the time. However, Lee cautioned against making the coverage interval inordinately large just to 

capture uncertainty. If this occurs, the coverage interval contains little useable information. 
B-7

Elicitation Exercise 

Each participant filled out an elicitation questionnaire dealing with age related health issues. The results 

from this exercise were reviewed with the meeting participants on Wednesday morning. As part of this 

exercise, Lee Abramson indicated that it is usually easier to determine relative rates versus absolute rates. 

Various absolute and relative questions were posed in order to demonstrate this concept. 

Definition of Terms for Elicitation 

Terms used during the elicitation must be commonly understood by the group in order to foster 

discussion, issue development, and subsequent elicitation. Certain key terms must be defined. 

Rob Tregoning indicated that, for this exercise, all definitions should be kept generic, not plant specific. 

The first term to be defined is LOCA. Rob Tregoning presented the NUREG/CR-5750, Appendix J 

definition as a starting point. This report defines a LOCA as “an unisolable breach of the Reactor Coolant 

Primary Boundary (RCPB) requiring ECCS initiation.”

The group felt that the term “unisolable” was not appropriate because the main point is to limit the scope 

to Class 1 piping. Also, the merits of the phase ECCS initiation were debated because the ECCS response 

in some plants requires use of normally operating plant equipment. Therefore, some plants might require 

a large leak before implementation of standby ECCS systems. There was also a discussion on the merits

of using break instead of breach, but the term breach was determine to be more generic than break. The 

addition of the term “sudden breach” instead of just “breach” was also neglected because of the vagueness 

of the word sudden. 

The group agreed to a definition of a general LOCA as follows. A LOCA is “a breech of the reactor 

coolant pressure boundary which results in a leak rate beyond the normal makeup capacity of the 

plant”. 

The next definitions are required to determine the size classifications of LOCAs. Once again, Rob 

Tregoning presented the definitions used in NUREG/CR-5750, Appendix J as a starting point. These 

definitions were also used in NUREG-1150 and form the basis of plant PRA event trees. This document 

defined three LOCA size categories: SB, MB, and LB. The NUREG/CR-5750 definitions are as follows: 

• SB LOCA - A break that does not depressurize the reactor quickly enough for the low pressure 

systems to automatically inject and provide sufficient core cooling to prevent core damage. 

However, low capability systems (i.e., 100 to 1,500 gpm [380 to 5,700 lpm]) are sufficient to 

make up the inventory completion. For a BWR, this translates to a pipe in the primary system 

boundary with a break size less than 0.004 ft2

 (370 mm2), or a 1 inch (25 mm) equivalent inside 

pipe diameter, for liquid, and less than 0.05 ft2 (4,600 mm2), or an approximately 4 inch (100 

mm) inside diameter pipe equivalent, for steam. For a PWR, this equates to a pipe break in the 

primary system boundary with an inside diameter between 1⁄2 to 2 inches (13 to 50 mm). 

• MB LOCA – A break that does not depressurize the reactor quickly enough for the low pressure 

systems to automatically inject and provide sufficient core cooling to prevent core damage. 

However, the loss from the break is such that high capability systems (i.e., 1,500 to 5,000 gpm 

[5,700 to 19,000 lpm]) are needed to makeup the inventory depletion. For a BWR, this translates 

to a pipe in the primary system boundary with a break size between 0.004 to 0.1 ft2

 (370 to 9,300 

mm2), or an approximately 1 to 5 inches (25 to 125 mm ) inside diameter pipe equivalent, for 

liquid, and between 0.05 to 0.1 ft2 (4,600 to 9,300 mm2), or an approximately 4 to 5 inches (100 

to 125 mm) inside pipe diameter equivalent, for steam. For a PWR, this equates to a pipe break 

in the primary system boundary with an inside diameter between 2 to 6 inches (50 to 150 mm). 
B-8

• LB LOCA – A break that depressurizes the reactor to the point where the low pressure system 

injection automatically provides sufficient core cooling to prevent core damage. For a BWR, this 

translates to a pipe in the primary system boundary with a break size greater than 0.1 ft2 (9,300 

mm2), or an approximately 5 inch (125 mm) inside diameter pipe equivalent, for liquid and steam. 

For a PWR, this equates to a pipe break in the primary system boundary with an inside diameter 

greater than 6 inches (150 mm). 

The elicitation panel questioned the basis of the equivalent pipe diameter relationships to break size

provided in the NUREG/CR-5750 Appendix J. Bill Galyean thought that they could be traced back to 

NUREG 1150 and possibly WASH-1400. It was quickly determined that “break” should be replaced by 

“breech” everywhere for consistency with the general LOCA definition. Also, the group decided that the 

formal definitions should be based on leak rate, and not equivalent break area or size. 

At this point, the need for additional LOCA size classification was revisited. This request was 

promulgated by Bruce Bishop based on discussions with the Westinghouse Owner’s Group (WOG). The 

System response and mitigation procedures for a 5,000 gpm (19,000 lpm) LOCA (lower limit LB LOCA 

within NUREG/CR-5750, Appendix J) and a DEGB of the largest class 1 pipe (flow rate up to 860,000 

gpm [3,250,000 lpm] according to WOG) are significantly different. Because, the elicitation results will 

used in existing PRAs, it was also stressed that the original leak rate classifications in NUREG/CR-5750, 

Appendix J should also be maintained. While the group also agreed that the leak rate threshold should

ideally be based on the equipment needed to mitigate a specific event, this information is highly plant 

specific and could not be approximated generically.

For the reasons stated in the above paragraph, the group decided to keep the NUREG/CR-5750, Appendix 

J leak thresholds of 100 gpm (380 lpm), 1,500 gpm (5,700 lpm), and 5,000 gpm (19,000 lpm), but to add 

several leak rate categories above 5,000 gpm (19,000 lpm). The highest category was set at 500,000 gpm 

(1,900,000 lpm) to capture the DEGB events of the largest primary system pipes. Additional ranges of 

25,000 gpm (95,000 lpm) and 100,000 gpm (380,000 lpm) were chosen to span the range from 5,000 gpm 

(19,000 lpm) to 500,000 gpm (1,900,000 lpm) in roughly equivalent magnifications. These leak rate 

categories were also chosen because they tend to group DEGBs by primary system functionality. 

The LOCA size classification thresholds adopted by the group are summarized in Table B.1.11

. A 

category 1 LOCA is defined as “a breach of the reactor coolant pressure boundary which results in 

a leak rate which is greater than 100 gpm (380 lpm). Similarly, a category 6 LOCA is a breach of 

the RCPB which results in a leak rate which is greater than 500,000 gpm (1,900,000 lpm). It should 

be stressed that category 1 LOCAs include contributions from all categories. The group preferred 

the threshold classification of LOCA sizes instead of partitioning the sizes into ranges as in 

NUREG/CR-5750, Appendix J. Care will be needed during the elicitation to ensure that these 

definitions are understood. 

 

1

 The nomenclature for the table and figure numbers is such that the letter B refers to Appendix B, the first number 

(1 or 2) refers to a figure associated with either the first or second panel meeting, and the second number refers to 

the numerical sequence of that particular table or figure in the text for the applicable meeting, i.e., either first or 

second. 
B-9

Table B.1.1 LOCA Size Classification Thresholds 

Category Leak Rate 

Threshold (gpm) 

1 > 100 

2 > 1,500 

3 > 5,000 

4 > 25,000 

5 > 100,000 

6 > 500,000 

It was determined by the group that these leak rates should be roughly correlated to breach area, and 

converted into an equivalent pipe diameter so that the UB leak rates for various piping systems could be 

determined. There was some concern about the feasibility of developing generic estimates. It was 

suggested that equivalent pipe sizes could be based on 250 gpm/in2 (1.47 lpm/mm2) for liquid PWR lines 

and 175 gpm/in2 (1.03 lpm/mm2) for liquid BWR lines. However, these estimates did not agree with the 

Westinghouse equivalent pipe diameter estimates. 

Presentation: SKI-PIPE Database: Background - Structure - Status - Applications (1994 - 

2002) 

This presentation by Bengt Lydell discussed the SKI-PIPE database evolution and background; the 

database structure and content; current database status; and LOCA frequency estimate conducted with the 

data base. Some of the specific key points from his presentation and subsequent discussion are outlined 

below: 

• Background: The database was motivated to create a tool that would serve both PRA and the 

PFM/material science practitioners. It’s structured to provide information to completely define 

the piping systems attributes (design characteristics) and the influence functions (operating 

history) which govern system failure probability. By thoroughly assessing these features it is 

possible to determine plant specific estimates of piping system reliability. 

• Structure and Content: The database covers pipe failures in commercial nuclear power plants 

from 1970 to the present. 

• It should be stressed that SKI-PIPE only includes failures in piping systems, external to the RPV. 

Non-piping system failures are not included. Also, SKI-PIPE contains only passive piping 

failures of metallic piping. 

• A pipe failure is defined in the database as any degradation that results in piping repair or 

replacement. 

• Each record in the database is indexed. References to the original data source (e.g., LER 

report) and supporting information are provided. All the supporting documented is stored 

electronically. 

• The database is organized by reliability attributes (i.e. design features such as material, 

dimensions) and influence factors (i.e. unique service conditions, including degradation 

susceptibility). 

• When the original record is incomplete (such as an LER), a best effort is made to fill in 

database gaps by directly contacting the plant operators. 

• It is noted in the database when each record consists of multiple flaws at a single component 

location. However, subsequent data entries are typically associated with only the largest flaw 

at that location. 
B-10

• The database includes both surface penetrating flaws and non surface penetrating flaws (i.e., 

embedded flaws). 

• Current Database Status: 

• The database is continually being updated. 

• The current OECD-sponsored OPDE project has participants from 12 nations. The first year 

of the three year effort is concerned with adding and validating database entries for each of 

the member countries from 1998 through 2001. 

• Raw data is currently obtained from over 40 different sources 

• Applications: Two relevant studies are the determination of LOCA frequencies for the 

Barsebäck-1 plant and examination of IGSCC in Russian graphite moderated reactors (RBMK). 

• The Barsebäck-1 study employed plant-specific attribute and influence functions which were 

comprehensively developed for all “known and credible” degradation mechanisms. 

• The Beliczey and Schulz conditional rupture probability was not used in the Barsebäck-1 

analysis. Instead a Bayesian update of a Jeffrey’s modified non-informative prior was 

employed. 

• The database results have been compared with PFM predictions for welds in certain systems 

with some success. 

• The RBMK studied indicated that the experience today with IGSCC in Russia is similar to 

US BWR IGSCC cracking experience in the late 70’s to early 80’s, before wide-spread 

mitigation was adopted 

Discussion: The panel asked if they could get copies of the SKI-PIPE database. Karen Gott of SKI 

indicated that it is possible to distribute a non-proprietary version of the database. This non-proprietary 

version contains piping failures thru 1998. 

Baseline LOCA Determination I 

Discussion: Rob Tregoning commented that the panel needed to define baseline LOCA frequencies in 

order to benchmark relative responses during the elicitation. He also mentioned that the SKI-PIPE 

database could be used to develop baseline frequencies if the group could develop well-defined “base 

case(s)”. The base case(s) will represent a set of conditions and physical phenomena. In theory, the

absolute LOCA frequencies associated with each base case are not important for the elicitation session

because all elicitation responses will be judged relative to the base case conditions. The absolute 

frequencies are only required to reconstruct the final results. However, the panel members decided that 

their elicitation responses might change depending on the exact LOCA frequencies associated with the 

base case conditions. That is, if a base case frequency was 10-8/year, the elicitation responses might be 

quite different than if the frequency was 10-2/year. The group therefore agreed that they will define 

rigorous conditions for each base case and also associate absolute LOCA frequencies with these 

conditions. 

Day 2 – Wednesday, February 5, 2003

Elicitation Exercise Review 

Lee Abramson reviewed the elicitation questionnaire results from Tuesday afternoon’s session. Overall, 

the results were good and consistent with expectations. The group tended to perform better on those 

questions that asked for the ratio of diseases between men in different age ranges (questions 3 and 4 in 

exercise). The mid value tended to reasonably close to the actual 2000 census values for these questions. 
B-11

Additionally, the true value was contained with the 50% interquartile region (75% - 25% percentiles) 10 

out of 12 times, or 83%, which is quite good. 

The average coverage interval for these questions was 71%. The coverage interval should theoretically be 

90%, so that group underestimated the uncertainty. Lee Abramson indicated that typically group 

uncertainty is about 1⁄2 of the true value. In other words, people tend to be more confident in their 

responses than they should be. 

The results were not quite as good when the group was asked to provide absolute disease rates for an age 

category (Question 2). The true value was inside the 50% interquartile range 4/6 times, or 67%. Also, 

the coverage interval only captured the true value 61% of the time, 10% less than for the relative 

questions. This performance demonstrates the supposition relative differences between conditions tend to 

be more accurate than absolute measures for a given condition. This will be a guiding principal in 

developing the elicitation framework. 

LOCA Issue Development 

LOCA Issue development required the group to brainstorm important LOCA issues. The group first 

defined a structure for categorizing issues in the form of a flowchart (Figure B.1.1). It was stressed that 

the LOCA frequencies in this exercise will consider only passive system failure. Active system failure 

will not be considered for the following reasons: 

1. The panel has no specific expertise in these types of failures. 

2. Failure of these components is not as rare and there is adequate data to assess their contribution to 

the LOCA frequencies. 

3. Active components are subject to ongoing maintenance which should diminish the likelihood of 

future failure rate increases. 

However, LOCA frequency contributions from active components will be combined with the passive 

component contributions to develop final LOCA estimates which can be supplied as PRA input. The 

estimation of these contributions will occur separately, but will be summarized for the panel at the wrapup meeting. 

The group divided the passive system LOCA sources (Figure B.1.1) into two classes: piping and nonpiping. Non-piping contributions include RPVs, steam generators, bolting flange failures, valves, pumps, 

etc. The distinction between piping and non-piping categorization is useful because piping has unique

issues. 

Piping LOCA Contributions 

The group first defined piping to include vessel penetrations (e.g CRDM housings, instrumentation lines), 

piping, and safe ends. The boundary between piping and non-piping components (e.g. vessels) was 

defined as the nozzle (or component) side of the safe-end/piping to nozzle weld. The group then decided 

that piping should be categorized by the specific plant system. The piping system is important because it 

defines the functionality and operating history, or influence factors. The plant system is also often

associated with specific piping designs and materials, or attribute functions. The relationship between the 

piping attribute and influence characteristics will determine its failure propensity. 

For a given plant system, the variables which affect the LOCA probability fall into one of the following 

five categories: geometry, materials, loading history, degradation mechanisms, and mitigation or 

maintenance procedures. The group decided to list all the possible contributors for each variable category 

and then link the dependencies with a given plant system. Obviously there is a synergistic effect among 

these variables. The piping system requirements result in geometrical and material selection constraints. 
B-12

Passive System LOCAs

Plant Piping Systems Component

Non-Piping LOCA 

Contribution 

Piping LOCA 

Contribution 

Degradation

Mechanisms

Mitigation/

Maintenance Geometry

Materials

Loading

History

Valves

Pressure

Vessel Pumps

Pressurizer

Steam 

Generator

Passive System LOCAs

Plant Piping Systems Component

Non-Piping LOCA 

Contribution 

Piping LOCA 

Contribution 

Degradation

Mechanisms

Mitigation/

Maintenance Geometry

Materials

Loading

History

Valves

Pressure

Vessel Pumps

Pressurizer

Steam 

Generator

Degradation

Mechanisms

Mitigation/

Maintenance Geometry

Materials

Loading

History

Degradation

Mechanisms

Mitigation/

Maintenance Geometry

Materials

Loading

History

Valves

Pressure

Vessel Pumps

Pressurizer

Steam 

Generator

Valves

Pressure

Vessel Pumps

Pressurizer

Steam 

Generator

The geometrical and material choices mesh with the system functionality and operating history to 

determine component loading history. This specific combination dictates the degradation mechanisms 

that emerge. Mitigation and maintenance procedures are developed to counteract these mechanisms. The

effectiveness of these strategies, however, is a function of all the other variables discussed. 

Material Variables: It was quickly determined that the materials category should also include 

fabrication procedures as well. Material variables are important because they give rise to specific 

degradation mechanisms. Fabrication variables can lead to variations in defect formation and residual

stress which can result in certain locations having a greater propensity for degradation. The important 

variables which affect the LOCA propensity are summarized in Table B.1.2. Table B.1.2 indicates 

whether the material is found in the base metal portion of the piping system or the welds. Issues that are 

strictly related to fabrication are separated in the table. 

The circumferential versus axial welded pipe issue was raised to differentiate between seamless and nonseamless pipe. While the bulk of current piping is seamless, there is some remaining seam welded piping 

in service. The concern is that seam-welded piping is more susceptible to degradation and leaking than is 

seamless. The shop versus field welded issue was raised due to possible weld quality differences between 

welds made in controlled conditions during piping fabrication and those made on site during piping 

system assembly. Finally, it was noted that defects, residual stress irregularities, and poor material 

properties can be associated with repair welding. Experience has shown that field cracking and leakage is 

often associated with repair welds. 

Figure B.1.1 Passive LOCA Contributions 
B-13

Table B.1.2 Piping Material and Fabrication Variables 

Materials Base metal Weld/HAZ 

 304, 304L, and 304NG stainless steel X X 

 316, 316L, and 316NG stainless steel X X 

 Carbon steel X X 

 Stainless steel clad carbon X X 

 Alloy 600 X 

 Alloy 82/182 X 

 Cast stainless steel X 

 Stainless steel bimetallic X 

Fabrication Issues 

 Circumferential vs. axial welded pipe X 

 Shop vs. field welds X 

 Repair welds X 

Geometric Variables: Table B.1.3 lists the geometric variables which influence the LOCA frequency 

distributions. These variables affect piping stress, system compliance, the propensity for a given 

degradation mechanism, and the likelihood of leaking versus catastrophic rupture. Many of the variables 

are obvious and include general system information such as piping diameter and thickness (NPS and 

schedule), the number of welds and their location, the types and numbers of specific piping components, 

and the layout of supports and snubbers. The system configuration is related to the layout, but also 

specifically considers where active components such as pumps, valves, and flow offices are located. The 

variable “connections” in this table are meant to distinguish between welded connections which are more 

typical and flanged connections which can segregate piping from active components. 

Table B.1.3 Geometric Piping Variables 

NPS and Schedule (diameter and thickness) 

Component Type (Elbow, tee, fittings, straight pipe, reducers, sockets, safe ends, end caps, etc. 

Field fabricated vs. drawn 

Crevice weld (thermal sleeves) 

Number of welds & location 

Configuration (active components, flow orifices, etc.) 

Layout including locations and types of supports and snubbers 

Connections (bolted flanges) 

The final two variables in this table are a combination of geometric and material/fabrication issues: the 

existence of crevice welds (e.g. at thermal sleeves), and the difference between drawn piping and piping 

which is field-fabricated. Drawn piping is cold worked to size and is usually limited to smaller diameter 

piping (< xx” NPS). Field-fabricated piping may be hot forged or cold work to some extent, but the final 

size is often achieved by machining. This is more typical for larger diameter pipes (> yy”). Crevice or 

filet welds are for sleeves and other piping system attachments. Partial penetration welding can result in a 

greater propensity for flaws and higher residual stresses and constraint than through-wall welding. 

Material Degradation Mechanisms: The next variable group describes material degradation 

mechanisms. As mentioned earlier, the degradation mechanisms are often associated with the piping 

system material. Each material is susceptible to each of the mechanisms listed, although the degree of 

susceptibility will greatly vary. The system loading history can also favor certain mechanisms. Table 

B.1.4 summarizes the mechanisms developed during the group brainstorming session. The mechanisms 
B-14

are segregated by the primary mechanism type. Additional sub-categories are used to identify either 

specific degradation mechanisms under the appropriate main category or features associated with the 

main category. 

Table B.1.4 Material Degradation Mechanisms 

Main Category SubCategory 1 

SubCategory 2 SubCategory 3 SubCategory 4 

Low Cycle Thermal Fatigue Crack 

Initiation 

Crack Growth 

High Cycle Mechanical Fatigue Vibration Pressure Temperature 

Stress Corrosion Cracking IGSCC TGSCC ECSCC PWSCC 

Localized Corrosion Pitting Crevice 

Corrosion 

 

General Corrosion Boric Acid 

(ID or OD) 

 

Fretting Wear 

Material Aging Thermal Dynamic Radiation Creep 

Fabrication Defects and Repair 

 

Hydrogen Embrittlement 

Flow Sensitive Erosion/ 

Cavitation 

FAC 

Unanticipated (New) 

Mechanisms 

Fatigue degradation was separated into low cycle fatigue which is primarily driven by thermal loading 

fluctuations due to plant heat-up and cool-down cycles and high cycle mechanical fatigue which could 

result from general loading fluctuation on the piping. This loading fluctuation could be induced by 

vibration, pressure or temperature fluctuation (e.g. striping). High and low cyclic loading are often

differentiated with respect to the loading magnitude relative to yield strength or number of cycles. For 

this elicitation, a rule of thumb differentiation of 1,000 cycles is sufficient to differentiate between lowcycle and high-cycle events. Both crack initiation and crack growth portions of life are important 

contributors to fatigue life, although crack initiation occupies a greater percentage of life in high cycle 

fatigue. 

Stress corrosion cracking (SCC) is listed as a main category and it includes IGSCC which was prevalent

in BWRs in the late 70’s, TGSCC which affects casting components, PWSCC which has more recently 

surfaced in PWRs, and ECSCC. The localized corrosion category includes both general pitting and 

crevice corrosion which is likely in tight, stagnate areas. While general corrosion is listed as its own 

category, boric acid corrosion is the principal contributor. Both internal (ID) and external (OD) boric acid 

corrosion are included in this sub-category. External corrosion can result from leaking fluid from another 
B-15

component in the plant which then impacts the pipe. Fretting wear describes material erosion usually 

resulting from external contact with other components. This wear usually occurs with vibration loading 

but should be distinguished from high cycle fatigue which results in crack formation. 

Material aging relates to changes in the intrinsic quasi-static material properties at the time when the 

component is placed into service. These properties include constitutive properties (stress/strain behavior), 

cyclic, crack growth resistance (da/dN versus ∆K), and resistance to crack initiation and tearing under 

monotonic loading (J-R behavior). Aging can occur due to thermal or radiation embrittlement; long term, 

or low temperature creep. Dynamic strain aging refers to the stress/strain and J-R curve resistance 

changes which result from dynamically applied loading. It is most evident in the ductile to brittle 

transition temperature shift of ferritic steels. Hydrogen embrittlement is somewhat related to the other 

material aging mechanisms in that strength and toughness can be affected over time. The distinction is 

that hydrogen embrittlement only becomes prevalent after a crack has formed. The other aging 

degradations do not require a preexisting defect although the system impact is certainly enhanced in their 

presence. 

The flow sensitive category is used to capture those mechanisms which are sensitive to the flow 

characteristics of the piping medium. The erosion/cavitation sub-category refers to material erosion due 

to cavitation which occurs when vapor bubbles collapse. This is distinct from FAC where downstream 

turbulence results in erosion/corrosion that is not accompanied by low pressure boiling. The category of 

unanticipated or new mechanisms covers those mechanisms which could surface in the future which are 

either unknown at the present time or not deemed to be important at this time. This category is purposely 

vague to capture the panel’s general uncertainty of the completeness of our understanding of future piping 

degradation mechanisms. For instance, just a few years ago, PWSCC would not likely have been 

considered to be an important degradation mechanism. Now however, it is of primary concern. 

Fabrication defects and the repair of those defects (or lack thereof) are distinct from repair welding

mentioned earlier in the material/fabrication issue table. This issue covers the likelihood of repair of 

fabrication defects and the possibility that these fabrication defects could lead to failure due to one of the 

other mechanisms listed in Table B.1.4. The repair component of this issue only considers the possibility 

that these defects are repaired (or not), not any new defects generated by the repair process. Defects 

generated by repair welds have been captured in the geometric variability section (Table B.1.2). 

Loading History Variables: The next variable which contributes to the piping LOCA frequency 

estimates is the system loading history. The term loading history considers both the magnitude and 

frequency of the loading applied to the piping system over its service history. The different types of 

applied loading are summarized in Table B.1.4. Again, as in Table B.1.4, the loading variables are 

divided into main and sub categories. 

The thermal loading category considers loading from differential expansion between dissimilar piping 

materials. This loading is potentially exacerbated at the ends of the piping if connected to a rigid 

component (e.g. steam generator or RPV). This loading is categorized as restraint-free expansion in 

Table B.1.4. Radial thermal gradients are induced in piping due to the temperature difference between

the pipe ID and OD. Insulation can diminish this gradient. Thermal stratification can occur under low 

flow rate conditions when hotter liquid flows on top of cooler water. Boundary layer fluctuations in this 

interface can induce thermal cyclic loading which is referred to as thermal striping. 
B-16

Table B.1.5 Loading History Variables 

Main 

Category 

SubCategory 1 SubCategory 2 SubCategory 3 SubCategory 4 SubCategory 5 

Thermal Differential 

Expansion 

Restraint Free 

Expansion 

Radial 

Gradient 

Stratification Striping 

Water 

Hammer 

Steam 

Hammer 

 

Seismic Inertial Displacement 

Pressure Normal Transients 

Residual 

Stress 

Design Repair welds Fabrication MitigationInduced 

Dead Weight 

Loading 

 

SRV Loading 

Overload 

(Ext. and Int.) 

Pipe Whip Jet 

Impingement 

Deflagration 

Support Snubber 

malfunction 

Hanger 

Misadjust. 

 

Vibration Mechanical Cavitation 

Water hammer, dead weight, and SRV loading are considered as separate main categories with no 

corresponding sub-categories. Water hammer is distinguished from other pressure transients because of

its potential severity. Safety relief valve loading describes the pressure transient which occurs when 

SRVs are opened or closed. Dead weight loading is contributed by the weight of the pipe and any 

unsupported attachments. The pressure loading category includes normal operating pressure and any 

pressure transients other than those specifically listed in Table B.1.5 (e.g. water hammer, SRV loading, 

internal overloads, and cavitation). 

Residual stress is a prominent loading category and includes contributions from locally-induced welding 

process stresses related to the as-designed weld (Design sub-category in Table B.1.5) and additional 

contributions due to weld repair. The Design sub-category also encompasses contributions from the pipe 

system compliance on the weld restraint during component assembly. System compliance will obviously 

influence the residual stress distribution which if formed at a particular weld joint. Fabrication residual 

stresses include cold-springing needed to align piping during plant construction. This residual stress 

contribution may not be apparent from the system design which is why it is distinguished from the Design 

sub-category. Finally, a unique sub-category is entitled “Mitigation-Induced” to account for residual

stresses which may be applied during plant operation to mitigate certain failure mechanisms. These 

stresses are induced by processes like weld overlay repairs (used during IGSCC mitigation) and 

mechanical stress improvement (applied for VC Summer). These stresses are certainly associated with 

the weld joint being treated, but also affect the residual stress distribution in surrounding welds and 

piping. 

Overloads can result from external failures in other plant systems and internal accidents. These accidents 

can result in loading which is potentially in excess of the structural design limits of the piping. External 

overloads include those induced by pipe whip and jet impingement. Both of these categories require 

failure outside of the reactor pressure boundary as a precursor event. The loading on the reactor pressure 

boundary piping occurs either by pipe whip of the failed components or through water or steam jet 

impingement caused by the breech in the other system. A special type of internal overload sub-category 

is Deflagration. This describes the loading due to hydrogen combustion which occurred at the Hamaoka 
B-17

and Brunsbuettal plants. This category would cover not only direct failure of pressure boundary piping, 

but also precursor failure of secondary piping that leads to conditional failure (from shrapnel or jet

impingement) of pressure boundary piping. 

Another loading history category is support structure loading. This includes loading due to a 

malfunctioning snubber or misadjusted hanger that leads to piping system loading that is beyond the 

intended design limits. Vibration loading includes classical displacement loading due to nearby active 

component vibration (Mechanical sub-category) and loading due to cavitation which may exist within the

piping system. 

The final loading category is seismic which will be treated uniquely from the other loading variables 

listed. The seismic loading category includes both inertial and displacement loading components. 

Seismic loading is treated separately within PRA models and the LOCA contribution from seismic 

loading is also calculated separately. However, many analysts frequently calculate conditional failure 

probabilities due to seismic loading and this is often the principal transient of interest in most piping 

systems. The panel therefore seemed generally comfortable with considering the possible effects of 

seismic loading if the loading magnitude was specified. For these reasons, it was decided to query the 

panel about seismic effects separate from any other loading history contributions to LOCA. The seismic 

contributions can then be segregated from the final results and used to examine the effect of conditional 

seismic events on the LOCA frequency distributions.

Mitigation and Maintenance Issues: The final piping variable which was discussed separately is in the 

area of mitigation and maintenance. This is an important topic area because these procedures have been 

developed to ensure piping system integrity and prevent piping rupture. The effectiveness of any 

particular procedure is often a function of the degradation mechanism, although some issues developed 

are not specific to any mechanism. These procedures can sometimes result in unintended consequences 

which actually exacerbate the piping failure likelihood. It should be stressed that the procedures and 

issues in this table are not just concerned with current practice, but also future application and possible 

improvements. Each panel member must express his or her expectations about future LOCA performance 

up to the end of the plant license-renewal period. 

The first four mitigation procedures (Table B.1.6) are related to piping system inspection and 

maintenance. In-service inspection and RI-ISI considers both current application of these programs and 

future industry trends. Currently, more US plants are adopting RI-ISI. The effectiveness of these 

techniques to find and determine the extent of degradation is considered in this category. Often a 

technique’s effectiveness is quantified by the POD for a certain degradation mechanism. Leak detection 

considers the broad array of leak detection methods (including plant walkdowns) and their effectiveness 

in uncovering piping degradation. Online monitoring considers the effect that current system 

performance indicators (pressure, temperature, etc.) may have on preventing failures as well as future

systems that could be utilized to monitor degradation in real-time. 

Planned maintenance accounts for programs which monitor degradation and then replace piping segments 

once the degradation exceeds allowable limits. This is a popular approach for dealing with FAC in 

carbon steel piping. Planned maintenance also considers any component cleaning or preparation for 

inspections and the effect on the failure likelihood. Planned maintenance can be either beneficial or

detrimental. For instance, maintenance requires closed systems to be opened which could introduce air if 

gas blanketing is not sufficient. Maintenance can then lead to more future problems than if the 

maintenance had not occurred. 
B-18

Table B.1.6 Mitigation and Maintenance Variables 

Mitigation/Maintenance Procedures & Issues Degradation 

Mechanism 

Specific 

Plant Specific 

Variable 

ISI/RI-ISI X 

Leak Detection (Plant Walkdown) X 

Online Monitoring X 

Planned Maintenance X 

Water Chemistry X 

Decontamination X 

Internal Linings and Coatings X 

Weld Overlay X 

IHSI/MSI X 

Pipe Replacement (New Materials, New Design & Layout) X 

Improved Weld Techniques/Materials X 

Improved Inspection Techniques X 

Socket Weld Replacement X 

Plant Operating Conditions X 

Stratification Mitigation X 

Utility Safety Culture X 

Regulatory Safety Culture 

The next group of procedures is concerned with changing either the piping medium environment or the 

metal/medium interface in order to impede degradation. Water chemistry is concerned with additions or

changes in the basic water chemistry in order to reduce the degradation rate of a certain mechanism. For 

instance, hydrogenated and noble metal additions to BWR water have proven effective in impeding the 

rate of IGSCC. This category also considers fluctuations in water chemistry over the plant’s operating 

cycle and the affect that this may have on failure rates. Decontamination is related to water chemistry and 

considers the removal of impurities in the water supply and the possible impact that this could have on the 

degradation rate of certain mechanisms. 

The application of internal linings and coatings is used to segregate a susceptible piping material from the 

environment using a coating or overlay of a more resistant material. The coating or lining performs the 

same role as stainless steel cladding does in protecting carbon steel piping. Weld overlays, induction 

heating stress improvement (IHSI), and mechanical stress improvement (MSI) all attempt to change or 

relieve weld joint residual stress in order to impede crack growth. Normally, they attempt to create 

compressive residual stresses at the inside surface, or throughout the entire, piping segment. However, as 

mentioned earlier, they can also affect the residual stress in other sections of the piping. 

The next group of mitigation and maintenance procedures is concerned with anticipated future 

improvements in materials, repair techniques, and inspection methods that could reduce the likelihood of 

future LOCAs. These improvements are captured by the categories for Improved Weld 

Techniques/Materials and Improved Inspection Techniques. The Pipe Replacement category considers 

not just the removal of possibly degraded piping, but also the replacement with new materials that are less 

susceptible to known, important degradation mechanisms for a certain system. This could also be 

coupled with new design and layout configurations to reduce residual stresses and improve accessibility 

for inspections. Socket weld replacement is a specific piping replacement program that is currently being 

considered. The effectiveness and scope of its implementation still is uncertain which is the reason that 

this has been included as a separate category. 
B-19

The next two mitigation procedures are related to current and future plant operating performance. 

Thermal stratification mitigation is a specific technique employed by some plant operators in order to

improve thermal mixing and reduce stratification and striping stresses which can occur in the surge line 

and other piping systems. General plant operating performance is a category that captures all other 

similarly related issues. This would include issues such as the effect of possible power upgrades, possible 

future changes in the heat-up and cool-down cycle, the possibility of increased time periods between 

outages, etc. 

The final two issues (Table B.1.6) are related to utility and regulatory safety culture in general. There are 

many specific issues that the group lumped into these broad categories. One such issue within the utility 

safety culture is human error. Human error was defined by the expert panel as the likelihood that 

incorrect action is taken during mitigation and maintenance. This includes improper application of 

techniques and procedures, misinterpretation of obvious indications (beyond the POD included in ISI), 

and omission of a prescribed procedure. Other issues within the broad category of utility safety culture 

include the adoption and implementation of risk-informed management strategy which requires a detailed

understanding of real-time plant risk and the objective to embrace changes that reduce overall plant risk. 

The impact of economic considerations is important in terms of choosing which mitigation strategies to

pursue. All decisions weigh plant risk with economic considerations to hopefully arrive at the optimal 

mitigation strategy. However, this mitigation strategy might not lead to the lowest possible plant risk. 

Flow accelerated corrosion monitoring programs illustrate this concept. While the absolute lowest plant 

risk could be achieved for a system by replacing the pipe with FAC-resistant materials, many plants have 

chosen to monitor the degradation and replace only when the failure risk becomes unacceptable. 

Also part of the general safety culture are the lessons-learned from past problems. This experience may 

decrease the response time for mitigating future problems. For instance, the industry experience with

mitigating IGSCC in the early 1980’s may provide some useful strategies for PWSCC issues that are 

currently surfacing. Response time is generally an important mitigation concept. When degradation 

mechanisms are identified, the failure likelihood due to these mechanisms may continue to increase with 

time until effective mitigation strategies are employed which reduce their propensity. Obviously, shorter 

response times are preferred. The industry required roughly three to four years to fully implement IGSCC 

cracking mitigation strategies after the issue was fully identified. A final related issue is technology 

transfer which is the training of and knowledge transfer to the next generation of plant operators and

engineers. As the workforce continues to age and is replaced by less experienced workers, it is possible 

that plant risk may be affected. 

The regulatory safety culture also encompasses many of the same issues discussed under utility safety 

culture. Certainly lessons-learned, regulatory response time, and technology transfer equally apply to the 

regulatory culture. The regulatory environment is also affected by the agency’s interaction with the 

public and the changing public perception of risk. Management philosophy and the adoption of riskinformed regulations may also influence the regulatory safety culture. 

The group next determined if the effectiveness of specific mitigation or maintenance procedures varied as 

a function of degradation mechanisms and materials being evaluated. This dependency is reflected in the 

second column of Table B.1.6. Plant and regulatory safety culture were considered to be general issues 

which do not vary significantly with the degradation mechanism. However, the utility safety culture was 

considered by the panel to vary from plant to plant, as indicated in Table B.1.6. Regulatory safety culture 

was not determined to be a plant-specific function. 

BWR Piping Systems: The variables just discussed (geometry, materials, loading, degradation 

mechanism, and mitigation) are important for determining the overall LOCA frequencies. However, 
B-20

these variables are linked to each specific LOCA-sensitive piping system (Figure B.1.1). The next group 

task was to identify LOCA-sensitive piping and assign only pertinent variables to each system. This task 

spanned both the Wednesday (2/5) and Thursday (2/6) meeting days, but it is summarized here for 

continuity. It should also be noted that Sam Ranganath provided most of the initial input for BWRs. 

The data presented in Tables B.1.7 (for BWRs) and B.1.8 (for PWRs) summarize information developed 

by the panelists at several brainstorming sessions. Therefore, they represent panelist expertise rather than 

a systematic analysis of a database of piping system geometries. In some cases, there are minor 

discrepancies between the pipe size ranges provided and actual plant survey data. For example, a review 

of the LBB database developed by Emc2 shows that there are 16-inch diameter surge lines, such as those 

at South Texas Units 1 and 2, while Table B.1.8 indicates that the range in surge line diameters is 10 to 14 

inches. However, these relatively small discrepancies in the outer ranges of these bound have no 

significant effect on the results and conclusions which focused on developing generic LOCA frequency 

estimates. The diameters provided in this table do represent nominal pipe dimensions (e.g., 12-, 16-, 28-

inch diameter) for the sake of simplicity, rather than inside pipe dimensions which would have some 

fractional component. This convention was utilized by consensus among the panelists. 

The BWR piping systems that can result in a LOCA were first identified (Table B.1.7). These include the 

recirculation (RECIRC), feed water, steam line, high pressure (HPCS), and low pressure core spray 

(LPCS), RHR, RWCU, CRD, standby liquid control (SLC), instrument lines in both the reactor and in 

other piping systems (INST), drain lines, head spray lines, steam relief valve lines, and the reactor core 

isolation cooling (RCIC) system. It should be noted that while drain lines are associated with each 

system, they were segregated into a separate category due to their common functionality. The materials 

commonly used for the piping within each system are identified (column 2 of Table B.1.7). Similarly, the 

safe end material (column 4) and the weld material (column 5) are also indicated. The intent of this 

identification was to be comprehensive and also indicate the most prevalent materials wherever possible. 

Table abbreviations are provided at the end of the table. 

Some additional clarification is required for certain entries in this table. In the recirculation piping 

system, the safe end materials (stainless steel or Alloy 600) are furnace-sensitized during manufacture. 

The feedwater safe end is manufactured either by interference fit, butt welded, or by a triple sleeve weld 

overlay. The HPCS and LPCS contain both creviced and non-creviced welds between the piping and safe 

end. Also, the bulk of this system’s piping material is carbon steel. The CRD system consists of a 

crevice Alloy 82/182 weld to the RPV head while the stub tube (“safe end” in this system) is stainless

steel and alloy 600 which is welded and furnace sensitized. It should be noted that no stainless steel clad 

carbon steel, cast stainless steel, or bimetallic stainless steel welds were indicated in any of these systems, 

although they are listed in Table B.1.2. 

The next variable listed in Table B.1.7 is the nominal piping size present in the system (column 3 in Table 

B.1.7). This is the only geometric variable (Figure B.1.1 and Table B.1.3) indicated in Table B.1.7. It 

was not possible to do an exhaustive listing of the possible geometric variables (as with materials) due to 

the complexity and plant-specific nature of variables related to layout, configuration, weld location, and 

component type. Each expert panel member must individually determine variability and influence of 

these parameters. The rationale for listing the piping size is only to provide the panel with an indication 

of the piping size for maximum leak rate assessment. Common piping sizes for a system are separated by 

commas in Table B.1.7. Size ranges are separated by a dash and the maximum piping size is given as < 4 

inch for many of the smaller systems. It should be noted that the feedwater system typically consists of 

10 or 12 inch diameter pipes. However, a range from 12 to 24 inch is also possible. 

Significant degradation mechanisms that could be associated with piping materials in each system are 

included in the sixth column of Table B.1.7. Unanticipated mechanisms (UA) and fabrication defect and
B-21

repair (FDR) issues are present in every system. Stress corrosion cracking is listed for all stainless and 

carbon steel materials, while global corrosion (GC) is associated solely with carbon steel piping. 

Localized corrosion (LC) is included for carbon steel and stainless steel piping for all systems, except 

SLC, CRD, and instrumentation lines. Material aging (MA) was considered for the higher temperature 

lines that see constant use for both stainless and carbon steels. Flow sensitive (FS) degradation is present 

in all carbon steel piping systems with constant use. Mechanical fatigue was judged to be significant in 

all of the smaller piping systems (< 4 inch diameter). However, it was also considered important in the 

SRV and feedwater systems. Thermal fatigue (TF) was judged to be important in the feedwater, RHR, 

RWCU, HPCS/LPCS, and head spray piping. 
B-22

Table B.1.7 BWR LOCA-Sensitive Piping Systems 

System Piping 

Matls. 

Piping 

Size (in) 

Safe End 

Matls. 

Welds Sig. Degrad. 

Mechs. 

Sig. Loads. Mitigation

/Maint. 

RECIRC 304 SS, 

316 SS, 

347 SS 

4, 10, 

12, 20, 

22, 28 

304 SS, 

316 SS, 

A600* 

SS, 

NB 

UA, FDR, 

SCC, LC, 

MA 

RS, P, S, T, 

DW, SUP, 

SRV, O 

ISI w TSL, 

REM 

Feed 

Water 

CS 10, 12 

(typ), 12 

- 24 

304 SS, 

316 SS* 

CS, 

NB 

UA, FDR, 

MF, TF, FS, 

LC, GC, MA 

T, TFL, 

WH,P, S, 

SRV, RS, 

DW, O 

ISI w TSL, 

REM 

Steam 

Line 

CS – SW 18, 24, 

28 

CS CS UA, FDR, 

FS, GC, LC, 

MA 

WH, P, S, T, 

RS, DW, 

SRV, O 

ISI w TSL, 

REM 

HPCS, 

LPCI 

CS 

(bulk), 

304 SS, 

316 SS 

10, 12 304 SS, 

316 SS, 

A600* 

CS, 

SS, 

NB 

UA, FDR, 

SCC, TF, 

LC, GC, MA 

RS, T, P, S, 

DW, TS, WH, 

SUP, SRV, O 

ISI w TSL, 

REM 

RHR CS, 304 

SS, 316 

SS 

8 - 24 CS, 

304 SS, 

316 SS 

CS, 

SS, 

NB 

UA, FDR, 

SCC, TF, 

FS, LC, GC, 

MA 

RS, T, P, S, 

DW, TS, O 

SUP, SRV 

ISI w TSL, 

REM 

RWCU 304 SS, 

316 SS, 

CS 

8 – 24 CS, 

304 SS, 

316 SS 

CS, 

SS, 

NB 

UA, FDR, 

SCC, TF, 

FS, LC, GC, 

MA 

RS, TS, T, P, 

S, DW, SUP, 

SRV, O 

ISI w TSL, 

REM 

CRD 

piping 

304 SS, 

316 SS 

(low 

temp) 

< 4 Stub 

tubes – 

A600 

and SS* 

Crevice 

A182 to 

head 

UA, FDR, 

MF, SCC 

RS, T, P, S, 

DW, V, O, 

SRV 

ISI w TSL, 

REM 

SLC 304 SS, 

316 SS 

< 4 304 SS, 

316 SS 

SS, 

NB 

UA, FDR, 

MF, SCC 

RS, T, P, S, 

DW, V, O, 

SRV 

ISI w TSL, 

REM 

INST 304 SS, 

316 SS 

< 4 304 SS, 

316 SS 

SS, 

NB 

UA, FDR, 

MF, SCC, 

MA 

RS, T, P, S, 

DW, V, O, 

SRV 

ISI w TSL, 

REM 

Drain 

lines 

304 SS, 

316 SS, 

CS 

< 4 304 SS, 

316 SS, 

CS 

SS, 

NB 

UA, FDR, 

MF, SCC, 

LC, GC 

RS, T, P, S, 

DW, V, O, 

SRV 

ISI w TSL, 

REM 

Head 

spray 

304 SS, 

316 SS, 

CS 

< 4 304 SS, 

316 SS, 

CS 

SS, 

NB 

UA, FDR, 

SCC, TF, 

LC, GC 

RS, P, S, T, 

DW, SRV, O 

ISI w TSL, 

REM 

SRV 

lines 

CS 6, 8, 10, 

28 

CS CS UA, FDR, 

MF, FS, GC, 

LC, MA 

RS, P, S, T, 

DW, SRV, O 

ISI w TSL, 

REM 

RCIC 304 SS, 

316 SS, 

CS 

6, 8 304 SS, 

316 SS 

SS 

NB 

UA, FDR, 

SCC, LC, 

MA 

RS, P, S, T, 

DW, SRV, O 

ISI w TSL, 

REM 

* See note in text. 

304 SS = 304 series stainless steel 

(Table B.1.2) 

316 SS = 316 series stainless steel 

(Table B.1.2) 

A600 = Alloy 600 

CS = carbon steel 

CS – SW = seam welded carbon steel 

NB = Nickel-based weld (Alloy 82/182) 

UA = unanticipated mechanisms 
B-23

MA = material aging 

LC = local corrosion 

FDR = fabrication defect and repair 

SCC = stress corrosion cracking 

MF = mechanical fatigue 

TF = thermal fatigue 

FS = flow sensitive (inc. FAC and 

erosion/cavitation) 

ISI w TSL = Current ISI procedures 

with technical specification leakage 

detection requirements considered. 

RS = residual stress 

P = pressure 

S = Seismic 

T = Thermal 

DW = dead weight 

SUP = support loading 

SRV = SRV loading 

WH = water (and steam) hammer 

O = overload 

V = vibration 

TFL = thermal fatigue loading from striping 

TS = thermal stratification 

REM = all remaining mitigation strategies 

possible (eg. not unique to piping system)

Significant loading sources for BWR piping systems are also listed (column 7). All systems undergo 

residual stress (RS), pressure, thermal, seismic, SRV and dead weight (DW) loading. Water (or steam) 

hammer (WH) was considered to be important in the feedwater, steam line, and HPCS/LPCS systems. 

Support loading (SUP) was mainly considered to occur through the snubber support. This is important 

for the recirculation (RECIRC), RHR, RWCU, and HPCS/LPCS systems. Vibrational loading is listed 

for the smaller diameter piping systems. This loading is always coupled with mechanical fatigue 

degradation (column 6). However, vibrational loading is conspicuously absent from the feedwater and 

SRV lines which both have MF in the list of significant degradation mechanisms. 

Overloads (O) are possible for all systems, but they are likely to be external due to pipe whip, jet 

impingement, or secondary system failure. However, the drain line, CRD, instrument lines, and SLC 

were deemed to be more likely to be susceptible to internal overloads. The thermal loading was broken

down into thermal fatigue loading due to striping (TFL) in the feedwater system, and thermal fatigue 

loading due to stratification (TS) in the RHR, RWCU, and HPCS/LPCS systems. The head spray line, 

which is also judged to be TF susceptible, does not have a corresponding thermal fatigue loading source 

considered. 

There were no mitigation and maintenance procedures that were identified by the panel as being unique 

for any particular BWR piping system. Standard mitigation and maintenance for all systems is ISI with

credit given for technical specification leakage (TSL) detection. The technical specification leakage

threshold is 1 gallon per minute. The effect of all remaining (REM) mitigation and maintenance 

procedures and issues (Table B.1.6) on the LOCA likelihood should be considered by the panel. 

 
B-24

Table B.1.8 PWR LOCA-Sensitive Piping Systems 

System Piping 

Matls. 

Piping 

Size (in) 

Safe 

End 

Matls. 

Welds Sig. Degrad. 

Mechs. 

Sig. Loads. Mitigation

/Maint. 

RCP: Hot 

Leg 

304 SS, 

316 SS, 

C-SS, 

SSC-CS 

CS – SW 

30 - 44 A600, 

304 SS, 

316 SS, 

CS 

NB, 

SS, 

CS 

TF, SCC, 

MA, FDR, 

UA 

P, S, T, RS, 

DW, O, SUP 

ISI w TSL, 

REM 

RCP: 

Cold 

Leg/Cros

sover Leg 

304 SS, 

316 SS, 

C- SS, 

SSC-CS, 

CS – SW 

27 - 34 A600, 

304 SS, 

316 SS, 

CS 

NB, 

SS, 

CS 

TF, SCC, 

MA, FDR, 

UA 

P, S, T, RS, 

DW, O, SUP 

ISI w TSL, 

REM 

Surge 

line 

304 SS, 

316 SS, 

C-SS 

10 - 14 A600, 

304 SS, 

316 SS, 

NB, 

SS 

TF, SCC, 

MA, FDR, 

UA 

P, S, T, RS, 

DW, O, 

TFL, TS 

TSMIT, ISI 

w TSL, 

REM 

SIS: 

ACCUM 

304 SS, 

316 SS, 

C-SS 

2 - 12 A600, 

304 SS, 

316 SS, 

NB, 

SS 

TF, SCC, 

MA, FS, 

FDR, UA 

(FAC) 

P, S, T, RS, 

DW, O 

ISI w TSL, 

REM 

SIS: DVI 304 SS, 

316 SS 

2 - 6 A600, 

304 SS, 

316 SS, 

NB, 

SS 

TF, SCC, 

MA, FS, 

FDR, UA 

(FAC) 

P, S, T, RS, 

DW, O 

ISI w TSL, 

REM 

Drain 

line 

304 SS, 

316 SS, 

CS 

< 2” MF, TF, GC, 

LC, FDR, UA 

P, S, T, RS, 

DW, O, V, 

TFL 

ISI w TSL, 

REM 

CVCS 304 SS, 

316 SS 

2 – 8 A600 

(B&W 

and CE) 

NB SCC, TF, MF, 

FDR, UA 

P, S, T, RS, 

DW, O, V

ISI w TSL, 

REM 

RHR 304 SS, 

316 SS 

6 – 12 SCC, TF, 

MA, FDR, 

UA 

P, S, T, RS, 

DW, O, 

TFL, TS

ISI w TSL, 

REM 

SRV 

lines 

304 SS, 

316 SS 

1 – 6 TF, SCC, MF, 

FDR, UA 

P, S, T, RS, 

DW, O, 

SRV

ISI w TSL, 

REM 

PSL 304 SS, 

316 SS 

3 – 6 NB TF, SCC, 

MA, FDR, 

UA 

P, S, T, RS, 

DW, O, 

WH, TS

ISI w TSL, 

REM 

RH 304 SS, 

316 SS 

< 2 A600 MF, SCC, TF, 

FDR, UA 

P, S, T, RS, 

DW, O, V, 

TS 

ISI w TSL, 

REM 

INST 304 SS, 

316 SS 

< 2 A600 MF, SCC, TF, 

FDR, UA 

P, S, T, RS, 

DW, O, V

ISI w TSL, 

REM 

C-SS = cast stainless steel 

SSC-CS = stainless steel clad carbon steel 

FW = fretting wear 

TSMIT = thermal stratification mitigation 

HREPL = vessel head replacement 

PWR Piping Systems: LOCA-sensitive PWR piping systems were also determined (Table B.1.8) along 

with associated piping safe end, and weld materials; pipe size; significant degradation mechanisms; 

significant loading sources; and system-dependent mitigation and maintenance procedures. The format of 

this table is identical to the BWR summary table (Table B.1.7) and the abbreviations have been retained. 
B-25

Unique abbreviations are defined at the bottom of Table B.1.8. It should be noted that the group did not 

discuss broad differences between Westinghouse (W), B&W, and Combustion Engineering (CE) designs. 

In fact the only place it is noted (Table B.1.8) is in the use of A600 safe ends in the CVCS system in CE 

and B&W plants. However, any plant design distinctions may be important for certain LOCA classes and 

should be considered by each expert during analysis. 

The LOCA-sensitive PWR systems listed include reactor coolant piping (RCP) hot leg, RCP cold leg, and 

RCP crossover legs, the surge line, SIS accumulator line (ACCUM) and SIS DVI line, drain lines, CVCS, 

RHR, SRV lines, pressurizer spray lines (PSL), CRDM lines, reactor head (RH), in-core instrumentation 

(ICI), and instrumentation (INST) lines. The hot leg was segregated from the other RCP components by 

the group due to its higher operating temperature. The SIS system was divided into the ACCUM and 

DVI components to account for the piping size, material, and functionality differences. The RH group 

was intended to capture all the non-CRDM lines that penetrate the upper reactor vessel head. This 

grouping is distinct from the ICI system. The INST line grouping here considers mainly lines within 

piping systems and not the reactor. It is worth noting that this grouping is different from the grouping for 

BWRs where INST lines capture both piping and reactor lines. 

The materials utilized in PWR piping systems are similar to those in counterpart BWR systems. One 

difference is the inclusion of cast stainless steel (C-SS) in RCP, surge line, and ACCUM PWR piping. 

Also, stainless steel clad carbon steel (SSC-CS) is prominent in certain plant designs within the RCP. 

There is also less use in general of carbon steel in PWRs. It should be noted that the only material listed 

in Table B.1.2 which is not explicitly listed in either BWR (Table B.1.7) or PWR (Table B.1.8) piping 

systems is bimetallic stainless steel welds. 

The degradation mechanisms are again tied to the material and functional considerations of the piping 

system. The FDR and UA categories are included for all systems, as is thermal fatigue. Stress corrosion 

cracking was affiliated primarily with stainless steel piping, but also for carbon steel. Material aging was 

listed for several higher-temperature, constant-service piping systems (PSL, RHR, RCP, SIS) and 

mechanical fatigue was deemed important for smaller diameter piping, including the CRDM. Flow 

sensitive degradation, specifically FAC, was determined to be important in only the SIS system piping,

while fretting wear (FW) is listed only for the ICI system. 

Significant loading sources are consistent with the BWR piping sources. Pressure, seismic, thermal, RS, 

DW, and overload loading histories are sources for all systems. Smaller lines again are again considered 

to be susceptible to vibration loading and this loading is linked to the MF degradation mechanism. The 

RCP system is considered to have additional support loading contributions, mainly due to snubber 

malfunction. Both thermal stratification and thermal fatigue loading due to striping and heat-up/cooldown were listed as significant for the surge line. 

Thermal fatigue loading is also important for the RHR and drain lines according to the group, while the 

reactor head and pressurizer surge lines are influenced by thermal stratification. The PSL also must 

consider water hammer. Only the SRV lines need to consider SRV transients which is quite different that 

the BWR classification. All of the major loading variables (Table B.1.5) were considered in either BWR 

or PWR systems. However, hanger misadjustment and cavitation loading were not specifically 

mentioned. They would certainly fall under the broader loading categories listed in Table B.1.5, but may 

need to be considered individually by each expert during the elicitation. 

As with BWR piping, ISI with credit for leak detection is existent for all piping systems. All remaining 

mitigation and maintenance issues should also be considered for their effect on the LOCA frequencies. 

However, some specific mitigation procedures have been highlighted. This includes thermal stratification 

mitigation which some operators practice to limit surge line loads. Also, reactor vessel head replacement 
B-26

(HREPL) is a solution being considered to alleviate CRDM cracking concerns. The group will need to 

consider the extent and effectiveness (now and in the future) of each of these specific procedures. 

Base Case Development 

During the elicitation, each panelist will determine the biggest contributors for each LOCA frequency 

threshold category (Table B.1.1) separately for BWR and PWR plants. The information summarized in 

Tables B.1.2 – B.1.8, and discussed in the previous sections, is simply intended to identify those issues 

and variables which contribute to the LOCA frequency distributions. Each panelist will determine the 

magnitude and likelihood of each variable separately, but more importantly will determine the importance 

of the interrelationships among the variables. 

Each panel member will not be asked to provide absolute LOCA frequencies. All questions will be 

structured so that relative differences with a specific base case will be queried. The base cases will be 

associated with absolute frequencies and quantitative LOCA estimates will be derived from these values

and the relative relationships provided during the elicitation. The group spent quite a bit of time and 

effort both understanding the role of the base cases in the elicitation and assessing their importance. 

Ideally the base cases are chosen to represent significant contributing conditions to the total LOCA 

frequency estimates in order to minimize the extrapolation required during questioning. Representative, 

significant base cases should therefore theoretically improve the elicitation accuracy. Once this concept 

was understood, the group settled on several base case conditions for PWR and BWR systems. Pete 

Riccardella provided the original suggestions which were largely adopted by the panel after the analysis 

framework was clarified. The base case discussion evolved over the Wednesday and Thursday meeting 

days. All of the discussion will be summarized in this section for consistency. 

Two base cases were developed for BWR piping systems (Table B.1.9). The first case will examine the 

recirculation system piping. All the various piping sizes will be considered, and original 304 stainless 

steel material will be assumed that has not been replaced during plant operation. The safe end is noncreviced Alloy 600 which is connected to the piping and vessel by Alloy 82/182 weld material. Only the 

IGSCC (subcategory of SCC, Table B.1.2) will be considered as the degradation mechanism. The 

loading will consist of pressure, residual stress, and dead weight nominal components. Transients to be 

considered include SRV loads and seismic. The base case will assume that NWC is used in the system. 

The next BWR base case will examine the feed water system. A 12 inch diameter carbon steel pipe will 

be analyzed. The safe end and weld materials were not specified and will need to be defined by the base 

case analysis team (to be discussed subsequently). The degradation mechanisms for this base are FAC 

and TF. Loading sources include pressure, thermal, residual stresses, and dead weight nominal 

components. Thermal fatigue loading from stratification and possibly striping will provide alternating 

loads, and water flow velocities will also be included to assess fluid loading. Transients for this base case 

will include water hammer and seismic. Once again NWC will be assumed. 

Three base cases were constructed for PWR systems (Table B.1.9). The first will examine the hot leg in 

the reactor coolant piping system. A 30 inch diameter Type 304 stainless steel pipe will be considered 

with Alloy 600 safe ends and Alloy 82/182 bimetallic welds. This base case will examine thermal fatigue 

and PWSCC. Loading will again include pressure, thermal, residual stress, and dead weight nominal 

loads and thermal fatigue alternating loads. Transients will include seismic loading and a pressure pulse 

transient, the magnitude and duration of which is still to be determined. 
B-27

Table B.1.9 Base Case Analyses 

Plant 

Type 

System Piping 

Size (in) 

Piping 

Material 

Safe End 

Material 

Weld 

Material 

Degradation 

Mechanism 

Loading Mitigation/

Maint. 

BWR RECIRC 12 – 28 Original 

304 SS 

Non creviced 

A600 

A82 IGSCC P, S, RS, 

DW, SRV 

NWC, leak 

detection, 

ISI 

 Feed 

water 

12 CS FAC, TF P, S, T, RS, 

DW, WH, 

Flow 

velocities 

NWC, leak 

detection, 

ISI 

PWR RCP – 

Hot Leg 

30 304 SS A600 A82 TF, PWSCC P, S, T, RS, 

DW, pressure 

pulse 

ISI, leak 

detection 

 Surge 

Line 

10 304 SS A82 at 

Pressurizer 

TF, PWSCC P, S, T, RS, 

DW, pressure 

pulse 

ISI, leak 

detection 

 SIS: DVI 

HPI/mak

eup 

4 SS/CS TF P, S, T, RS, 

DW, pressure 

pulse 

ISI, leak 

detection 

The next PWR base case is a 10 inch diameter surge line. The surge line material is Type 304 SS and an 

Alloy 82/182 bimetallic weld will be included at the pressurizer. No safe end materials will exist. Once 

again, thermal fatigue and PWSCC will be considered. Loading will include pressure, thermal, residual

stress, and dead weight nominal loads and thermal fatigue alternating loads. Transients will include 

seismic loading and a pressure pulse transient, the magnitude and duration of which is still to be 

determined. 

The final PWR base case was the most ill-defined case because it was added after initial group 

discussions. It was added to provide a base case for a smaller diameter piping system. This base case 

will need to be defined more completely prior to analysis. A 4 inch diameter high pressure 

injection/makeup (HPI/MU) line will be examined for thermal fatigue degradation. The piping material,

welding, and safe end materials still need to be specified. Nominal loading is once again provided by

pressure, thermal, residual stress, and dead weight loads. Thermal alternating loads will be defined and 

seismic and pressure pulse transients will be considered. 

Absolute LOCA frequencies will be developed for each base case and for each threshold leak rate 

category defined in Table B.1.1. There are six leak rate categories and five base cases; therefore at least 

thirty separate calculations will be required to fully define the base case frequencies. The base cases will 

include analysis of many welds and other piping components. The LOCA frequencies for the system will 

obviously be the summation of the contributions from all system components. The frequencies will also

be determined as a function of time. Three time periods will be evaluated: 25 years after plant startup 

(current-day), 40 years after start-up (end-of-plant-license), and 60 years after start-up (end-of-plantlicense-renewal). 

The panel decided that seismic transients would be handled as part of a sensitivity study. As mentioned 

previously, seismic-induced LOCAs will not be determined as part of this elicitation for several reasons: 

the PRA models that these estimates will be used for do not consider seismic loading; there has been 

significant work in developing seismic LOCA estimates; and the group has no specific expertise in 

seismic analysis. However, many panel members are experienced in conducting analysis when the 

seismic loading history is provided. Many panel members also seem comfortable with comparing other 

loading histories to seismic events. Therefore, it was decided that the elicitation would ask for 
B-28

comparisons with conditional seismic loading (probability of occurrence of 1) separately in order to 

conduct a sensitivity analysis of the effect of seismic loading. 

Originally, the group determined that the seismic loading magnitude will be 0.3 g for the sensitivity study. 

Pete Riccardella suggested changing the 0.3 g criteria for the seismic condition to an ASME Code 

faulted-stress condition. By doing this soil conditions, damping characteristics, etc. do not have to be 

considered. While there was some discussion on the merits of this suggestion, the issue was not finalized. 

The base cases will also perform a sensitivity analysis to determine the effect of the frequencies both with 

and without ISI. Standard ISI techniques will be considered in the analysis. Credit will also be given in 

this analysis for leak detection. The leak detection threshold for the base case analysis will be leak rates 

which are commensurate with the data defined in the SKI-pipe database. The specific leak rate threshold 

associated with the database must be defined. 

A base case team was established to develop four separate LOCA frequency estimates for each of the five 

base cases. The base case team will consist of Vic Chapman, Bengt Lydell, David Harris, and Bill 

Galyean. The team will model a specific piping system and define plant operating characteristics for 

each base case. Then, the team will develop input for each of the five LOCA variables (material, 

geometry, loading, degradation mechanism, and mitigation/maintenance) within the parameter constraints

identified in Table B.1.9. The team will share information to ensure that each analysis is considering the 

same nominal conditions for each base case. 

Each base case team member will develop their own LOCA estimates for each system using whatever 

methodology they choose. The likely general approaches for each team member are summarized in Table 

B.1.10. Each specific methodology will require additional assumptions. It is incumbent that each team 

member catalog required assumptions and document the methodology used to arrive at their base case 

LOCA estimates. This information will then be rigorously, yet concisely, presented to the remaining 

panel members. This should allow each panel member to completely understand the assumptions, 

methodology, and results generated by each base case team member. It needs to be stressed that once the 

general conditions are developed, the base case members should independently develop their estimates 

without further consultation. This step is necessary to retain realistic sample uncertainty in the calculated 

results. 

Table B.1.10 Base Case Approaches 

Base Case Team Member Analysis Approach 

Vic Chapman PFM using PRODIGAL code 

Bill Galyean Direct analysis of operating experience 

David Harris PFM using PRAISE code 

Bengt Lydell Direct analysis of operating experience 

The base case team would collaborate to ensure that the PFM analyses accurately capture that leaking 

pipe operating experience. This is the one aspect of the exercise that contains plant operating experience 

data. Initial PFM calculations will be conducted based on best-estimate assumptions and the current leak 

rate frequency predictions will be compared with the operating experience. At this point, PFM input 

assumptions may be changed in order to match the operating history. Each PFM model should accurately 

document the input variables, any model changes, and results both before and after benchmarking. This

benchmarking exercise will help the remaining panel members gauge uncertainty in the calculations. 

The panel will supply background information to the base case calculation team as required. All requests 

for background information will be coordinated by Rob Tregoning to ensure proper cataloging and 
B-29

dissemination to the group. Some volunteers already offered certain background information. Bruce 

Bishop has run his own PFM models for seven Westinghouse plants that could ultimately be used to help 

verify the base case calculations. 

It was also stressed by the panel that it is important to make the PFM modeling conditions as close as

possible to the postulated service conditions so that the various base case approaches can be directly

compared to assess uncertainty and possible inaccuracies. For instance, many existing PFM models 

assume that all repairs are perfect (no defects). However, many repairs introduce new defects and most 

large flaws are associated with repairs. This fact (and other similar issues) is naturally captured within the 

operating-experience database, and needs to be considered within the PFM modeling if possible. 

Confidentiality 

A discussion was held on the confidentiality of participant’s responses during the exercise. Rob 

Tregoning indicated that all information provided as part of this exercise will remain confidential and will 

not be distributed to anyone not specifically involved in the exercise. The kick-off meeting has been

videotaped, but this will not be distributed outside of the group. Elicitation sessions will be taped for 

accuracy, but this information will also not be made public. There will be public reporting of the 

assumptions, methodology, elicitation results, and calculated LOCA frequencies that stem from this 

exercise. However, the summary reporting will only identify the names, affiliations, and possibly 

credentials of the expert elicitation panel and the facilitation team early in the report. No reference to 

individual opinions will be documented. 

Day 3 – Thursday, February 6, 2003

Use of Base Case in Elicitation 

The base cases LOCA frequencies will provide absolute LOCA estimates that each panel member will 

use to anchor the relative likelihood of LOCAs in other (non-base case) piping systems. Each panel 

member will also determine how well the base cases depict expected current and future LOCA 

performance in the piping systems that they model. It is therefore not important that a panel member 

agree with the modeling assumptions, approach, and results provided by the base case team. However, it 

is imperative that the base case development is completely understood by each panel member. Each panel

member will be able to correct perceived deficiencies in the calculated base case frequencies during the 

elicitation. Each panel member will also determine, relative to the base case results, the LOCA 

contributions of other (non base case) piping systems, and the contributions and uncertainty induced by 

the primary piping system variables. 

Reference Case Development 

In order to decompose the elicitation topics further, the group determined that it would be useful to further 

decompose non base case piping systems and variables. This was accomplished by defining a set of 

reference conditions for each LOCA-sensitive piping system identified in Tables B.1.7 and B.1.8. The 

reference conditions are similar to the base cases in that they define a unique set of conditions (materials, 

geometric variables, mitigation and maintenance procedures, and degradation mechanisms) that can be 

analyzed. They are different from the base cases in that absolute LOCA frequencies will not be 

developed for the reference cases by the base case team. The reference cases for various systems will be 

compared to determine the relative LOCA-severity among piping systems. LOCA-severity variability 

within any specific system can then be gauged with respect to the reference case for that system. It will 

be up to each panel member to determine the method for developing these relative comparisons. 
B-30

The BWR reference cases (Table B.1.11) represent specific combinations of the possible BWR piping as 

previously listed in Table B.1.7. In general, one material and degradation mechanism has been chosen is 

ideally representative of each piping system. The effect of fabrication defects and repair (discussed

earlier) should be considered for its effect on the other degradation mechanism in all cases. Nominal

pressure, thermal, RS, and DW loading should be considered for all cases. One loading transient was 

identified for each system. All the transients should be fairly well identified based on past discussion, but 

the overload transient for the control rod drive (CRD) piping needs to be better defined. In all cases, the 

snubber is considered to be functional. 

Table B.1.11 BWR Reference Case Conditions 

System Piping 

Material 

Piping 

Sizes (in) 

Safe end Welds Degradation 

Mechanisms 

Loading Mitigation and 

Maintenance 

RECIRC 304 SS 10, 12, 

20, 22, 28 

304 SS SS SCC, FDR P, T, RS, DW, 

SRV 

NWC, ISI w. TSL, 

88-01 (AI), 182 

Feed 

Water 

CS 10, 12, 12 

- 24 

304 SS CS FAC, FDR P, T, RS, DW, 

WH, TFL 

NWC, ISI w. TSL, 

88 

Steam 

Line 

CS – SW 18, 24, 28 CS CS FAC, FDR P, T, RS, DW, 

SRV 

NWC, ISI w. TSL, 

88 

HPCS, 

LPCS 

CS 10, 12 304 SS CS TF, FDR P, T, RS, DW, 

TS, SRV 

NWC, ISI w. TSL, 

88 

RHR 304 SS 8 – 24 304 SS SS SCC, FDR P, T, RS, DW, 

TS, SRV 

NWC, ISI w. TSL, 

88 

RWCU 304 SS 8 – 12 304 SS SS SCC, FDR P, T, RS, DW, 

TS, SRV

NWC, ISI w. TSL, 

88 

CRD 

piping 

304 SS < 4 A600 

and SS 

Creviced 

NB 

welds 

SCC, FDR P, T, RS, DW, 

O 

NWC, ISI w. TSL, 

88 

SLC 304 SS < 4 304 SS SS SCC, FDR P, T, RS, DW, 

SRV 

NWC, ISI w. TSL, 

88 

INST 304 SS, < 4 304 SS SS MF, FDR P, T, RS, DW, 

V, SRV 

NWC, ISI w. TSL, 

88 

Drain 

lines 

304 SS < 4 304 SS SS SCC, FDR P, T, RS, DW, 

SRV 

NWC, ISI w. TSL, 

88 

Head 

spray 

304 SS, < 4 304 SS SS TF, FDR P, T, RS, DW, 

SRV 

NWC, ISI w. TSL, 

88 

SRV 

lines 

CS 6, 8, 10, 

28 

CS MF, FDR P, T, RS, DW, 

SRV 

NWC, ISI w. TSL, 

88 

RCIC 304 SS 6, 8 304 SS SS SCC, FDR P, T, RS, DW, 

SRV 

NWC, ISI w. TSL, 

88 

Specific piping, safe end, and weld materials were not determined by the group. Table B.1.11 represents 

an initial attempt to select these materials based on the general discussion. However, the group did decide 

to consider uncrevised Type 304 where proper in the reference cases. The mitigation and maintenance for 

all systems should assume NWC. Standard ISI with technical specification leakage detection should be 

considered along with augmented inspection as defined in generic letter 88-01. The mitigation and 

maintenance also has listed alloy 182 and stress improved, but these concepts need to be better defined 

and summarized. 

The PWR reference case conditions are provided in Table B.1.12. The reference cases were again 

distilled from the LOCA-sensitive PWR piping systems (Table B.1.8). The philosophy behind this table 

was consistent with the BWR reference case development (Table B.1.11) with a few notable exceptions. 
B-31

The effects of fabrication defects and repair on the other listed degradation mechanisms should again be 

considered in all piping systems. However, several PWR reference cases list multiple other degradation 

mechanisms which is a departure from the BWR approach (Table B.1.11). The PWR piping reference 

cases also account for nominal loading supplied by pressure, thermal, residual stress, and dead weight

loading to each system. However, very few PWR systems have associated loading transients while all 

BWR systems do. It may be necessary to add associated PWR transients for consistency. The PWR 

piping, safe end and weld materials were also not specified. Some initial choices have been made in 

Table B.1.12, but feedback from the group is required in order to finalize selection. The mitigation and 

maintenance to be considered for each reference case consists of ISI with TSL. No other special 

mitigation procedures were identified. 

Table B.1.12 PWR Reference Case Conditions 

System Piping 

Material 

Piping 

Sizes (in) 

Safe 

end 

Welds Degradation 

Mechanisms 

Loading Mitigation and 

Maintenance 

RCP: Hot Leg 304 SS 30 - 44 A600 NB TF, 

SCC,FDR 

P, T, RS, DW ISI w TSL 

RCP: Cold/ 

Crossover Legs 

304 SS 22 - 34 A600 NB TF, FDR P, T, RS, DW ISI w TSL 

Surge line 304 SS 10 - 14 A600 NB TF, FDR P, T, RS, 

DW, TFL, TS 

ISI w TSL 

SIS: ACCUM 304 SS 10 - 12 304 SS SS TF, FDR P, T, RS, DW ISI w TSL 

SIS: DVI 304 SS 2 - 6 304 SS SS TF, FDR P, T, RS, DW ISI w TSL 

Drain line 304 SS < 2” SS MF, TF, 

FDR 

P, T, RS, 

DW, V 

ISI w TSL 

CVCS 304 SS 2 – 8 SS TF, MF, 

FDR 

P, T, RS, 

DW, V 

ISI w TSL 

RHR 304 SS 6 – 12 TF, FDR P, T, RS, 

DW, TS 

ISI w TSL 

SRV lines 304 SS 1 – 6 TF, FDR P, T, RS, 

DW, SRV 

ISI w TSL 

PSL 304 SS 3 – 6 NB TF, FDR P, T, RS, 

DW, WH 

ISI w TSL 

RH 304 SS < 2 A600 TF, FDR P, T, RS, 

DW, TS 

ISI w TSL 

INST 304 SS < 2 MF, TF, 

FDR 

P, T, RS, 

DW, V 

ISI w TSL 

It is important that the baseline and reference case conditions be clearly defined prior to the start of the 

elicitation so that each panel member understands the general attributes of each of these cases. As 

mentioned previously, the elicitation questions will be structured to query the variability and uncertainty 

associated with each piping system with respect to the reference cases. Each panel member will compare 

the reference and base cases to assess the relative importance of each piping system to the total LOCA

frequencies. Every effort will be made to accommodate all requests and information will be shared 

among the group. Additionally, any areas or issues which are not clear to a panel member should be 

raised to Rob Tregoning as soon as it arises. 

Non-Piping LOCA Contributions 

The final portion of the meeting concentrated on developing issues associated with non-piping 

contributions (passive failures only) to the LOCA frequencies. Active components will be analyzed 

separately during this program from operating experience. Because active components have maintenance 
B-32

plans, the group in general expects that the failure rate of these components will not increase in the future. 

Operating experience should therefore adequately represent active component failure rates. This rationale 

is the basis for considering only passive component failures within this elicitation. 

The non-piping contributions will be combined with the piping component to determine the total LOCA 

frequency (Figure B.1.1). The group decided to break these issues down by component functionality. 

This is an analogous approach used to tackle the piping contribution which initially segregated piping

systems by functionality. Five main components were determined as candidates for passive failures: the 

pressurizer, the RPV, valves, pumps, and steam generators/steam systems. The valve component 

category encompasses both pressure isolation valves at Class 1 to Class 2 piping boundaries and also 

loop-stop valves. The pumps category only considers pumps in the reactor coolant or recirculating water 

system. 

For each component category, the panel developed sub-categories which represent specific possible 

failure modes (e.g. what portions of the component could fail passively). Each failure mode is governed 

by the same variables that are important for piping systems (material, geometry, loading, degradation 

mechanisms, and mitigation/maintenance). Unfortunately the group did not have sufficient time or 

resources to fully develop comprehensive variable lists in the same manner as for piping systems. 

Table B.1.13 illustrates the failure modes developed for pressurizer failures. Please note that all table 

abbreviations for this and all subsequent tables in this section are as previously defined unless indicated. 

Bold items in the failure mode sub-category of Table B.1.13 (and all following tables) indicates that 

operational data exists which captures that component failure. For instance, in Table B.1.13, the group 

thinks that data is available on heater sleeve failures. 

Table B.1.13 Pressurizer Failure Scenarios 

Component Geometry Material Degradation 

Mechanisms 

Loading Mitigation/ 

Maintenance 

Comment 

Shell A600C-LAS, 

SSC-LAS 

GC, SCC, 

MF, FDR, UA 

 Boric acid wastage 

from OD 

Manway NB-LAS, 

SSC-LAS, 

LAS, 

HS-LAS 

(Bolts) 

GC, SCC, 

MF, SR, FDR, 

UA 

 Bolt failures 

Heater 

Sleeves 

Small 

diam. (3/4 

to 1 in) 

A600, SS TF, MF, 

SCC, FDR, 

UA 

 Req. multiple 

failures 

Bolted relief 

valves 

 C-SS MA, FDR, 

UA 

 

Nozzles SSC-LAS 

C-SS 

CD, TF, SCC, 

MA, FDR, 

UA, GC 

 Same as surge line 

NB-LAS = nickel-based clad low alloy steel 

SR = Stress Relaxation and loss of preload 

The panel identified failures in the pressurizer shell, manway, heater sleeves and nozzles as passive 

LOCA candidates. Also, the pressurizer bolted relief valves could fail. The group generally did not have 

information on component geometries and loading and mitigation/maintenance were not discussed by the 

panel due to lack of time. However, some specific issues were discussed for each of these failure modes. 

The shell failure envisioned would most likely occur by boric acid wastage from the outer diameter of the 
B-33

shell. Manway failures would result by multiple bolt failures. Heater sleeves fail due to PWSCC, but as 

a result of their size, multiple failures are required in order to result in a LOCA. Bolted relief valves 

could fail due to steam cutting or localized bolt corrosion resulting from boric acid leaks. 

The RPV failure modes (Table B.1.14) focused on vessel head bolt failure, failure of CRDM connections,

nozzle failure, RPV wastage, and RPV corrosion fatigue. Upper head vessel head bolt failure is most 

likely due to human error during removal at each refueling cycle. Human error could occur as a result of 

improper installation procedures. Problems, however, could be identified during prestart-up inspection. 

The lower head bolts are not removed during refueling and they could be susceptible to common cause 

failure resulting from local bolt corrosion leading to several simultaneous bolt failures. A certain 

percentage of these bolts are inspected at each outage and the assumption is that inspection would not be 

effective in identifying the degradation prior to failure. These requirements may uncover the likelihood 

of common cause errors leading to some latent failure that is not immediately evident and shed light on 

other possible failure mechanisms. An example of a common cause failure is a torque wrench/tensioner 

which is out of calibration so that all bolts are improperly installed and then can possibly fail during 

operation. 

CRDM connections far outside of the reactor could be welded, bolted, or threaded and seam welded. The

degradation mechanism would be a function of the specific connection. For instance, welded connections 

would be susceptible to the mechanisms and loading discussed previously for CRDM components. 

Bolted CRDM connections would be subject to steam cutting, boric acid corrosion, aging and other 

degradation mechanisms that are unique to bolts. It must be stressed that the CRDM connections in this 

table refers to the CRDM which connects to the drive mechanism. Inboard connections are considered to

be part of the “CRDM piping system” discussed earlier. For bolted connections, this demarcation line is 

the flange joint. The group identified failure data for CRDM leakage from bolted flanged connections.
B-34

Table B.1.14 Reactor Pressure Vessel (RPV) Failure Scenarios 

Component Geometry Material Degradation 

Mechanisms 

Loading Mitigation/ 

Maintenance 

Comment 

Vessel Head 

Bolts 

 high strength 

steel 

GC, FDR, UA Human error Removal leading to 

human error 

(common cause 

failure) during 

refueling 

RPV 

wastage 

 SSC-LAS 

LAS 

GC, FDR, 

UA, MA 

 LAS = some BWR 

upper head, 

Boric acid wastage 

(upper & lower 

head, shell) 

CRDM 

connections 

 SS FDR, UA welded, bolted, 

threaded + seal 

weld 

CRDM 4-6 A600 base 

nozzle, SS, CSS, and NBLAS housing 

with NB weld 

SCC, TF, MF, 

LC, GC, FDR, 

UA 

P, S, T, 

RS, DW, 

O 

HREPL, ISI w 

TSL, REM 

Nozzles and piping 

up to connection 

Nozzles LAS, 

SSC-LAS, 

TF, MF, LC, 

GC, SCC, 

FDR, UA 

 LAS = BWR only 

ICI < 2” 304 SS, 316 

SS 

MF, SCC, TF, 

FW, FDR, UA 

P, S, T, 

RS, DW, 

O, V 

ISI w TSL, 

REM 

RPV 

Corrosion 

Fatigue 

 SSC-LAS 

LAS 

LC, MF, MA 

FDR, UA 

 LAS = some BWR 

upper head, 

Initiate at cladding 

cracks (upper & 

lower head, shell) 

BWR 

penetrations 

 SS SCC, LC, 

FDR, UA 

 Stub tubes, drain 

line, SLC, 

instrumentation, 

etc. 

PWR 

penetration 

 SS, 

A600 

SCC, FDR, 

UA, LC, MF, 

TF 

 

NB-LAS = nickel-based clad low alloy steel 

SR = Stress Relaxation and loss of preload 

There are two RPV degradation mechanisms which were specifically discussed. The first was 

degradation of the shell, upper head, or lower head due to boric acid corrosion. The second mechanism

was corrosion fatigue developed at through-thickness cladding cracks in the shell, upper head, or lower 

head. The nozzle category is subject to similar degradation mechanisms as in the attached piping. It

should be stressed that the nozzle category only considers the flared portion of the nozzle up to the reactor 

shelf. The nozzle safe end was earlier defined as part of the piping system. The group identified that 

some data on nozzle issues exists. 

Valve failure modes are summarized in Table B.1.15. The cast stainless steel valve bodies are susceptible 

to an array of potential degradation mechanisms. These include cavitation (CAV), thermal fatigue (TF), 

and material aging (MA). Casting defects (CD) are another particular concern. Failure due to the other 
B-35

mechanisms listed could initiate at either the defects, or repairs of those defects. The main steam 

isolation valve (MSIV) body is associated with similar failure modes. Specific failure modes for valve 

bonnets, and valve bonnet bolts were not discussed. Presumably, bonnet bolt failures would be 

susceptible to the same failure mechanism of other bolts: aging, boric acid corrosion, steam cutting, etc. 

The hot leg/cold leg loop isolation valve failure modes were also not discussed. However, failures in

these valves could be described in terms of the bonnet, body, or bonnet bolt failure sub-categories listed 

earlier. It should also be noted that valve sizes are generally consistent with the piping system where they 

are located. 

Table B.1.15 Valve Failure Scenarios 

Component Geometry Material Degradation 

Mechanisms 

Loading Mitigation/ 

Maintenance 

Comment 

Valve Body CS, SS 

C-SS 

FAC, CAV, 

LC, TF, MA, 

GC, CD, 

SCC, FDR, 

UA 

 CS, SS = BWR 

only 

Valve Bonnet CS, SS 

C-SS 

FAC, LC, GC, 

SCC, MA, 

CD, FDR, UA 

 CS, SS = BWR 

only 

Bonnet Bolts HS-LAS GC, SCC, 

FDR, UA SR 

 

Hot Leg/Cold 

leg loop 

isolation 

valves 

 FDR, UA 

MSIV Body CAV, TF, 

MA, CD 

 

HS-LAS = High Strength Low Alloy steel (SA540 GrB23, SA193 GrB7) 

CAV = Cavitation Damage 

SR = Stress Relaxation and loss of preload 

Steam generator tube rupture (Table B.1.16) can occur from a variety of different mechanisms including

thermal fatigue, mechanical fatigue, SCC, and general corrosion. The tubes can also be degraded by 

mechanical deformation (MECDEF), or denting, during installation, inspection, or cleaning. Steam 

generator tubes are too small to lead to a LOCA due to a single tube failure. Therefore, multiple tube 

rupture needs to also be considered in order to achieve a certain size LOCA. There is data which exists 

for SGTR. 

Steam generator failure can also occur at the manway (specifically bolt failure), the steam generator shell, 

or the nozzles. These various failure modes were also not sufficiently discussed so little information has 

been defined in Table B.1.16. However, the nozzle failure issues will likely be similar to the associated 

piping system, while manway bolt failure would be caused by the same types of mechanisms as for other 

bolt failures. 

The pump failure modes (Table B.1.17) are similar to many of the failure modes already discussed for 

other components. The cast pump bodies are potentially subject to the same degradation mechanisms 

(CAV, TF, CD, MA) as other cast components. The recirculation (RECIRC) bonnet bolts and RCP 

nozzle are also susceptible to mechanisms discussed earlier. The only unique mode considers an incipient 

failure of a pump flywheel which could initiate collateral damage in other components or in other piping 

systems. There was no appropriate passive pump failure data that was identified by the group. 
B-36

Table B.1.16 Steam Generator/Steam System Failure Scenarios 

Component Geometry Material Degradation 

Mechanisms 

Loading Mitigation/ 

Maintenance 

Comment 

Tube 

Rupture 

5/8 to 3/4“ 

diam. 

A600 TF, MF, 

SCC, GC, 

LC, FRET, 

MECHDEF, 

FDR, UA 

 single and 

multiple tube 

rupture 

Manway 

Bolts 

 CS, 

LAS 

SCC, GC, LC, 

SR, 

FDR, UA 

 

Shell CS, LAS, GC, LC, MF, 

TF, FDR, UA 

 

Nozzles to 

safe end 

 SSC-LAS 

CS, LAS 

SSC-CS 

FAC, SCC, 

FDR, UA 

 

Tube Sheet 

Failure 

 NB-LAS 

A600 

SCC, GC, 

FRET, MF, 

FDR, UA 

 

FRET = fretting or mechanical wear 

Table B.1.17 Pump Failure Scenarios 

Component Geometry Material Degradation 

Mechanisms 

Loading Mitigation/ 

Maintenance 

Comment 

Pump Body C-SS, 

SSC-CS 

CAV., TF, 

CD, MA, 

SCC, fatigue 

 

RECIRC 

Bonnet Bolts 

 HS-LAS SCC, GC, SR 

RCP nozzle 

Flywheel 

failure 

 initiating collateral 

damage – 

secondary pipe 

failure 

HS-LAS = High Strength Low Alloy steel (SA540 GrB23, SA193 GrB7) 

SR = Stress Relaxation and loss of preload 

It is obvious that the non-piping passive LOCA sources have not been nearly as well-defined as the piping 

system sources, and they must be better defined prior the elicitations. However, due to the number and 

complexity of the components, the panel realized that it may not be possible to fully define all the 

variables listed in the tables above. At a minimum, the group decided that it would need isometric 

drawings for as many of these components as possible. 

The manner for arriving at the LOCA contributions of these other components will be similar to the 

approach followed for the piping contribution. Reference cases will be developed and absolute LOCA 

estimates will be assigned to those numbers. However, these reference cases will be based strictly on

data. The bolded items in Tables B.1.13 – B.1.16 are component failures that are supported by passivesystem failure data. This data will first need to be accumulated and analyzed. Karen Gott and Bill 
B-37

Galyean are possible sources for some of this data. There is an EPRI database called PM-BASIS which 

consists of mainly active components, but there may be some data on bonnets and packings. Also, 

Spence Bush may have some data for these components that might be useful to the group. Finally, the 

group discussed that there may be data available for feed water nozzles. 

Once the data is developed, it will be made available to the group. This data will make up the base case 

information for the non-piping components. The group will also be asked how representative the base 

case data is for future (end-of-plant-license-renewal) LOCA estimates. The LOCA propensity (for each 

leak threshold rate) for the components without data will also be queried relative to these base cases. This 

approach is identical to the development of the piping LOCA contributions discussed earlier. 
B-38

MEETING MINUTES FOR SECOND ELICITATION MEETING 

FOUR POINTS SHERATON, BETHESDA MD

Day 1 – June 4, 2003 - Base Case Review

Dr. Rob Tregoning (USNRC) welcomed everyone to the Second Elicitation meeting and reviewed the 

agenda for the two days. Everyone in attendance introduced themselves. A package of the Day 1 

presentations was provided to everyone. Rob warned the group that there was a lot of material to cover in 

each presentation. Furthermore, he indicated that we should not treat what is presented at this meeting as 

final, but more of a snapshot of where we are presently. Next, Rob reviewed the objectives for the first 

day of the meeting. The objectives for the first day were: 

1. Review base case conditions 

2. Understand assumptions, methodologies, and results calculated by each base case member 

3. Understand important factors and variables that lead to differences among results 

4. Determine what additional calculations are required to complete the base case analysis 

Rob also indicated that he was not asking everyone to agree with the results to be presented, but simply to 

understand what was done. The panel members can state differences of opinion during their individual 

elicitations. 

Next, the second day agenda (Elicitation Coordination) was discussed. The second day agenda has been 

adjusted slightly to ensure adequate time for the topic of non-piping LOCA frequency determination since 

it received less attention at the last meeting 

Rob then reviewed the meeting objectives for Day 2. The Day 2 objectives were: 

1. Finalize elicitation question sets and provide consistent understanding of each question. 

2. Determine methodology for evaluating non-piping LOCAs and identification of non-piping base 

case data. 

3. Determine methodology for evaluating conditional seismic loading including determination of 

seismic loading magnitude. 

4. Determine what information panelists will require prior to their elicitations and assign action 

items for providing information. 

5. Develop final schedule and time-frame for upcoming elicitations. 

Bruce Bishop (Westinghouse) asked if the panel would get a status report on the new PFM code being 

developed as part of the USNRC program. Rob indicated that time was not available at this meeting, but 

that sometime in the fall or winter an initial meeting will be set up where this new code being developed 

by Battelle and Emc2

 could be presented.

Presentation 1 – Base Case Review and Summary Results 

by Rob Tregoning of the USNRC

The purpose of this presentation was to review the conditions analyzed by the base case team and 

summarize the calculated results to date. This talk served as a prelude for the next four presentations by 

the base case team members. 

The base case results will be used to anchor elicitation responses by the elicitation panel members as part 

of their individual elicitations. The elicitation members can use one or all of the base case results directly 

for their anchoring, or provide their own base case analysis if they choose. There were a total of five (5) 

base cases defined at the first elicitation meeting in February, see Table B.2.1. 
B-39

Table B.2.1 Base Case Conditions

Base Case 

Identification 

Piping System Pipe 

Diameter, 

inches 

Piping Materials Degradation 

Mechanisms 

Loading Mitigation 

BWR-1 Recirculation 12 to 28 Type 304 

stainless 

(originally), 

non-creviced 

A600 safe ends, 

nickel based 

(NB) welds 

IGSCC Pressure, RS, 

DW, safety 

relief valve 

transient 

(SRV) 

NWC, leak 

detection (LD), 

ISI, augmented 

inspection per 

Generic Letter 88-

01 (88-01) 

BWR-2 Feedwater 12 Carbon steel FAC, thermal 

fatigue (TF) 

P, RS, DW, 

thermal (T), 

water hammer 

(WH), 

NWC, LD, ISI, 

88-01 

PWR-1 Hot leg 30 Type 304 

stainless, A600 

safe ends, NB 

welds 

PWSCC, TF P, RS, DW, 

T, pressure 

pulse (PP) 

LD, ISI 

PWR-2 Surge line 10 Type 304 

stainless, A600 

safe ends, NB 

welds at 

pressurizer 

PWSCC, TF P, RS, DW, 

T, PP 

LD, ISI 

PWR-3 High pressure 

injection 

makeup 

nozzle 

(HPI/MU) 

(B&W) 

4 Stainless and 

carbon steel 

TF P, RS, DW, 

T, PP 

LD, ISI 

As part of the base case effort, the base case members were to evaluate the LOCA frequencies at 25 years 

(current-day), 40 years (end-of-plant-license), and 60 years (end-of-plant-license-renewal). These results 

will then be used by the individual elicitation panel members to anchor their respective responses so that 

they can estimate the various LOCA frequencies at these same time periods. 

The goal for the base case members is to calculate results for the set of conditions listed in Table B.2.1. 

The base case members also shared their results and presentations prior to this meeting so that there was a 

common format for the presentations. At this time the base case results comparison charts in the handouts 

should be viewed as works in progress. Furthermore, some results from Vic Chapman (OJV 

Consultancy) are still forthcoming. Vic and Chris Bell need to provide additional information to the 

panel members on how they conducted their base case analyses. 

The current base case results are summarized in slides 16 through 20 of this presentation. For David 

Harris’s calculations, the frequency results at 60 years were averaged over the 20 year time period from 

40 to 60 years while the 25 year estimates were averaged over the first 25 years of operation. It is 

important that results are consistent (with consistent assumptions and conditions) among each base case 

team member. 

The hot leg results (PWR-1 on page 18 of the handout) indicates large initial uncertainty. Bruce Bishop 

questioned if the results are for individual welds or the overall system. The response was that the intent 

was that these results should reflect the frequencies for the overall system. It was pointed out that the hot 
B-40

leg results should reflect the LOCA frequencies for the hot leg only. The results should not consider all 

the other lines associated with the RCS, i.e., the cold leg, cross over, surge line, etc. 

The slides 21 and 22 (Remaining Work and Differences Among Methodologies) of Rob’s presentation 

(Base Case Conditions and Summary Results) were not included in the handout and were to be filled out 

later by the team members. These updates will be posted on the ftp site once available. It was indicated 

that the LOCA frequencies for the lower leak rate categories (> 100 gpm [380 lpm]) include all the 

incidences of LOCAs in the higher leak rate categories, e.g., the 100,000 gpm (380,.000 lpm) bin should 

include all incidences of LOCAs in the 500,000 gpm (1,900,000 lpm) bin. 

Presentation 2 – Report No. 1 by Base Case Team to Expert Panel on LOCA Frequency 

Distributions 

By Bill Galyean, INEEL 

Bill employed a “top down” approach in his analysis of operating experience. His database represents 

approximately 2,600 LWR-years of operating experience. The resultant average age of a plant is 23 

years. In that 2,600 years of LWR operating experience there have been no passive system LOCAs with a

resultant leak rate greater than 100 gpm (380 lpm) (Category 1 LOCAs). 

As part of this analysis, Bill assumed that cracks and leak events are indicators of LOCA frequencies. 

They indicate system susceptibility. In order to get a LOCA, Bill’s analysis assumed that a piping system 

must first have a leak or a crack. In the first 2,647 years of US LWR experience represented in Bill’s 

database there have been approximately 1,100 crack and leak events, but no LOCAs. Note, at the first 

elicitation meeting the demarcation between leaks and breaks was set at a 100 gpm (380 lpm). 

A comment was made that small pipes are more susceptible to LOCAs than large pipes, i.e., large pipes 

are less likely to fail catastrophically. A question was raised as to why limit the analysis to US operating 

experience only. Bill did not categorically know whether there have been any 100 gpm (380 lpm) 

LOCAs worldwide. Another reason to limit his analysis to US operating experience is that there are some 

fundamental design differences between US and other overseas plants. Pete Riccardella (Structural 

Integrity Associates) thought that if Bill had included foreign experience that the number of years of

operating experience would have about doubled so Bill’s LWR LOCA frequency number of 1.9E-04/year 

be would reduced by a factor of two to approximately 1.0E-04/year. It is important to understand the 

basis for this 1.9E-04/year number since everything else is referenced to this number. This number is the 

total number of LOCAs of all sizes. 

It was noted that this is a different approach than followed in NUREG/CR-5750. This analysis was not an 

attempt to update NUREG/CR-5750. Pete Riccardella asked if this database included all of the small 

diameter socket weld cracks that occur due to vibration fatigue. Bill indicated that this was the case, even 

though these small diameter lines could not result in a 100 gpm (380 lpm) leak. It was pointed out that 

there was no distinction between cracks and leaks in Bill’s analysis. Any crack deeper than 10 percent of 

the wall thickness was included in the analysis. 

Gery Wilkowski (Emc2) asked if the analysis of the feedwater system (BWR-2) included FAC as a failure 

mechanism. Gery noted that on the secondary side there have been large breaks in some piping systems 

due to FAC. Bill indicated that he limited his database search to those systems that affected reactor

coolant pressure boundary integrity. 

Karen Gott was surprised at the low number of incidences for the feedwater system. Most of the 

problems seen to date with the feedwater systems have been outside the primary portion of system. 

Furthermore, cracks in nozzles are associated with the RPV and not piping. For the PWR systems there 
B-41

has been a lot of feedwater cracking, but those cracks have been on the secondary side. It was also 

pointed out to the panel that everyone has access to the SLAP database that Bill used for his analysis. It 

is now on the ftp site. Rob encouraged everyone to use it as part of their elicitation exercises. The SLAP 

database is current up to the end of 1998. 

A question was raised about the validity of the single IGSCC failure reported in the carbon steel 

feedwater system. Bill indicated that this was the reported database value. Karen Gott said they’ve seen 

such cracking in Sweden as well. 

The BWR recirculation system provided a unique problem for the base case analysis. From a materials 

standpoint, the base case was the old system. Bill segregated the data by old pipe (Type 304 stainless) 

versus new pipe (Type 316 nuclear grade [NG]). For the old pipe (Type 304 stainless) there were 127 

events in 550 years of operating experience versus 3 events in 410 years of operating experience for the 

new pipe (Type 316NG). 

The resultant leak/crack frequency for the old pipe (127 events/550 years = 0.231 events per year) is 

about a factor of 2 greater than the overall leak/crack frequency for the overall recirculation system

history (old plus new), i.e., 130 events/960 years = 0.135 events per year. It was pointed out though that 

this improvement may be more due to other factors than pipe replacement only. The improvement could 

also be due to changes in water chemistry, or the installation of weld overlay repairs. Hence, it may be 

more appropriate to refer to the pipe systems as mitigated (new) or unmitigated (old) pipe systems. 

It was stated that the base case is unrealistic in that it is for the old pipe case (Type 304 stainless) and no 

one uses that material anymore. Also, the base case does not account for the incorporation of water 

chemistry improvements which all plants have already implemented. 

As part of his analysis, Bill made an assumption that the LOCA categories/sizes (e.g., 100, 1500, 5000

gpm, etc. [380, 5,700, 19,000, lpm, etc.]) are related on a logarithmic sense (1, 0.3, 0.1, 0.03, 0.01, 0.003). 

Half likelihood on logarithmic sense realizing that larger LOCAs are a subset of Category 1 (100 gpm 

[380 lpm]) LOCAs. 

The 40 welds for the PWR-1 case (hot leg) include the cold leg and cross over leg welds. This is 

inconsistent with the assumption stated above that the PWR-1 case only considers the hot leg (not the 

cold leg or cross over leg). (Note, there are typically 3 loops in a PWR plant and there can be 5 to 7 

welds per loop, but the loading is not the same for all these welds.) 

Rob Tregoning indicated that the correlation between pipe size and LOCA size (gpm) that were originally 

supplied are subject to change. 

Bill’s aging correction factor is for thermal fatigue and should not be used for the other failure 

mechanisms. 

The non-pipe LOCAs that Bill included are for passive systems only (bolted flanges, etc.). He did not

include active system contributions, e.g., stuck open valves, to the non-pipe break frequencies. 

Bill’s results for the “Current Estimate” are significantly smaller than the LOCA frequencies reported by 

others (WASH-1400, NUREG-1150, NUREG/CR-5750) due to the larger database (more years of 

operating experience without a LOCA). The best agreement is with the NUREG/CR-5750 results. Bill 

employed a Bayesian approach as part of his LOCA frequency analysis by assuming a half of failure for 

these very low occurrence events. This is a very common data analysis practice. Since Bill’s analysis is 
B-42

based on the analysis of past passive-system failure data, the data implicitly include ISI and other 

mitigation experience implemented by industry. 

Each plant has a PRA which includes LOCA frequency estimates for the plant. These LOCA frequency 

estimates are often based on WASH-1400 or NUREG-1150 and the ranges shown on pages 36 and 37 of 

Bill’s presentation are the ranges for the IPEs. Thus, the IPE range, WASH-1400 and NUREG-1150 

frequencies shown by Bill on pages 36 and 37 of his handout are closely related/interlinked. As described 

in Slides 36 and 37, Bill expressed the uncertainty in his analysis by assuming an error factor of 10. This 

is somewhat crude and somewhat arbitrary, but the scope of the base case analysis did not ask for 

uncertainty, just a best estimate. 

Bill indicated that he has no strong technical basis for extending his “Current Estimate” analysis out to 40 

or 60 years. With Bill’s approach, the LOCA frequencies will go down as more years of operating 

experience are accumulated, unless a LOCA event occurs in the future. Bill presented information 

predicting 8 years to double the frequencies for thermal-fatigue events. This estimate conservatively

assumes no industry-wide mitigation programs. Rob Tregoning indicated that he did not include results

for Bill in the summary table for 40 and 60 years of operation because they are most appropriate for 

current estimates. 

Bruce Bishop was very uncomfortable predicting the future out to 40 or 60 years. He felt that the 

uncertainties are going to increase dramatically. Lee Abramson (USNRC) responded that this is a natural 

experience and likely shared by others on the panel. However, it’s incumbent that each member to 

attempt these predictions. Rob and Lee stressed again that the panel will not be forced to answer any

questions that they are very uncomfortable with. Rob further stressed that the March 2003 SRM 

specified that the NRC staff needs to revisit the LOCA frequency estimates every 10 years and the effort 

will be most concerned with the next 10 years. However, it is still important to gain longer-term insights. 

A question was raised as to whether or not to have Bill extend his analysis out to 40 and 60 years? Gery 

Wilkowski said, yes, he wanted to see Bill’s assumptions. Fred Simonen (PNNL) would like to see more 

partitioning by pipe size as part of Bill’s analysis. Dave Harris (Engineering Mechanics Technology) 

concurred. Sam Ranganath (formerly of General Electric) specifically indicated that it would be helpful 

for the BWR recirculation system because they replaced the smaller diameter pipes, but not the 28-inch

diameter pipes. Bengt Lydell is to check on the statistics in his database to see if any of the 28-inch 

diameter BWR recirculation lines had leaks and provide this information in his final base case report.

Pete Riccardella and Sam Ranganath were unaware of any General Electric large diameter recirculation 

pipes that leaked. 

Report No. 2 by Base Case Team to the Expert Panel on LOCA Frequency Distributions 

by Bengt Lydell (Erin Engineering and Research) 

Bengt, like Bill Galyean, used passive-system failure data in his analysis, but in a much different manner. 

Whereas Bill Galyean followed a “Top Down” approach, Bengt followed a “Bottoms Up” approach. 

Bengt’s presentation assumed that all BWR welds were category D & E welds. For BWRs, Bengt 

indicated that Category D & E welds specify inspection criteria based on Generic Letter 88-01. Category 

D welds have been subjected to weld overlay repairs and Category E welds are subject to IGSCC. 

Bengt’s used a different database than Bill Galyean did is his analysis. Bengt’s database is proprietary 

(PIPEex) and the panel will not have access to this during the elicitation. The SLAP database that Bill 

used is available to the panel members on the website. The cut-off date for PIPEex events is the end of 

2002 while the cut-off date for the SLAP database is the end of 1998. PIPEex has about twice the number 

of data entries as does SLAP and includes international experience. 
B-43

Bengt only looked at welds in his analysis. He did not consider base metals. His database did not show 

any occurrences of base metal indications. However, he invoked a wide definition for what was 

encompassed by the term “weld”. He included the heat-affected-zone (HAZ) and counter bore region into

his definition of what a weld was. Bengt also did not consider degradation of non-piping passive 

components in his analysis. 

For PWR systems, he assumed that the V.C. Summer and Ringhals cracks were circumferentially oriented 

cracks, not axially oriented. Whereas Bill did a “top down” analysis, Bengt did a “Bottoms up” analysis. 

The failure rates were derived for individual welds, and then an integration system level model was 

formed by combining the contributions from each individual weld failure to an overall pipe system failure 

frequency. 

The term “prior” has very specific meaning in this analysis. It means “before mitigation/remedial action

in response to a significant pipe failure”. Hence, failure rates that are input to LOCA frequency 

calculations explicitly account for reliability improvements (mitigation methods) made in response to past 

pipe degradation histories. Failure in Bengt’s analysis is defined as a “through-wall flaw resulting in 

leakage”. Bengt did not include surface cracks found during ISI in his data reduction process. 

Karen Gott has a report in Swedish that may be valuable in this effort. This report gives the number of 

leaks and the number of ISI detected surface cracks. Gery Wilkowski thought that was important 

information since leaking through-wall cracks are readily detected, but the surface cracks that would grow 

to be long in length are more of a LOCA threat. The number of records shown on slide 10 from Bengt’s 

presentation includes both leaks and cracks. 

 

There was much discussion among the group in an effort to understand slides 11 through 14. In slide 13, 

Bengt did not eliminate welds if mitigation was performed prior to 15 years of operation in developing his 

prior distribution. There were no leaks in BWRs after 15 years. From years 10 to 15, it is possible that 

there may have been some plants that used mitigation, but those mitigated plant weld numbers were still 

included in the weld failure rate analysis, i.e., that may account for why the weld failure rate was not 

accelerating as the number of years increase. Bengt used a Monte Carlo simulation to create this plot, 

where he needed to estimate the number of welds for the number of plants that were at a certain age. 

Results in slide 14 include results from US, Spanish, Swedish, and Japanese plants. The results were 

adjusted by the number of welds and type of welds. 

These preliminary results are used to determine the “prior” LOCA frequencies (weld failure rate). The

next step is to determine the “posterior” frequencies based on “prior” distributions. Bengt accounts for 

uncertainty in the knowledge base, which is fundamental difference between his analysis and Bill’s. 

Dave Harris thought the “posterior” frequencies should be equal to the Prior LOCA frequencies times the 

Likelihood Function. Lee Abramson explained that the likelihood function was built in. 

It is noted in slide 19 that the Bayesian update strategies are different for each base case. It was also 

noted again that the weld failures represent leaks only and not cracks. Weld failure was defined as a

through-wall flaw with leakage less than or equal to the tech spec limit for undefined leakage. 

Gery Wilkowski noted that in January 2003, PWSCC was found in a surge line bimetallic weld at the 

pressurizer in a Belgium plant. It was noted that some transient event is needed to cause a tech spec limit 

flaw to propagate to a higher category (well beyond 100 gpm [380 lpm]) LOCA. Slide 23 shows that the 

conditional probability of failure (PL/F) is a function of the nominal pipe diameter (DN), i.e., PL/F = a x 

DNb

, much like what is in NUREG/CR-5750 (Beliczey and Schulz, i.e., PR/TWC = 2.5/DN). While the 
B-44

exact formulations are different, the conditional failure probability in all cases is an inverse function of 

pipe size. 

The form of the conditional leak probability given a failure is inconsistent with the original development 

of the Beliczey and Schulz correlation which was developed to relate leaks to breaks as a function of pipe 

size. This use (see slide 24) may not be physically realistic because it assumes that larger diameter pipes 

are less likely to result in a category 0 leak. The implication is that larger diameter pipes are less likely to 

reach a Category 0 leak, and then progress to higher leak rates. Slide 26 presents information on the

aspect ratios of IGSCC cracks. It was asked how the deep, full circumference cracks (a/t = 0.5, 2/Β = 1.0) 

formed. The expectation is that these are likely crevice cracks. It was suggested that it would be nice to 

break down this data by pipe size in order to assess the relevance. 

Bengt assumes that a through-wall crack can only propagate into a large leak if there is a large transient 

event. Slide 27 documents the loading categories assumed to drive the crack among various LOCA 

categories. Category 0 to Category 1 LOCA progression can occur assuming moderate loading, while to 

go from a Category 0 or Category 1 LOCA to a Category 6 LOCA, would require an extreme loading 

transient. The general consensus of the panel members was that this was a very subjective approach. It 

was noted that there were about 400 water hammer events reported, but Bruce Bishop and Guy DeBoo 

said that if there were this many water hammers, then the plant piping system was probably redesigned. 

The extrapolation of results from the “Current Estimate” to 40 and 60 years is based on posterior analysis 

of prior results assuming no additional failures. If one assumes no additional failures, the failure rates 

will go down with time. Bengt will examine possibly extrapolating his base case results out to 60 years 

using another assumption. 

Report No. 3 by Base Case Team to the Expert Panel on LOCA Frequency Distributions 

by Dave Harris (Engineering Mechanics Technology) 

Dave prefaced his comments with the thought that he thinks that the base case results are surprisingly

close considering the differences in the approaches. Dave indicated that an important input to any PFM 

analysis is the stress history. This requirement is contrary to the operating-experience approaches where 

the stress history is indirectly reflected in the incidence of cracking events. Dave’s analysis is performed 

on individual pipe locations which are then integrated to determine the overall system frequency. 

Some key points from the crack initiation and crack growth portion of Dave’s presentation are that piping 

failures occur due to the initiation and growth of cracks. Cracks initiate due to stress corrosion or fatigue. 

Growth is controlled by fracture mechanics (other than early SCC). The question was asked as to what 

assumption Dave used as to the size of the crack once it initiates. Dave’s PRAISE code assumes that 

fatigue cracks are 0.3 inch (7.6 mm) deep (per a criteria proposed by Argonne National Laboratories 

[ANL]). This is based on an assumed 25 percent load drop definition of crack initiation from an S-N 

specimen test. In addition, in PRAISE the SCC rules differentiate between early SCC growth and fracture 

mechanics growth since the early growth is faster than calculated by fracture mechanics analysis. The

SCC rules in PRAISE assume a 0.001 inch (0.025 mm) deep surface crack with some distribution 

function on length. The latest version of PRAISE (2002) includes updates to the S-N curves that 

incorporate environmental effects. Pete Riccardella noted that Art Deardorff of SAI was doing an update 

of some of the environmental S-N results from EPRI. For crack growth, the focus of PRAISE is semielliptical part-through surface cracks. PRAISE considers crack growth in both the depth and length 

directions (K at both the maximum depth and at the ends of crack.) 

Dave pointed out that fatigue failure of welds is dominated by growth from pre-existing crack-like 

fabrication defects. Thus, the flaw distribution of initial fabrication defects is an important parameter to 
B-45

define. PRAISE stipulates that the final failure is controlled by tearing instability. PRAISE treats the 

stresses at maximum load as load-controlled stresses from stability analysis perspective. 

The leak rate in PRAISE is computed based on the length of the leaking through-wall crack on the inside 

pipe surface using the SQUIRT leak-rate code. The SQUIRT code has recently been updated with 

numerous technical enhancements as part of the USNRC Large Break LOCA program. PRAISE includes 

some of the mechanistic dependent crack morphology parameters from some of the earlier versions of 

SQUIRT, but not the new COD-dependent roughness, number of turns, and flow path length to thickness 

ratio parameters. In addition SQUIRT has been made more user friendly by incorporating a graphical 

user interface. Note, WinPRAISE is PC-PRAISE with a Windows pre-processor for entering input 

parameters. 

Dave used a stratified sampling technique that allows for the evaluation of extremely small probabilities 

events such as the case of fatigue crack growth from pre-existing defects (10E-17 frequencies). The 

approach also assumes that all cracks with leak rates greater than 5 gpm (19 lpm) are discovered and 

subsequently removed from service which implies that getting a higher Category LOCA (e.g., a 1,500 

gpm [5,700 lpm] LOCA) would most likely result from the growth of a long surface crack that pops 

through the wall thickness and immediately becomes a TWC with a length equal to length of the surface 

crack on inside pipe surface. (The only other means of achieving a higher Category LOCA would be 

through some sort of transient event.) 

Dave postulates that inaccuracies in leak-rate calculations will not significantly impact final LOCA 

frequencies. Even assuming the leakage detection capability equals zero should not have a large effect. 

Sam Ranganath felt that a surface crack grows 3 or 4 times faster in the length direction than it does in 

depth. He wasn’t sure if that was due to multiple initiation sites, or the surface growth rate being higher. 

The analysis in PRAISE considers crack growth in both the depth and length directions (K at both depth

and ends of the crack) with an RMS value of K in each direction. 

The detection probabilities shown in slide 14 are based on depth only, not length. It is likely that current 

technology has better performance. It was also noted the fatigue crack growth parameter used could be

improved based on newer results. Dave Harris indicated that sensitivity studies show that using improved 

crack growth parameters in PRAISE (e.g., including environmental effects) will result in changes in 

LOCA frequencies on the order of a factor of 2. Slide 16 presents an example S-N initiation curve for a 

low alloy steel. Vic Chapman raised the concern that there may not be a plateau or fatigue limit with the 

higher number of cycles. Gery Wilkowski commented that the default flow stress values (slides 17 and 

18) are very tight from a standard deviation perspective, especially in light of what was seen in 

NUREG/CR-6004 from an analysis of PIFRAC data. 

The NUREG-6674 stresses have been downgraded from the design basis stresses by Jack Ware of INEEL 

to make them more realistic with fewer transients. The stresses may have been elastically calculated 

values, which can go well above yield and still be allowed for secondary stresses by the Code. PRAISE

uses the most realistic values available. The final results (frequencies) from Dave’s analyses are very 

dependent on stress input values. Dave commented that one only needs stresses at the high stress 

locations. These high stress locations dominate the final frequency answers. 

Slide 21 shows the surge line stresses. These stresses are probably for the flank of the elbow, not the 

weld. These values can’t be used for the girth weld location (i.e., they are too high). The stresses shown 

are stress amplitudes (the stress ranges will be twice these values). In addition to stresses and number of 

occurrences, one also needs some input as to the spatial distribution for these stresses. One of the short 

comings of PRAISE is that PRAISE doesn’t have a model for FAC for the feedwater lines. The fatigue 
B-46

initiation models are only in latest versions of PC-PRAISE, i.e., from NUREG/CR-6674 published in 

June 2000. 

An ad hoc procedure was used with pc-PRAISE in order to obtain results for larger leak rates (stratified 

sampling for fatigue crack growth is not available for fatigue crack initiation). One of the handouts

provided shows this ad hoc procedure. 

A question arose with slide 26 concerns the fact that Dave’s analysis shows that the cumulative failure 

probabilities continue to increase after 20 years after a weld overlay repair is applied at 20 years. Past 

experience at Battelle as part of the Degraded Piping Program showed that weld overlays are very 

effective. They have much higher strength than the base metal. Another aspect of their application is that 

they apply a very high compressive stress at the crack plane. These high compressive stresses should 

restrict any further crack growth of the surface crack. In addition, it was noted that these high 

compressive stresses may preclude the environment from getting to the crack tip. Dave noted that he put 

in a linear approximation of the stresses through the thickness, the increased thickness of the overlay, and 

the crack growth equations for Type 316 NG into his analysis. WIN-PRAISE uses an adjusted weld 

residual stress pattern (linear gradient) for the case of post-weld overlay residual stresses (see slide 38). 

Dave will also present the failure probabilities without the weld overlay so that an assessment of its effect 

can made. 

Bruce Bishop asked how much was Dave’s results affected by inspection. Dave didn’t think that the final 

frequencies would be affected that much. Dave’s results showed that there was a minimal change in 

LOCA frequencies for the hot leg as a result of the application of a 5SSE earthquake. This was not 

surprising to Dave since he has found similar behavior in a previous study. Lee Abramson pointed out 

that one means of seeing the effect of the earthquake is to compare conditions for equal probabilities. For 

example, for the 40 year time period analysis, no earthquake results in 1.3E-18 LOCA frequency for the

no leak case while for the same 40 year time period analysis, a 5SSE earthquake results in 1.3E-18 

frequency for a DEGB. 

Dave, generally found that the LOCA frequencies were not highly dependent on JIc and dJ/da. Gery 

Wilkowski thought that if the toughness was low enough that one was operating in the EPFM regime then 

the LOCA frequencies may be more dependent on toughness. Gery indicated that he thought that the 

toughness values used in Dave’s base cases were too high for weld crack locations, or aged cast stainless 

steel pipe and fittings. 

PRAISE can’t account for time dependent material properties. Thus, to account for aging, one would 

need to input aged properties in at time equal to zero. 

The very low LOCA frequencies for the hot leg in slide 31 may be an artifact of the failure mechanism 

(fatigue) chosen for analysis. Higher frequencies may be seen for some other mechanism, such as 

PWSCC. This is a case that we may want to analyze in future analyses. 

Report No. 4 by Base Case Team to the Expert Panel on LOCA Frequency Distributions 

by Vic Chapman and Chris Bell 

The first part of the presentation was presented by Chris Bell (Rolls Royce) with the final few slides

presented by Vic Chapman. Chris presented the general assumptions and methodology of the 

PRODIGAL code. Note that PRODIGAL actually has several modules. One of them is to determine 

weld defect size from welding information. Another is to determine failure probabilities for navy nuclear 

power plants. All results presented are for a per weld basis. 
B-47

In PRODIGAL, surface imperfections with a depth of 0.004 inch (0.1 mm) are assumed, but there are no 

SCC initiation/growth models in PRODIGAL. Past study has shown that there is little sensitivity to the 

assumed depth and sizes much less than 0.004 inch (0.1 mm) have little effect. For surge line case, there 

is sensitivity to the existence of 0.04 inch (1 mm) defects, see slide 19. 

Based on some work of Ritchie, Pete Riccardella felt that the 0.004 inch (0.1 mm) initial defect size was 

near the LB of the region where fatigue crack growth (da/dN) models were valid. It was noted that 

Omesh Chopra from Argonne thinks that this lower limit is closer to 0.02 inch (0.5 mm). It was 

suggested that this limit is dependent on the grain size of the material. Note, cast stainless steels can have 

very large grain sizes. 

Vic and Chris felt that they could not do a generic analysis for seismic considerations since the effect of 

seismic has been found to be highly dependent on plant layout. Therefore they didn’t consider seismic

stresses in their analysis. In addition, they only considered the three PWR base cases since they had little 

experience with BWRs. 

They used the same stress data as Dave Harris did for consistency purposes. As Dave did, they used a 

second order distribution of stresses thru the thickness per NUREG/CR-5505 criteria that was done by 

PNNL in 1998. 

The failure criterion for their instability analysis is based on the FAD approach in R6 using KIC, i.e., crack 

initiation would equal failure. The crack initiation used for stainless steel was closer to wrought base 

metal rather than weld metal which would be an order of magnitude lower, or aged cast stainless steel 

which could be another factor of 3 lower than the weld metal toughness. 

Bruce Bishop asked what the mean temperature in the analysis is used for. It is used for material property 

considerations, such as flow stress, but not for subcritical crack growth. Slide 16 from Vic and Chris’ 

presentation shows the cumulative probability of a TWC (probability of a leak occurring). The 

implication from this slide is that if you are going to have a leak, it will occur in the first 25 years of 

operations. It was noted that slide 16 is conditional on having a crack (probability of having a crack is 1). 

Rob Tregoning indicated that he wants each of the panel members in the next week to make list of what 

they want to see (e.g. data they used in doing their analysis) from either individual participants or from 

the group as a whole. 

It was noted that the dominant hot leg cycles are those due to heat up and cool down. There are only on 

the order of 5 of these cycles per year. 

Pete Riccardella thought that there were a lot of cycles on the PWR HPI/Make up nozzle each year. Dave 

Harris countered that he thought that there were only about 40 of these cycles in 40 years of operations. 

Bengt Lydell agreed with Pete and thought there were a lot of thermal cycles. Pete thought that 

something was missing here. Bengt said there was a nice ASME paper on the cyclic stress history of 

these nozzles that we could use in the analysis. 

Chris indicated that although they typically keep the aspect ratio constant in their PRODIGAL runs, they 

have the ability to grow cracks in length, and often get very irregular crack shapes. 

In slide 24, the “separation” referred to is the crack-opening displacement. In this slide, at the surge line 

elbow, Vic speculates that the crack starts to act as a hinge so that crack opening becomes very large for 

the longer crack lengths. This assertion is the basis for slide 25 illustration of the crack frequency versus 
B-48

angle distribution that was assumed in the analysis. This relationship was created by Vic with guidance 

from metallurgist to rectify predicted and experimental crack lengths. 

It was commented that for the same input, we are seeing similar results from PRAISE (Dave Harris’ 

results) and PRODIGAL (Vic Chapman and Chris Bell’s results). It was noted that as far as this exercise 

is concerned, PRODIGAL’s strength is the defect distribution analysis. PRODIGAL only looks at 

thermal fatigue while PRAISE can look at other failure mechanisms. PRAISE can also look at crack 

initiation while PRODIGAL cannot. However, neither accounts for FAC or PWSCC at this time. 

PRAISE can also look at crack initiation whereas PRODIGAL assumes crack growth from weld defects 

or surface imperfections. 

Vic pointed out that embedded cracks can straddle the compressive zone of the residual stress field 

through the thickness so that once they break thru to inside surface they are already through the 

compressive zone. This is in contrast to case where a crack is growing through the wall thickness from 

the inside pipe surface and the crack gets trapped in the compressive zone near mid thickness. There was 

also the question of whether embedded cracks are affected by the environment. 

The next topic for discussion was to plan the next step for the base case calculations. 

Bruce Bishop would like to bench mark the PFM results against the 25 operating-experience estimates 

developed by Bill Galyean and Bengt Lydell, and then use the PFM models (PRAISE and PRODIGAL) 

to predict the 40 and 60 year results. Rob Tregoning thought that this was an excellent idea. Rob also 

indicated that we could do this for some cases, but not all cases, e.g., FAC in the feedwater system or 

PWSCC. 

Sam Ranganath would like to know when we compare results where do we get good agreement and 

where not. Rob indicated that we haven’t made comparisons on a consistent basis as of this date but that 

this would be rectified. 

The next issue focused on additional stresses to consider in the PFM results. The surge line and HPI/MU 

cases were mentioned. Dave Harris has already done some additional analysis for the surge line, based on 

refined stresses developed by Art Deardorff. Gery Wilkowski noted that he had surge line stresses from a 

Westinghouse Owner’s Group report used for LBB analyses. Pete agreed to verify that the previous 

stresses provided by Art are appropriate. Pete also agreed to provide more accurate HPI/MU stresses. 

Sam Ranganath would like to lower the stresses to 10 ksi (70 MPa) for the recirculation line (BWR-1) 

and to lower the feedwater line stresses (BWR-2) by 20 percent. 

The next area where the panel thought we may want to focus is some sensitivity analysis using different 

material properties. Gery Wilkowski volunteered to supply some distributions of material properties 

(mainly toughness values for welds and aged cast stainless steels) developed as part of NUREG/CR-6004. 

Gery Wilkowski also wanted to see the ratio of surface crack to through-wall cracks removed from 

service, and the distribution of the lengths and depths of those service removed surface cracks. He then 

wanted to see a comparison of the PFM probability of leaks for IGSCCs from PRAISE and the 

distribution of surface cracks that might exist up to the time that piping might have been replaced/repaired 

(15 service years?). Karen suggested looking at the more recent results on a yearly basis because ISI

wasn’t sensitive enough to find surface defects in early years. Cracks were only discovered once they

became a leaking crack. Also, in Sweden, even if whole pipe sections were removed, all the welds were

inspected, whereas in the US if the pipe system was replaced they did not spend the effort to inspect 

welds that were being removed from service. Hence, the database of cracks removed from service should
B-49

separate Swedish and US plants. The important aspect of this comparison is to see the population of the 

ISI remove surface crack lengths compared to the surface flaw sizes calculated by PFM. If the service

removed crack lengths from ISI are much longer than calculated by the PFM analyses, then the PFM 

analysis should underestimate the future failure probabilities. Gery noted several times that failure

probabilities should be controlled by development of long surface cracks, not the growth of leaking 

through-wall cracks. Bengt Lydell has some papers relating ISI-detected surface-crack geometries to 

through-wall crack leaks that he can provide on the ftp site and can provide to Rob Tregoning. 

It was suggested to include PWSCC in the hot leg base case analysis. Dave Harris has initially done this 

using IGSCC relationships for preexisting flaws. Initiation is not accounted for. Gery Wilkowski noted 

that the crack growth rate through the weld metal (along the dendritic grains) is much faster than IGSCC. 

Gery will work with Karen Gott and Bill Cullen to see how Dave can adjust the PRAISE model to get 

initiation and growth for PWSCC. Gery will provide the IGSCC initiation and growth equations that 

PRAISE uses to Bill Cullen so that appropriate constants can be provided for PWSCC in PRAISE for 

base case calculations. Gery and Karen will provide information to Dave Harris on PWSCC crack 

initiation and growth models that he can use to evaluate the impact of PWSCC on the appropriate base 

case calculations. 

The panel thought it was important to address Dave Harris’ strange results for weld overlay repairs. What 

is leading to high growth rates after the overlay is applied. Can a comparison be shown with what would 

happen if no weld overlay had been applied? Dave Harris is to address this concern of the unexpectedly 

high growth rates after the weld overlay repair is applied. 

When the summary comparison tables are completed, the PFM subgroup needs to clearly identify where 

the PFM conditions do not agree with operating experience. Rob Tregoning indicated that the base case

subgroup needs to finish up the base case calculations by the end of the month. We cannot delay 

individual elicitations any longer. 

Day 2 (June 5) – Elicitation Coordination

Rob Tregoning started the morning by reviewing the agenda for the second day. There were six (6) basic 

items to cover. These were: 

• Reviewing the elicitation questions 

• Reviewing the leak rate versus pipe break size evaluation 

• Addressing the non-piping LOCA evaluations 

• Reviewing the conditional seismic evaluation 

• Addressing the additional information required prior to the individual elicitations for piping and 

non-piping evaluations 

• Elicitation scheduling 

The specific objectives for the second day included: 

• Finalize the elicitation question sets and to provide a consistent understanding of each question. 

• Determine the methodology for evaluating non-piping LOCAs and the identification of nonpiping base case data. 

• Determine the methodology for evaluating conditional seismic loading including determination of 

the seismic loading magnitude. 

• Determine what information the panelists will require prior to their elicitations and assign action 

items for providing information. 

• Develop the final schedule and time-frame for the upcoming elicitations. 
B-50

Presentation 1 (Day 2) – Elicitation Questions: Structure and Review 

by Rob Tregoning (USNRC) 

Rob stressed that prior to their individual elicitations, each panel member needs to do their homework and 

answer as many of the elicitation questions as possible. Panel members can change their answers at any 

time during this exercise, including during the actual elicitation and afterwards. 

It was indicated that the term “LOCA frequencies” should really be “LOCA probabilities” in this 

presentation during the analysis of the conditional emergency loading. Rob agreed to make this change to 

the presentation. 

Pete Riccardella questioned whether we should expand the conditional seismic loads to conditional 

emergency and faulted loads which include transients like water hammer as well as seismic. Fred 

Simonen thought that we needed to talk with the PRA people. Alan Kuritsky (USNRC) indicated that 

these other potential LOCA causing events were currently not included in the PRAs. Rob suggested 

tabling this discussion until later in the agenda. Bruce Bishop thought that if we continued with the

traditional seismic approach then we need to consider the fact that the snubbers may not work probably. 

Bruce indicated that there is a significant probability that the snubbers may not work as advertised. 

Rob next addressed the top down elicitation structure. Bill Galyean used a top down approach where he

assigned an overall piping LOCA contribution, and then looked at the breakdown in the contribution due

to piping system, geometry, load history, mitigation, materials, and degradation mechanisms. Vic 

Chapman questioned what the top down approach gave us (we start with the final answer that we are 

looking for). Lee Abramson indicated that at the end of the day we will get numbers in each of the blocks 

on slide 3 so that we can decompose the problem into the small pieces. The panel members can initially 

choose either a top down or bottoms up approach. 

Bruce Bishop asked how soon the panel members are going to have the final base case results prior to the 

first elicitation. Rob indicated that he would be working with the base case members the week of June 9 

to finalize their answers. He hoped to have the final results by the end of June. However, the base case 

conditions are well-known 

The panel members need to supply their answers (on a pre-established form) and the facilitation team will 

work with the individual panel members individually. The important point of the pre-elicitation exercise 

is to quantify the median value results, provide some qualitative uncertainty and also rationale. During 

the elicitations the panel members can change answers as they interact with the facilitation team and 

points are clarified. 

Bruce Bishop felt that the “utility safety cultural” should be “utility operations cultural” in that safety and 

economic drivers both feed into the operations cultural. Karen Gott indicated that the IAEA definition of 

safety cultural (and how we defined safety cultural at the kick off meeting) includes both safety and 

economic aspects. With regards to the ratios on safety cultural issues, ratios greater than 1.0 indicate that 

things are getting worse; less than 1.0 means things are getting better. 

Slide 6 shows a flow chart in which the question is which variables are independent. Variables in this 

context are geometry, load history, mitigation, materials, and degradation mechanisms. For these 

variables, each panelist may need to estimate the future impact of that variable, e.g., what new materials, 

or what new degradation mechanisms, should be expected in the future. We may want to look at the past

to estimate what might happen in the future (e.g., Gery Wilkowski’s plot of new failure mechanisms with 

time which shows a new mechanism approximately every 7 years). 
B-51

There ensued a long discussion on the comparisons of reference cases (defined at the kick off meeting) 

with baseline cases. We defined a reference case for each piping system (e.g., diameter, material, 

degradation mechanism, etc) whereas we only defined a few baseline cases to which the reference bases 

are to be compared (i.e., anchored). For example, the hot leg (base case) may be a natural comparison for 

the reference case for the cold leg. 

For Questions 3A.1 and 3A.2 (slide 7) and all related questions, Rob will change the “surge line” to “cold 

leg” example so questions don’t refer to a system that is both a base case and a reference case, realizing 

that we will end up asking same question for a surge line as well. For those systems which have both a 

base case and a reference case, these comparisons may be more natural and easier than inter-system 

comparisons. 

It was decided to eliminate the assessment of which variables are independent or dependent. All variables 

will be considered to be dependent as originally defined during the kick-off meeting. Thus, the original 

Question 3A.2 will be eliminated. Other References to correlated or independent variables in other 

questions should be eliminated in the final presentation version. For original Question 3A.3, the 

requirement to list at least 80 percent gets the most significant contributions, but not all of them. 

There was considerable discussion on what the panel members would need to provide for Question 3A.3 

and 3A.4. We need to know what variables have a major impact on LOCA frequencies to quantitative 

that impact as best as possible. Lee Abramson tried to make the point that things would become clearer 

once individual panel members got into their elicitations and tried to put numbers to the answers to the 

questions. 

A question was asked if any attempt is going to be made to look at plants and determine how many plants 

have a certain combination of variables (V1, V2, V3, etc.), and how many plants have another set of 

variables (V2, V4, V5, etc.), and how many have another set. It was noted that some variables will be

important at certain plants and other variables will be important at other plants. Rob indicated that it 

would be nice to have such information, but it was not practical to get such information in the time frame 

we have. This could possibly be a follow-on effort. However, it should be stressed that the plant design 

information will not likely result in a significant change in the analysis. If certain designs do not

contribute to the LOCA frequencies then they are not significant contributors and the panelists can focus 

on designs that do as long as they exist in several plants. If the population of the significant contributors 

is in error by 2 to 3, it will likely not matter. 

The only issue to avoid during quantification is if you believe that only a few (1 – 2) plants of a certain 

design, operating experience, etc. significantly contribute to the generic LOCA frequencies. These plants 

should not be explicitly considered in these generic estimates. However, but possibly applicability of 

these generic results to those design conditions should be discussed during the elicitation. 

Elicitation Question (EQ) 3A.1 compares a base case to a reference case, then EQ 3A.4 will use the 

impact of the important variables to compare other similar piping systems to the reference cases. 

Reference cases are the link back to the base case for which we will have actual LOCA frequency 

estimates. The panel members don’t have to do the mapping back to the base cases, the facilitation team 

will do that. Then the facilitation team will filter the results up (bottoms up approach) to get an overall 

LOCA frequency. 

There was some discussion about how the facilitation team would integrate these results and how the 

panelists could account for these individual contributions. 
B-52

Again, it was emphasized that if panel members are not comfortable in answering specific questions, then 

they need to say so. If the panel members need to make a crude assumption, then do so, but indicate that 

during the individual elicitations so the facilitation team can help estimate the level of uncertainty. Dave 

Harris thought that it won’t be clear to him how this all fits together until he starts the process. Then he is 

sure that he will have a lot of questions. Rob Tregoning and Lee Abramson told him, and the rest of the 

panel members, to call them for any clarification of questions. Rob also indicated that he will be 

contacting each panelist prior to their elicitation to discuss issues. 

The flow chart on slide 9 is for a “top down” approach, much in the motif of what Bill Galyean discussed 

on Day 1. As part of this approach, one only needs to tie one system (of those identified as being 

important contributing systems to LOCA frequencies) to the base case since previously we had identified 

individual piping system contributions. One can make this connection through a reference case if not a 

base case system, or can tie directly to the base case if the system is a base case system. This approach 

may be more straightforward for people who need to integrate variables in mind, which might lead to 

more uncertainty. 

Panel members can chose which approach to follow (top down or bottoms up), and they can switch back 

and forth depending on different systems. As part of their homework prior to their elicitation, they only 

need to do one approach, but the facilitation team may ask about each approach during the elicitation. 

There will be different tables to fill out with each approach. The bottoms up approach may be more 

rigorous with less subjectivity, but the top down approach may be easier to understand. One can get top 

down answers from the bottoms up approach, but can’t do reverse. By doing both approaches for certain

systems, panelists can search for consistency in their analysis. 

Rob discussed the elicitation questions related to non-piping components starting at slide 12. At the kickoff meeting in February we didn’t spend as much effort developing the base case and reference cases for 

non-piping components as we did for the piping systems. Thus, these gaps needed to be filled during the 

remainder of this meeting. 

Slide 12 illustrates the flow chart for the “bottoms up” approach for the non-piping components. At the 

kick-off meeting, we had identified five (5) non-piping components to consider (pressurizer, valves, 

pumps, RPVs, and steam generators). In order to estimate the frequencies for these non-piping 

components, the panel members need to pick either a piping or non-piping base case for comparison. The 

facilitation team will integrate the results in a manner similar to the bottoms up approach for piping

systems. 

There was a question about the nature of the non-piping base case conditions, especially in light of the 

thorough discussion about the piping base cases during the previous day. We had originally planned to

have precursor data for the non-piping base cases for comparison. However, we do not have data 

identified yet. We may have to drop the idea of using non-piping base cases and only have piping base

cases to compare to. We will come back to this issue later in the afternoon. 

Bruce Bishop felt that tying non-piping components back to piping base cases would be difficult. He 

foresaw lots of dissimilarities between non-piping and piping in failure mechanisms, etc. He suggested 

that we try our best to come up with some non-piping base cases. Even if we can’t come up with base 

cases for all 5 of the components, if we could come up with base cases for a few, that would be better than 

nothing. 

Pete Riccardella asked if we have defined the failure mode for these non-piping components. Rob felt 

that we don’t have a clear definition at this time. Rob felt that we had to take more of a mechanistic 

viewpoint. There was also general confusion about the definition of “failure mode” for the non-piping
B-53

issues. Rob indicated that this means the failure mechanism. It was agreed that “failure mechanism” is a 

preferable term and Rob will change the phrasing for the elicitation questions from “failure mode” to 

“failure mechanism” to avoid any confusion. 

It was again emphasized that the panel members will not be asked to provide absolute LOCA frequencies. 

However, if a panel member prefers to think in terms of frequencies, they should feel free to do so. 

Elicitation questions will ask for relative comparisons with the base cases and other conditions. The

facilitation team will then make the calculations to get the absolute LOCA frequencies. The panel 

members should make the best comparisons possible and are not compelled to answer questions in areas 

where they have no expertise. 

Rob showed an example of a table that he may provide the panel members for them to fill out for EQs 

3B.1 and 3B.2 for the top down approach, see Table B.2.2. 

The complete set of tables to be filled out for the elicitation will be provided electronically by Rob. There 

will be a space for comments in each table row to initiate discussion during the elicitation process. The 

tables will be provided in Excel format. If a panel member wants to change the Excel spreadsheet format 

they should feel free to do so as long as the cell references for each answer remains unchanged. The final 

calculations will be done in Excel. Therefore, the elicitation results should be provided to Rob in the 

excel spreadsheets if at all possible. Hardcopies or MS Word versions of the tables can provide upon 

request.

Rob will provide the panel members with a copy of the spreadsheet that he will use to calculate LOCA 

frequencies sometime during the elicitation process. Rob will attempt to complete this spreadsheet to the 

individual elicitations so that the panel members can see how their responses reflect their calculated

LOCA frequencies. However, this will be a lower priority than coordinating the information exchange 

among the expert panel and finishing the base case calculations. 
B-54

Table B.2.2 Elicitation Questions 3B.1 & 3B.2 

BWR Piping Systems: Important System Contributions to LOCAs 

 25 Years of Plant Operation 40 Years of Plant Operation 60 Years of Plant Operation 

LOCA 

Cat. 

Systems System 

Cont. 

5% 

LB 

5% 

UB 

Systems System 

Cont. 

5% 

LB 

5% 

UB 

Systems System 

Cont. 

5% 

LB 

5% 

UB 

 

1 

Total Total Total 

2 

Total Total Total 

3 

Total Total Total 

4 

Total Total Total 

5 

Total Total Total 

6 

Total Total Total 

Presentation on Non-Piping LOCA Evaluation: Base Case Data and Remaining Issues 

by Rob Tregoning (USNRC) 

Prior to this presentation, Rob noted that this presentation was not included in the handout, but will be put 

on the ftp site. Rob also asked that the base case team provide him in an electronic format with any 

references that they used so that the references can be put on the ftp site. 

It was first noted that we have not been successful in locating additional failure data for several of the 

components where we were lacking data. Fred Simonen had spoken with Spencer Bush and was not able 

to locate additional data. Rob and others were also somewhat unsuccessful. 

Based on the leak rate versus opening area from one of the prior presentations, Pete Riccardella 

questioned if we needed multiple failures for the heater sleeves as shown in slide 3 in Rob’s presentation. 

Bruce Bishop indicated that thermal fatigue needed to be added to the degradation mechanisms for vessel 
B-55

head bolts (slide 4). Rob indicated that these tables are not filled in completely at the present time. These 

tables are neither comprehensive nor as complete as the piping component table. We focused more on the 

piping issues at the kick off meeting than we did on the non-piping issues. 

For all of the major groups, we initially listed at least one component (bolded item in the original nonpiping tables in presentation and kick-off meeting notes document) where the panel thought that failure 

(leaks or cracks) data is available. It was thought that these bolded items represented potential nonpiping base cases for anchoring. If these non-piping base cases can’t be develop, then we will have to use 

a piping base case for anchoring. As Bruce Bishop indicated earlier that would be an unnatural 

comparison, making it somewhat difficult. The question was asked how we develop these non-piping 

base cases. It was thought that areas where we could come up with passive-system failure data, e.g., 

steam generator tubes, would be logical first choices. It was thought that data on steam generator tube 

failures should be easy developed. 

Bruce Bishop indicated that there was an INEEL NUREG report by Vic Shah that has piping and nonpiping failure data that could possibly be used for base cases. Obviously, this would be a good place to 

start. It was noted though that this INEEL report is for PWRs only, and the steam generator tube failure 

data will be included in this report. Bill Galyean and Rob Tregoning are to locate and distribute copies of 

this report to the panel members. 

Pete Riccardella volunteered to run a base case analysis for feedwater nozzles and the belt line region for 

the RPV due to LTOP. Pete thought he could have some analysis results by the end of the month. These

would be for BWRs and will be done using predetermined flaw distributions. Note, to date there have 

only been cracks, there have been no leaks to date.

Another potential non-piping base case for anchoring is the PTS study for the belt line region of the RPV. 

This would be for PWRs. Rob Tregoning will extract this data. Gery Wilkowski noted that we should 

use caution in using the PTS results and should only use the contribution from non-pipe break transients 

to ensure that the comparison is consistent. In a related action, Bruce Bishop volunteered to provide a 

Westinghouse nozzle study for PWRs and will also provide PWR vessel failure probability for areas 

outside the beltline region covered by the PTS study. 

Pete Riccardella suggested classifying the alloy 600 penetrations as non-piping failures for consistency 

with other nozzles. A suggestion was made to move the in-cores and CRDMs to the non-piping category. 

Fred Simonen suggested putting them all under a separate category called vessel penetrations, with a 

separate bin for RPVs and pressurizers. Karen Gott and Pete Riccardella are to create a base case for

penetrations using the CRDM data based on a prior MRP study. Rob Tregoning will move all the vessel 

penetrations from piping to the vessel bin and supply updated tables. 

Bengt Lydell has a non-piping data base that he will query by end of the month. He will also query the 

IRS data base (Incident Reporting System by INEA) by end of the month. Note the IRS database only 

includes data countries chose to include. It was also noted that MITI and NUPEC (both in Japan) have a 

data base for non-piping components. The NRC supposedly has a copy of this database. Rob will check 

into relevancy and availability. Bill Galyean volunteered to examine his database to see any relevant nonpiping events, i.e., stream generator tube rupture statistics, incidents of bolting connections, etc., by the 

end of the month. Rob will be the conduit for getting the results from the various individuals searching 

the databases out to the rest of the group. Results will be posted to the ftp site and more important items 

will be bulk emailed to the panel.

The question was asked if the panel members could take home the modified tables for the pressurizer, 

RPV, pumps, valves, and steam generators and fill them out and return them back to Rob within two 
B-56

weeks. Rob will modify these non-piping tables (Tables B.1.7 and B.1.8 from the First Elicitation 

Minutes) with the additions made during the second meeting discussions and send them out to the panel 

members. Each panel member is to modify these non-piping tables and get them back to Rob Tregoning. 

Fred Simonen has an electronic version of the GALL report (Generic Aging Lessons Learned) which 

identifies the key degradation mechanisms that could be used to help fill out these tables. Fred will 

extract the relevant tables to be used in identifying the key degradation mechanisms. 

Presentation on Conditional Seismic Evaluation 

By Rob Tregoning (USNRC) 

Each base and reference case includes at least one transient since transients are needed to initiate a LOCA 

event. At this time the transients are poorly defined. 

Fred Simonen has seen reports that show seismic stresses, but he has no idea of the magnitude of the nonseismic transients (e.g., water hammer, safety relief valve transients). He and Gery Wilkowski would like 

help in establishing a rough order of magnitude for these types of transients. This information could be 

best expressed as a percentage of the Service Level stresses. Pete Riccardella indicated that SRV 

transients could be on the order of a small earthquake, just with a higher frequency. Bruce Bishop agreed 

to provide some water-hammer transient stresses for the pressurizer. Gery Wilkowski volunteered to 

provide some summary information from past probabilistic LBB analyses

Bengt Lydell indicated that the water hammer frequency is about 5E-3. There are more water hammer 

events on the secondary side, but there are design basis events that can cause water hammer on the 

primary side. It as noted again that without a transient a large LOCA is highly unlikely. The cracks will 

just leak until they are detected, and then will be repaired. Long surface cracks that don’t leak are drivers 

for large LOCAs. 

Guy Deboo (Consolidated Edison) volunteered to provide stresses and frequencies for transients (e.g., 

feedwater line water hammer, SRV, and seismic from the LaSalle plant). Gery Wilkowski to provide 

some tables from NUREG/CR-6004 showing the N+SSE stresses for about 30 piping systems. This data 

was originally developed for the ASME Section XI Working Group on Pipe Flaw Evaluation. 

Additionally, Gery Wilkowski and Guy Deboo will provide stresses (not frequency) for some large 

faulted loads that are not really expected to occur over the life of plant. Gery is to examine the N+SSE 

stresses from the USNRC LBB submittal database. Guy noted that a 1SSE amplitude earthquake, based 

on seismic hazard curves, is expected to occur once over 40 years (design basis). The frequency (not 

amplitudes) of the seismic hazard curves are generally considered to be conservative. Gery noted that the 

design basis apparently is conservative since he is sure that an SSE event has ever occurred at any US (or 

other) plant. Hence, the seismic event frequency could perhaps be down graded to 0.5/2,600 events/year 

rather than 1/40 events per year. 

For the smaller transients, Dave Harris will extract stresses from a NUREG report by Fred Simonen. Guy 

Deboo, Pete Riccardella, and Sam Ranganath will provide some data on normal operating transients. Pete 

Riccardella will get some data showing comparison of design versus actual transients based on some 

thermal fatigue analysis from a Sandia report. 

Bruce Bishop has some plant specific ISI data that he could provide which provides transient information, 

but he won’t be able to provide it expeditiously due to other commitments. In fact, it is unlikely he will 

be able to provide this during the timeframe of this effort. Pete Riccardella concluded that the actual 

transients were not as severe as the design basis transients, but there are typically more of them. 
B-57

It was also agreed that the group should have isometric drawings for the LOCA-sensitive piping. Bengt

Lydell will inventory his electronic drawing database. The purpose is to look at generic systems to get an 

idea of how many welds are involved, pipe sizes, etc. Guy DeBoo will also evaluate the ISO drawings in 

his archive. Bengt will coordinate with Guy Deboo on this effort. For the time being it was decided that 

we would not seek additional isometric drawings until we determine if we are missing any of the major 

piping systems. Guy Deboo will then help obtain drawings (e.g. ISI drawings) for missing systems that at 

least indicate the number of welds. In general, multiple isometrics of similar systems are useful since 

each plant design is somewhat unique. 

Dave Harris thought we only needed census data, i.e., number of welds as a function of pipe sizes, etc. 

Dave will provide a census that he has developed of number of welds for the base case piping systems. 

Dave and Bengt Lydell will coordinate on this action. Gery Wilkowski indicated that there is a MRP 

report with locations and numbers of bimetal welds that may be useful to review. Gery will provide a

table of Inconel weld locations in different piping systems from the MRP-44 report. 

Rob Tregoning asked if the panel should consider redefining the base and reference cases. The biggest

concern is that the loadings and the mitigation/maintenance may need to more accurately reflect the 

operating experience. The panel could redefine these variables in a way that is more consistent with the 

quantitative analyses that was presented for the base cases earlier. The approach is to define these base 

and reference cases load and mitigation variables to reflect historical plant operating experience for the 

first 25 years of plant life (e.g. the current LOCA estimate). The consensus opinion agreed with this

proposal. Guy Deboo added that he would like to see the base cases run out to 40 and 60 years, with a

seismic event included. It was again noted that the objective of the “current estimate” analyses (i.e., out 

to 25 years of plant life) was to provide a benchmark against historical data. 

Rob also asked if the panel wanted to consider more than one degradation mechanism for each reference 

case since the operating experience contains contributions from all applicable degradation mechanisms. 

Rob Tregoning wants to make sure each panel member is making the same relative comparisons with the 

reference cases and that these relative comparisons are natural. However, the group consensus was that it 

is easier to use the reference cases for anchoring when only considering one degradation mechanism and

that no other changes in the reference cases should be adopted. 

Vic Chapman argued that if we run the probabilistic fracture models for the first 25 years for 

benchmarking purposes, and we see a failure, then we should exclude that result since in reality we have 

not seen any failures to date. Lee Abramson agreed with this thought. The operating experience-based

estimates developed by Bill Galyean and Bengt Lydell are inherently benchmarked in this manner. The 

probabilistic fracture models (David Harris and Vic Chapman) still need to benchmark their data. 

Bruce Bishop would like Karen Gott, Bill Galyean, and Bengt Lydell to extract the failure mechanisms as 

a function of piping system from their databases. Bill will have one of his colleagues do this. Bengt and 

Karen will query their databases to get a list of degradation mechanisms as a function of piping system. 

Bruce Bishop would like someone to publish a list of plants by design type. Rob Tregoning indicated that 

he would provide this information. 

Several people wanted Dave and Vic to run their models using refined stress histories. Pete Riccardella 

agreed to redefine the loads for the HPI/MU nozzle. Gery Wilkowski volunteered to get some more 

realistic surge line stresses for the surge line elbow case. It should actually be for a crack in the girth 

weld at the elbow, not a crack in the body of the elbow. 

Bill Galyean and Bengt Lydell will determine the frequency of IGSCC leaks and surface cracks as a 

function of time and pipe size using their respective databases. The surface crack data should be 
B-58

characterized by the length and depth of the flaws if possible. The Swedish data should be separately

characterized from the US plant data, since the US plants may not have characterized the flaw shapes by 

ISI if the piping was being removed from service. In Sweden, flaws from removed pipe systems have 

been characterized. A further requirement of this query to examine the recirculation system piping 

studied in the base case would also be helpful. Bengt will provide all this information in an Excel 

spreadsheet. 

Presentation on Leak Rate versus Pipe Break Size Evaluation 

By Rob Tregoning (USNRC) 

Rob started by reviewing the history behind this analysis. Rob was not able to find a reference for the 

basis of PWR correlations developed in NUREG-1150. The BWR correlations were extracted from GE 

NEDO studies performed in the early 1980’s. New correlations have been based on several closed-form 

solutions appropriate for BWR steam and liquid lines and PWRs. 

Rob still needs to add a column correlating the pipe diameter to the leak rate/area in slides 8 of 9 of this 

presentation. It was agreed that the pipe diameter should assume a single ended guillotine break. Rob 

also agreed to change the word “axial” to “transverse” on slide 9 of the original presentation which relates 

the pipe fracture area to transverse piping displacement. Gery Wilkowski noted that once a full DEGB 

occurs that the two ends of the pipe are jets that move away from each other so that he would be hesitant 

to use the analysis on slide 9 which assumes a transverse displacement. The base case team members will 

update their results using the new pipe break size to leak rate correlations developed for this presentation. 

The final topic on the agenda was a discussion on the schedule for the elicitations. 

Rob Tregoning and Lee Abramson want to do 2 elicitations early, possibly the week of July 14th. Then 

take a month off and restart the elicitations in mid August with the goal of completing all of the 

elicitations by the end of September. That would leave about a month to analyze the results. Rob is 

looking for volunteers to be the first two individuals to go through the elicitation process. All panel 

members should give their schedule to Rob Tregoning so that the individual elicitations can be scheduled. 

Once the elicitations are complete and the results analyzed, a wrap-up meeting will be held in the 

October/November timeframe. As part of this meeting the calculated LOCA frequencies will be 

presented and any interesting and surprising results from individual questions will also be presented. The 

panel will also be solicited for feedback on the process. We will try to identify strengths and weaknesses 

with the process. Any necessary follow-on work will be defined during this meeting. We are also open

to suggestions for conducting a reanalysis of these results in ten years. It is anticipated that the wrap-up 

meeting will take two days. 

Again, it was noted that the panel members can change the results after their individual elicitations and 

after the wrap-up meeting. However, the final report must be submitted by the end of December. It is too 

premature to speculate on the form of the final report. However, confidentiality of the individual 

elicitation opinions will be maintained. 
B-59

MEETING NOTES FROM US NRC LOCA ELICITATION WRAP-UP 

MEETING 

DOUBLETREE HOTEL, ROCKVILLE, MD 

FEBRUARY 10-12, 2004

Day 1 – February 10, 2004

Rob Tregoning from the USNRC welcomed everyone and explained logistics for the meeting. Rob had 

everyone introduce themselves. Next, Rob reviewed the agenda for the three-day meeting. Day 1 would 

focus mainly on piping; Day 2 on non-piping, and Day 3 on emergency and faulted loads plus soliciting 

feedback on the process. There are no results to present on the topic of emergency and faulted loads. 

Only the basic approach will be shown. 

Presentation #1 – Elicitation Project Plan, Schedule, and Milestones 

By Rob Tregoning 

The NRC has initiated some ongoing work looking at active mechanisms, e.g., stuck open valves. Bill 

Galyean is doing this. 

There is a SECY paper due to Commissioners on March 31, 2004 with LOCA frequencies for normal 

operating loads. 

Rob will distribute a draft NUREG documenting expert elicitation results so the panel can provide 

feedback on the NUREG. Rob expects that the panelists will only a have short time (~2 weeks) to 

provide feedback. 

For the April-June public meetings, Bruce Bishop from Westinghouse suggested meeting with the WOG 

risk-based group. 

Most critical future milestone is finalizing individual expert responses for normal operating loading 

frequencies by February 25th

. 

Presentation #2 – Elicitation Results (Box and Whisker Plots 

By Rob Tregoning

Rob made a presentation on the details of the box and whisker plots that will be shown over the next 3

days. Many different methods of calculating percentiles; we used Standard method; fundamental message

is that doesn’t make much difference in final analysis. 

Presentation #3 – Safety Culture 

By Rob Tregoning 

Bruce commented on VG4 with respect to poor US safety culture that didn’t see a problem until starting

seeing circumferential cracks; Pete commented that it was an economic issue since US utilities charge 7 

to 10 cents per KW-hr while overseas may charge 40 cents per KW-hr 
B-60

Karen Gott pointed out that her experience was that the first time a plant experiences a problem it is a big 

problem so she would agree with second major bullet on VG#4; on second sub-bullet she thought a better

word is experience instead of sensitive. 

On VG#5 it was pointed out that while the NRC may have only one vote on code changes, it has the 

ultimate veto vote. 

Bruce felt day of single plant utility is numbered. Helmut Schulz and Karen disagreed with the last subbullet on VG6; utilities are willing to invest in older plants since they are already paid for (less capital 

investment); may be a difference in international experience and US practice. 

Helmut commented with regards to VG9 on decommissioning that they have not seen any increase in 

LER events in the last few years before decommissioning for those plants that were decommissioned. 

With regard to VG10 and negative bullets related to risk-informed regulation. Bill Galyean commented 

that utilities have limited resources and risk informed process helps prioritize; Vic Chapman cautioned 

against turning crank and getting an answer without thinking of why. 

Bruce commented that end of plant license renewal may be 80 years not 60; based on comment made at 

NRC recent meeting. 

Helmut pointed out that boric acid corrosion of manway bolts of 15 to 20 years ago was more serious 

from a LOCA perspective that Davis Besse head problem of today. 

Bottom line is no effect of safety culture on LOCA frequencies. There will be no adjustments to 

frequencies; no major discussion on part of panel with regards to this bottom line conclusion. 

Presentation #4 – Piping Base Case Evaluation I 

By Bengt Lydell 

Bengt used a bottoms up evaluation based on operating experience. 

Markov is standard approach common to any advanced reliability approach; this was the technical basis 

used by Bengt to develop time dependency. 

Bengt’s model allows for imperfect repairs or inspections. 

Can go from S (unflawed condition) to R (rupture condition) if have some extraordinary event such as gas 

accumulation at Hamaoka in Japan. 

VG13 results are per “weld year”; some of earlier plots are “per reactor year”. 

Presentation #5 – Piping Base Case Evalution II 

By Dave Harris 

Vugraph #3 is a summary of revised results since July 2003. 

Could use VC Summer Hot Leg/RPV nozzle weld crack as benchmark, but need to be careful to consider 

all aspects such as differences in weld residual stresses due to repairs. 
B-61

VG12 shows an order magnitude difference in LOCA frequency for a difference of 2 ksi (14 MPa) in 

normal operating stress; this was thought to be pretty sensitive result. 

Dave assumed a linear residual stress field through the thickness due to the weld overlay repair. 

The solid line in VG13 is PRAISE result while prior and post symbols come from Bengt’s results. 

Presentation #6 – Elicitation Question I: Base Case Evaluation 

By Rob Tregoning 

There was quite a bit of disagreement on first bullet on with regards to the perceived disadvantages of the 

various approaches. 

Bruce argued that PFM approaches have been benchmarked against operating experience and shown to 

agree well. 

Helmut and Bruce stressed that we are not in a position to review various approaches and we should not

provide such a review in the NUREG report. 

Need to stress in NUREG that general comments are not a group consensus, but individual responses. 

Presentation #7 – Elicitation Calculation Framework 

By Lee Abramson 

Split distributions necessary if UB and LB are not symmetric with respect to mid value. 

Used log normal distribution since results provided by participants fit log-normal distribution; also lognormal distribution easily to manipulate; also tradition is that log-normal is used in risk based approaches; 

bottom line is that due to variability in responses should not make that much difference as to what 

distribution chosen. 

Analysis will yield medians of mid values and bounds as well. 

Presentation #8 – PWR Piping 

By Rob Tregoning and Paul Scott 

The second main bullet on VG7 should be decrease with “decreasing” piping size. 

Dave Harris disagreed with comment that PFM models had problems modeling mitigation. 

There was a problem with VG16 with interpreting results for Panelist L. 

A number of the panelists were surprised with VG17 that surge line results for Cat 5 are comparable to

that of cold leg. 

There is a discrepancy in maximums for Category 6 LOCAs at 25 years between VG19 and 20. VG20 

shows participant L having the maximum value while VG19 shows participant B as having the maximum 

value. 

Participant J showed the most impact of age on the LOCA frequencies; really obvious in 

VG21. 
B-62

Bruce pointed out that for higher Category LOCAs that there was less uncertainty; may be an artifact that 

there are less systems than can contribute (only large diameter); also these larger systems are better

inspected, i.e., better controlled. 

Every plot shown is for 25 years unless specifically stated. 

Presentation #9 - BWR Piping 

By Rob Tregoning and Paul Scott

Similar degradation mechanisms as with PWR mechanisms: thermal fatigue, mechanical fatigue, 

Sam commented with regards to Slide #4 that analysis of BWR feedwater says it should crack, but don’t 

find these cracks in service. 

For Slide 8 it was suggested to change “risk” to “high” in last bullet. 

Maximum diameter of RWCU system is 6 inch; not 24 inch as shown in VG17. 

Day 2: February 11, 2003

Rob started the second day at 0810 by reviewing agenda for the next two days. 

Presentation #10 – Non-Piping Database Development 

By Bill Galyean 

LER database at ORNL will no longer be available after 2/29/04; the database will be moving to INEEL 

but in a different format. 

Failures defined as leaks or cracks. 

Presentation #11 – Non-Piping Base Case Development: CRDM and LTOP LOCAs 

By Pete Ricaradella 

Pete used the VIPER program to predict beltline failure frequencies (per vessel year) for typical BWRs. 

VG3 and 4 are frequency plots and not probability plots. 

For VG7 for large Category LOCAs, see big impact with time between 40 and 60 years; attributed to 

effect of radiation embrittlement; for smaller LOCAs don’t see much effect of time. 

EDY stands for Effective Degradation Years; used to normalizes degradation to a reference of a 600o

F 

operating temperature. 

For CRDM nozzle ejection probability the assumption in VG11 that immediately have circumferential 

TWC of 30 degrees is highly conservative according to Bruce in that most are axially oriented. 

Of 30 plants, there were 11 nozzles that had circumferential cracks; all of these plants were at about 20 

EDY so Pete could take time factor out; total number of nozzles in 30 plants was 881 
B-63

POD curve for NDE; cracks were EDM notches that were compressed to make them tight; eventually will 

get some real cracks from the North Anna head that can be used for calibration/validation 

VG15 shows the probability of leak in one of 98 nozzles in this plant per vessel year; shows effect of

NDE on probability of leakage 

VG15 and 16 show effectiveness of NDE and how PFM models can account for inspection in their 

analyses. 

VG17 shows decreasing frequency with time which reflects benefit of inspection. 

To get 5,000 gpm (19,000 lpm) leakage, need ejection of 2 nozzles; most likely scenario is for collateral 

damage as one ejects and causes damage to adjacent nozzles. 

Presentation #12 – Steam Generator Tube Rupture Frequencies 

By Rob Tregoning 

Almost everyone agreed that for PWRs the dominant failure scenario for Category 1 LOCAs was steam 

generator tube failures. 

Used non-piping database which was augmented back to 1987 to capture 2 major events in ‘87 and ’89. 

There have been 15 leaks since 1990 with 4 events over 100 gpm (380 lpm) since 1987. 

On VG2 the reference to Nine Mile Point should be to Indian Point. 

Fred Simonen stated that usual scenario assumed for higher category LOCAs is common cause failures 

such as losing pressure on secondary side causing pressure differential across tube and failure of multiple 

already degraded tubes. 

Rob’s analysis of independence (ignoring common cause) was viewed with a great deal of skepticism;

Bill Galyean commented that if look at LERs always see multiple tube degradation but only see single 

tube rupture in the LERs. 

This analysis is for 25 years, panel members left to their own devices for later years. 

Presentation #13 – Overview of PTS Re-Evaluation Project 

By Rob Tregoning

For this analysis all of the crack growth is from a PTS event; not fatigue. 

For plants with multiple pass cladding there is a very low probability of flaw penetrating multiple passes; 

exception was Oconne that was single pass cladding.

Big driver were flaws between plate and axial flaw region. 

For those that want to anchor against PTS, Rob will use updated results base on some average values for 

Oconee, Beaver Valley, and Palasdies 
B-64

Presentation #14 – Elicitation Question VI: PWR Non-Piping 

By Rob Tregoning 

Analyses tend to get easier as go up on LOCA sizes in that less systems to worry about. 

As a group the panel was uncomfortable with comments on Effect of Operating Time for Category 4 

LOCAs. 

Bill Galyean suggested that for presentation to public, may want to consider some other means of 

reporting extremely low frequencies (~1e-15); Rob wasn’t sure he could define a cut off value; also when 

look at median values, these low numbers don’t impact final answer; Helmut supported this approach. 

Helmut suggested that we explicitly state that a major assumption was that everything was fabricated in 

accordance with Code standards; no counterfeit bolts, etc. 

Participant J only has fatigue in his analysis, once he includes other mechanisms, then his frequencies will 

go up by 2 or 3 orders of magnitude, but will still be low and probably will not alter the final median 

results. 

Presentation #15 – Elicitation Question VI: BWR Non-Piping 

By Rob Tregoning 

Two of the six respondents who responded to this question only provided frequencies out to Category 5 

LOCAs even though BWR non-piping could cause a Category 6 (Vessels) LOCA; thus Rob only provides 

results for Category 5 LOCAs here. 

Concern with thermal aging of cast stainless steel is when it is present in concert with some other 

degradation mechanism. 

CRDM refers to stud tube housing on bottom head of BWRs. 

Panelist F didn’t consider RPV for Category 1 LOCAs; question was whether they didn’t think important 

or did they not have a means of making an evaluation. 

There is much less spread in results for RPV than valves and pumps; panelists spent more time and more

work in past on RPV than pumps and valves. 

Panelist C sees a decrease in freq with time but may be an artifact of the fact that he anchored against 

BWR recirculation line that shows a decrease in LOCA frequency with time; he provides no rationale for

why he would expect non-piping LOCA frequency to decrease with time. 

Presentation #16 Piping and Non-piping combined Results 

By Rob Tregoning 

Ratios of non-piping to piping for various category LOCAs are for 25 years only. 

Pete made the point that most of the plots that Rob has shown are for 25 years, while he thought 40 and 

60 years more important since 25 years is in past and the associated problems have been addressed while 

40 and 60 years are for future; Rob responded that not that much difference between 25 and 40 years with 

some effect for certain category LOCAs at 60 years.
B-65

Sam commented that he was somewhat surprised that non-piping contribution less than piping 

contribution for BWRs in that piping has some active mechanisms that have been successfully mitigated 

in past; Karen responded that non-piping components more robust. 

Bill Galyean warned about combining group distributions and panel distributions that may introduce a 

bias in that various members of group defined boundaries of system differently. 

Much discussion on whether to chose group median or individual medians; Rob and Lee haven’t done 

panel distributions yet. 

Comparison of BWR and PWR – effect of mitigation encompassed in results for BWRs, but not PWRs – 

BWRs have been doing mitigation for 15 to 20 years whereas PWRs are just starting with mitigation for 

PWSCC. 

Inclusion of S/G tube rupture for Category 1 LOCAs will be problematic for some people in that PRA 

people aren’t used to accounting S/G tube failures in with rest of data; typically S/G tube rupture data is 

presented separately; in future Rob will present data both with S/G tube rupture data and without. 

Categories 1, 2, and 3 are historically same as small, medium, and large break LOCAs respectively. 

For PWR MB LOCAs, major contributor is CRDM, not S/G tube ruptures. 

Discussion of whether MB LOCA was Category 2 or Category 3; some thought that MB LOCA was more 

in line with Category 3 LOCA. 

Bengt felt we are comparing apples and oranges as we try to compare our results with historical results; 

NUREG/CR-5750 didn’t look at non-piping per se whereas we did, although Bill Galyean indicated that 

if there had been indications of TWC in non-piping components then he would have included that data in

his analysis in 5750; some thought that due to apples and oranges nature of our approach with 5750 that 

we shouldn’t present these comparisons but Rob argued that if we don’t present these comparisons then 

others will; Some argued that we should present frequencies for multiple LOCA categories when we 

compare with small, medium, and large break LOCAs for 5750 

Lee reviewed the feedback questionnaire. 

Day 3: February 12, 2004

Presentation 17: Emergency and Faulted Loading: Elicitation Approach and Responses 

Water hammer type loadings should be in normal operating loading history. 

What we are asking panel members to estimate is only the conditional failure probability given a stress 

with magnitude i (PL/Si) 

Ken commented that the seismic anchor motion (SAM) stress which is a secondary stress may be a bigger 

contributor than some primary stresses such as inertial stresses. 

On VG entitled Elicitation Requirements, we are asking panel to do first bullet, we will do 2nd bullet, and 

plants would do 3rd and 4th bullets. 
B-66

Asking them for a given system and degradation mechanism for their estimate of L50, P50, Ppl, Lpl, Ltsl, 

and Ptsl and then we will interpolate to get entire curve. 

Some people argued during their elicitations that non-piping and large piping are non contributors to 

LOCA frequency due to seismic; we will look at results from piping and then decide what and if we will

do anything for non-piping considerations. 

Bruce argued that can get some very high loads due to malfunctions of snubbers. 

Pbc and Lbc are probabilities for base case and likelihood of base case 

Presentation 18: Remaining Work 

By Rob Tregoning 

Rob discouraged panel to make changes to bring their results more in line with others, would encourage

panel to make changes if they heard something technically that made them rethink their answers. 

Everyone will be involved in reviewing and critiquing NUREG reports; everyone wanted to be involved 

with the process. 

The question of how: possibly another meeting, VTC, circulate vugraphs for review and feedback 

(electronically); possibly couple with some other meeting (ASME, PVP, etc). 

Karen and Dave would want to meet before the NUREG was finalized; others seemed to agree with this. 

Ideally we would circulate draft, we would then get comments back, we would then synthesize comments 

and then feed them back to the group and then meet; all before finalizing NUREG. 

Helmut and Bruce supported idea of VTC (maybe limit to a few sites). 

Other option is provide slides; review slides on computer and then have a conference call to review; limit 

to a few hours at a time (bite off small chunks). 

Presentation #19 – Remaining work on Active Systems 

By Bill Galyean 

PORV stands for pilot operated relief valve. 

Difficult to correlate stuck open valves categories with leak rate sizes/categories; size of valves will vary 

between plants. 

Presentation #20 – Emergency and Faulted Loading Base Case Development 

By Gery Wilkowski 

No uncertainties applied to loads in Gery’s analysis. 

Base case assumes idealized TWC geometry. 

Did not do any subcritical crack growth; thus did not consider residual solutions. 
B-67

LBB.ENG2 is in form of closed form solutions. 

For predicting large crack growth in a pipe tests it is better to use J-M than J-D. 

Duane Arnold crack would have failed if subjected to Level B type loading. 

Global secondary stresses act as primary stresses if crack large enough such that failure stress is below 

yield strength. 

Presentation #21 – Lee reviewed feedback 

It would have helped to have had Gery’s presentation earlier, before the panel tried to answer seismic

question. 

Would have been nice to have a video of plants showing various systems as one tours plants with video 

camera. 

Amount of information available was overwhelming; try to do a division of labor so one or two people 

review something and provide a tutorial to others so everyone is working from same basis; otherwise 

everyone is inventing the wheel themselves; maybe have a meeting to review these tutorials. 

Need a roadmap of where information can be found. 

Periodic/weekly update of changes made to ftp site; alternatively an alert message when something added 

to site; maybe a readme file when something added and what was added and when. 

NRC management must make sure that staff are available to panel members during the process; Rob 

getting pulled off for Davis Besse was a problem; delayed things and then panel members only had a few

weeks to respond at the end. 

Bruce would like time at meetings to do actual work on elicitations because once they get back home they 

will get pulled off on to other things and won’t be able to get back to answering questions for a long time. 

APPENDIX C 

ELICITATION TRAINING EXERCISE RESULTS

C-1

APPENDIX C 

ELICITATION TRAINING EXERCISE RESULTS

As part of the panel member kick-off meeting in February 2003, elicitation training was provided for the 

elicitation panel. The training involved the panel members answering a series of almanac-type questions 

for which numerical answers were available. The panel members provided both their best estimate of the 

answer as well as relative ratios with respect to other quantitative responses. In this way the panel

members got an appreciation of the benefits of the anchoring process used throughout the elicitation 

process. 

C.1 Training Questions

The following questions were used in the training exercise. 

Q1. According to the 2000 census, how many men 65 or over were there in the U.S.? 

 

Q2. In 1995, how many American men age 65 or older suffered from the chronic conditions listed? 

Q3. What is the ratio of the rate for men 45- 64 years old to the rate for men 65 and older for each of the 

conditions listed? 

Q4. What is the ratio of the rate for men under 45 years old to the rate for men 45 - 64 years old for each 

of the conditions listed? 

The answer to Q1 is 14.4 million. The chronic conditions referred to in Q2, Q3, and Q4 and the 

corresponding answers are listed in Table C.1.

Table C.1 Correct Value (CV) Results to Elicitation Training Questions 

 Q2 Q3 Q4 

Condition Rate per 1000 (Age 45-64 rate) / 

(Age 65+ rate) 

(Under 45 rate) / 

(Age 45-64 rate) 

Arthritis 404.7 0.44 0.13 

Cataracts 125.1 0.13 0.11 

Diabetes 123.6 0.50 0.10 

Hearing Loss 366.8 0.56 0.20 

Heart Disease 362.4 0.40 0.17 

Prostate Disease 118.0 0.30 0.054 
C-2

C.2 Elicitation Training Responses

As described in Section 3.3.2, the panelists were asked to supply three numbers for each question: a MV, 

a LB, and a UB. The MV has a nominal 50/50 chance of falling above or below the correct value. The 

interval (LB, UB) has a nominal 90% chance of covering the correct value. 

The following tables summarize the responses made in the training exercise. There were between 15 and

17 sets of responses to each question. (Although there were only 12 panelists on the panel, members of 

the facilitation team were also invited to participate.) The number of respondents is indicated following 

each question. The table columns summarize the responses relative to the CV. The first column 

indicates the number of respondents where CV < LB, i.e., where the coverage interval fell above the CV; 

the third column indicates the number of respondents where CV > UB, i.e., where the coverage interval 

fell below the CV. Thus, the total of the first and third columns is the number of respondents whose 

coverage intervals did not cover the CV. The second column lists three numbers that summarize the set

of MVs provided for each row of the table. These are the lower quartile (LQ), median and upper quartile 

(UQ), respectively. About one quarter of the MVs are less than the LQ and about one quarter of the MVs 

are greater than the UQ. Hence the interquartile interval (LQ, UQ), denoted by IQI, contains about one 

half of the MVs. (These three summary statistics are used to construct box and whisker plots, as 

described in Appendix L.) For ease of reference, the rounded correct values are listed following the 

conditions for the Q2 - Q4 tables. 

Q1. According to the 2000 census, how many men 65 or over were there in the U.S.? (N = 17) 

(CV = 14.4 million) 

Table C.2 Summary of Respondent Results for Question Q1 

CV < Coverage Interval LQ, Median, UQ CV > Coverage Interval 

N = 3 16, 20, 28 N = 0 

Respondents tended to overestimate the CV. Since LQ = 16, about three quarters of the MVs were larger

than the CV. However, percent coverage at 82% was near the nominal 90%, with 3 (18%) lying above 

the CV and none lying below. 

Q2. How many American men age 65 or older suffered from the following chronic conditions in 1995? 

(N = 15)
C-3

Table C.3 Summary of Respondent Results for Question Q2 

 Rate per 1000 

Condition CV < Coverage Int. LQ, Median, UQ CV > Coverage Int. 

Arthritis (405) N = 1 135, 200, 400 N = 9 

Cataracts (125) N = 2 50, 150, 200 N = 2 

Diabetes (124) N = 0 90, 150, 250 N = 3 

Hearing Loss (367) N = 1 200, 300, 500 N = 5 

Heart Disease (362) N = 1 150, 200, 375 N = 6 

Prostate Disease (118) N = 3 125, 200, 375 N = 2 

Four of the six IQIs covered the CV, and the two which did not almost did. Three of the medians were 

above the CV and three were below. Thus, the MVs for the six conditions as a whole exhibited no 

systematic bias in estimating the CVs. However, the coverage intervals tended to underestimate the CVs. 

Of the 90 coverage intervals, 27 (30%) lay below the CV and 8 (9%) lay above. The average percent 

coverage of all 90 intervals was 61%. Over the six conditions, the percent coverage ranged from a low of 

33% to a high of 80%. 

Q3. What is the ratio of the rate for men 45- 64 years old to the rate for men 65 and older for each of the 

conditions listed? (N = 16) 

Table C.4 Summary of Respondent Results for Question Q3 

 (Rate for ages 45-64) / (Rate for age 65+) 

Condition CV< Coverage Int. LQ, Median, UQ CV > Coverage Int. 

Arthritis (0.44) N = 1 0.20, 0.30, 0.50 N = 5 

Cataracts (0.13) N = 2 0.10, 0.20, 0.30 N = 3 

Diabetes (0.50) N = 0 0.25, 0.40, 0.50 N = 3 

Hearing Loss (0.56) N = 0 0.25, 0.30, 0.30 N = 6 

Heart Disease (0.40) N = 0 0.30, 0.30, 0.50 N = 2 

Prostate Disease (0.30) N = 2 0.20, 0.20, 0.40 N = 3 

Five of the six IQIs covered the CV, but respondents tended to underestimate the CV. Five of the six 

medians were below the CV. Of the 96 coverage intervals, 22 (23%) lay below the CV and 5 (5%) lay 

above. The average percent coverage of all 96 intervals was 72%. Over the six conditions, the percent 

coverage ranged between 62% and 88%. 
C-4

Q4. What is the ratio of the rate for men under 45 years old to the rate for men 45 - 64 years old for each 

of the conditions listed? (N = 16) 

Table C.5 Summary of Respondent Results for Question Q4 

 (Rate for under 45 ) / (Rate for ages 45-64) 

Condition CV < Coverage Int. LQ, Median, UQ CV > Coverage Int. 

Arthritis (0.13) N = 2 0.10, 0.20, 0.30 N = 2 

Cataracts (0.11) N = 1 0.05, 0.10, 0.20 N = 1 

Diabetes (0.10) N = 8 0.20, 0.30, 0.50 N = 0 

Hearing Loss (0.20) N = 1 0.10,.0.20, 0.30 N = 2 

Heart Disease (0.17) N = 2 0.12, 0.20, 0.30 N = 2 

Prostate Disease (0.054) N = 6 0.05, 0.10, 0.20 N = 1 

Five of the six IQIs covered the CV, but respondents tended to overestimate the CV. Four of the six 

medians were above the CV, and two were equal or almost equal to the CV. Of the 96 coverage 

intervals, 20 (21%) lay above the CV and 8 (8%) lay below. The average percent coverage of all 96 

intervals was 71%. Over the six conditions, the percent coverage ranged between 50% and 88%. 

C.3 Discussion

The results of the training exercise were consistent with several of the basic premises underlying the

elicitation structure and methodology. First, apart from Q1, the responses to the other three questions as 

a whole did not exhibit any systematic over- or under- estimation bias. Q2 had no systematic bias, Q3

tended to underestimate, and Q4 tended to overestimate the CVs. This result is consistent with the basic 

premise of the elicitation process, which is that the panel responses as a whole have no systematic bias 

(see Section 3.3). 

Second, the percent coverage of the (LB, UB) intervals were less than the nominal 90% for all four 

questions. Q1 had the highest percent coverage at 82%, perhaps because the question dealt with 

demographic data with which the respondents were relatively more familiar. Q3 and Q4 had the next 

highest percent coverage at about 71% each and Q2 had the lowest percent coverage at 61%. This result

is consistent with the rationale for the overconfidence adjustments made to the panelists’ uncertainty

intervals (see Section 5.6.2). 

Third, the two questions (Q3 and Q4) that asked about ratios of rates had higher percent coverage than

the question (Q2) that asked about absolute rates. This result is consistent with the rationale for the basic 

structure of the elicitation questions, which ask about relative rather than absolute LOCA frequencies

(see Section 3.8). 
APPENDIX D 

PIPING BASE CASE RESULTS OF 

BENGT LYDELL 

D-1

APPENDIX D 

PIPING BASE CASE RESULTS OF BENGT LYDELL

An Application of the Parametric AttributeInfluence Methodology to Determine Loss of 

Coolant Accident (LOCA) Frequency Distributions 

Report No. 2 to the NRC Expert Panel on

LOCA Frequency Distributions

Prepared for 

U.S. Nuclear Regulatory Commission 

Washington (DC) 

June 2004
D-2

ACKNOWLEDGEMENTS 

The work documented in this report was performed for the U.S. Nuclear Regulatory Commission under 

Subcontract No. 177115 (Battelle Memorial Institute). 

Mr. Karl N. Fleming (Technology Insights, Inc., San Diego, CA) provided constructive review comments on 

a draft of this report. Mr. Fleming also provided the Markov model solutions supporting the calculation of 

time-dependent LOCA frequencies. 
D-3

TABLE OF CONTENTS 

D.1.0 BACKGROUND.................................................................................................................................. D.9 

D.1.1 Objectives ...................................................................................................................................... D.9

D.1.2 Base Case Definition ..................................................................................................................... D.9 

D.1.3 BWR Base Case System Descriptions ........................................................................................ D.10 

D.1.3.1 Reactor Recirculation (RR) System...................................................................................... D.10 

D.1.3.2 Feedwater (FW) System ....................................................................................................... D.11 

D.1.4 PWR Base Case System Descriptions......................................................................................... D.11 

D.1.4.1 Reactor Coolant (RC) System (Plant A.a)............................................................................ D.11 

D.1.4.2 High Pressure Injection (HPI)/Normal Makeup (NMU) Line (Plant A.b) .......................... D.12 

D.1.5 Summary of Scope Limitations................................................................................................... D.12 

D.1.6 Technical Approach to LOCA Frequency Estimation................................................................ D.12 

D.1.7 Study Conventions....................................................................................................................... D.13 

D.1.8 Report Organization..................................................................................................................... D.14 

D.2.0 TECHNICAL APPROACH............................................................................................................... D.14 

D.2.1 Overview of Analysis Steps ........................................................................................................ D.14 

D.2.2 Sensitivity Analyses..................................................................................................................... D.15 

D.3.0 SERVICE EXPERIENCE DATA APPLICABLE TO THE BASE CASE STUDY ....................... D.15 

D.3.1 PIPExp Database, Revision 2003.1............................................................................................. D.15 

D.3.2 Review of BWR-Specific Piping Service Experience ................................................................ D.16 

D.3.2.1 RR Piping Service Experience.............................................................................................. D.16 

D.3.2.2 FW Piping Service Experience............................................................................................. D.22 

D.3.3 Review of PWR-Specific Piping Service Experience................................................................. D.23 

D.3.3.1 RC & HPI/NMU Piping Service Experience ....................................................................... D.23 

D.4.0 DATA PROCESSING & DATA REDUCTION............................................................................... D.26 

D.4.1 Strategy for Data Processing & Data Reduction............................................................................. D.27 

D.4.1.1 Informative vs. Noninformative Prior Distributions............................................................. D.28 

D.4.2 BWR-Specific Apriori Pipe Failure Rates .................................................................................. D.29 

D.4.2.1 RR Pipe Failure Data ............................................................................................................ D.29 

D.4.2.2 FW Pipe Failure Data............................................................................................................ D.32 

D.4.3 PWR-Specific Apriori Pipe Failure Rates................................................................................... D.33 

D.4.4 Prior Distributions for Bayesian Updating.................................................................................. D.35 

D.5.0 DATA FOR LOCA FREQUNCY ESTIMATION............................................................................ D.37 

D.5.1 Posterior Weld Failure Rates....................................................................................................... D.37 

D.5.2 Conditional Weld Failure Probability ......................................................................................... D.40 

D.5.3 Conditional Failure Probability & Flow Rate ............................................................................. D.43 

D.6.0 BWR AND PWR BASE CASE LOCA FREQUNCIES................................................................... D.45 

D.6.1 LOCA Frequency Models ........................................................................................................... D.45 

D.6.2 BWR Base Case LOCA Frequencies at T = 25 .......................................................................... D.45 

D.6.3 PWR Base Case LOCA Frequencies at T = 25........................................................................... D.47 
D-4

D.6.4 Time-Dependency of LOCA Frequency Results........................................................................ D.48 

D.6.4.1 Use of Markov Model to Determine Time-Dependency...................................................... D.48 

D.6.4.2 Speculative LOCA Frequency at T = 40 & T = 60 .............................................................. D.59 

D.6.5 Influence of Service Data on LOCA Frequency......................................................................... D.60 

D.6.6 Service Data & Conditional Failure Probabilities....................................................................... D.61 

D.7.0 SUMMARY OF RESULTS............................................................................................................... D.61 

D.7.1 Discussion of Assumptions ......................................................................................................... D.61 

D.7.2 Summary of Input Data & Results .............................................................................................. D.62 

D.7.3 Benchmarking.............................................................................................................................. D.69 

D.7.4 Comparison to Historical LOCA Frequency Estimates.............................................................. D.70 

D.8.0 REFERENCES ................................................................................................................................... D.72 

ATTACHMENT A: SUMMARY OF PIPExp DATABASE ...................................................................... D.74 

ATTACHMENT B: BASIS FOR LOCA FREQUENCY MODELS.......................................................... D.83 

ATTACHMENT C: BASIS FOR LOCA FREQUENCY MODELS.......................................................... D.96 

ATTACHMENT D: SIGNIFICANT FAILURES OF SAFETY RELATED PIPING ............................. D.101 
D-5

LIST OF TABLES 

Table D.1 LOCA Size Classification Threshold Values............................................................................... D.13 

Table D.2 Estimated Flow Rates from Restrained Double-ended Guillotine Break (DEGB) ..................... D.13 

Table D.3 Selected Weld Counts in RR Piping............................................................................................. D.30 

Table D.4 Number of Through-wall Flaws in RR Piping Attributed to IGSCC .......................................... D.30 

Table D.5 IGSCC Susceptibility by Weld Configuration – Selected Parameter Values.............................. D.32 

Table D.6 Selected Weld Counts in ASME XI Class 1 FW Piping.............................................................. D.33 

Table D.7 Summary of FW Pipe Failure Data .............................................................................................. D.33 

Table D.8 Selected Weld Counts in Code Class 1 PWR Piping................................................................... D.34 

Table D.9 Degradation Susceptibility by Weld Configuration ..................................................................... D.34 

Table D.10 Prior RR and FW Weld Failure Rates........................................................................................ D.36 

Table D.11 Prior RC and HPI/NMU Weld Failure Rates............................................................................. D.37 

Table D.12 Posterior RR and FW Weld Failure Rate Distributions – BWR Base Cases............................. D.38 

Table D.13 Posterior RC and HPI/NMU Weld Failure Rate Distributions – PWR Base Cases.................. D.38 

Table D.14 Prior Beta Parameters (Constrained Noninformative Prior) ...................................................... D.42 

Table D.15 Probability of LOCA Given Severe Overloading ...................................................................... D.44 

Table D.16 Plant B LOCA Frequencies at T = 25 ........................................................................................ D.46 

Table D.17 Plant A.a/A.b LOCA Frequencies at T = 25 .............................................................................. D.48 

Table D.18 Four-State Markov Model Root Input Parameters..................................................................... D.50 

Table D.19 Inspection Cases Evaluated for Selected Pipe Segments........................................................... D.51 

Table D.20 Base Case LOCA Frequency Results (T = 25, 40 & 60 Years)................................................. D.60 

Table D.21 BWR LOCA Frequency Sensitivity Analysis Results............................................................... D.61 

Table D.22 Summary of Key BWR Base Case Input Data........................................................................... D.62 

Table D.23 Summary of Key PWR Base Case Input Data........................................................................... D.63 

Table D.24 Calculated LOCA Frequencies (T = 25 Years).......................................................................... D.63 

Table D.25 Calculated LOCA Frequencies (T = 40 Years).......................................................................... D.63 

Table D.26 Calculated LOCA Frequencies (T = 60 Years).......................................................................... D.63 

Table D.27 BWR & PWR LOCA Frequency Elicitation Anchor (MV) Values (T = 25 Years)................. D.64

Table D.28 Benchmarking of WinPRAISE vs. Service Experience............................................................. D.69 
D-6

LIST OF FIGURES 

Figure D.1 Weld Failures Due to IGSCC in Code Class 1 & 2 Piping (1970-2002) .................................. D.17 

Figure D.2 IGSCC Experience by Year(s) of Operation.............................................................................. D.18 

Figure D.3 IGSCC Data by Failure Mode.................................................................................................... D.18 

Figure D.4 IGSCC Experience by Pipe Size & Material Type.................................................................... D.19 

Figure D.5 Crack Depth versus Crack Length in Austenitic Stainless Steel ............................................... D.19 

Figure D.6 Crack Depth versus Crack Length in Austenitic Stainless Steel (NPS4, NPS12)..................... D.20 

Figure D.7 Crack Depth versus Crack Length in Austenitic Stainless Steel (NPS22, NPS28)................... D.20 

Figure D.8 Summary of IGSCC non-Through Wall Flaws by ‘a/t-Ratio’................................................... D.21 

Figure D.9 Service Experience with FW Piping (i)...................................................................................... D.22 

Figure D.10 Service Experience with FW Piping (ii) .................................................................................. D.23 

Figure D.11 Weld Failures in PWR RC-, CV- and SI-Piping (1970-2002) ................................................ D.24 

Figure D.12 Weld Failures in PWR RC-, CV- and SI-Piping (1970-2002) ................................................ D.25 

Figure D.13 Weld Failures in Small-Diameter PWR Piping ....................................................................... D.25 

Figure D.14 Pipe Failures Attributed to Thermal Fatigue in PWRs Worldwide......................................... D.26 

Figure D.15 Four-State Markov Model of Piping Reliability...................................................................... D.27 

Figure D.16 Rate of IGSCC-Induced RR Pipe Failure (‘Prior State-of-Knowledge’)................................ D.30 

Figure D.17 RR Weld Failures as a Function Weld Configuration ............................................................. D.31 

Figure D.18 Prior Frequency of Non-Through Wall IGSCC in RR Piping................................................. D.35 

Figure D.19 Posterior IGSCC Frequency (non-Through Wall)................................................................... D.39 

Figure D.20 Prior & Posterior IGSCC Frequency (non-Through Wall) for NPS12 Welds........................ D.39 

Figure D.21 Prior & Posterior IGSCC Frequency (non-Through Wall) for NPS22 Welds........................ D.39 

Figure D.22 Prior & Posterior IGSCC Frequency (non-Through Wall) for NPS28 Welds........................ D.40 

Figure D.23 Likelihood of Structural Failure According to Service Experience with 

Light Water Reactor Piping System..................................................................................................... D.41 

Figure D.24 Event Tree for Definition of LOCA Categories....................................................................... D.43 

Figure D.25 Conditional Probability of Weld Failure Given Through-wall Flaw & Severe Overloading . D.44

Figure D.26 Time-Dependent BWR-1 Cat1 LOCA Frequency Given ‘Cause-Based’ ISI......................... D.52 

Figure D.27 Time-Dependent BWR-1 Cat1 LOCA Frequency Assuming no ISI...................................... D.52 

Figure D.28 Time-Dependent BWR-1 Cat1 LOCA Frequency Given ‘Comprehensive ISI’..................... D.53 

Figure D.29 Time-Dependent BWR-2 Cat1 LOCA Frequency Given ‘Cause-Based’ ISI......................... D.53 

Figure D.30 Time-Dependent BWR-2 Cat1 LOCA Frequency Assuming no ISI...................................... D.54 

Figure D.31 Time-Dependent BWR-2 Cat1 LOCA Frequency Given ‘Comprehensive ISI’..................... D.54 

Figure D.32 Time-Dependent PWR-1 Cat1 LOCA Frequency Given ‘Cause-Based’ ISI ......................... D.55 

Figure D.33 Time-Dependent PWR-1 Cat1 LOCA Frequency Assuming no ISI....................................... D.55 

Figure D.34 Time-Dependent PWR-1 Cat1 LOCA Frequency Given ‘Comprehensive ISI’..................... D.56 

Figure D.35 Time-Dependent PWR-2 Cat1 LOCA Frequency Given ‘Cause-Based’ ISI ......................... D.56 

Figure D.36 Time-Dependent PWR-2 Cat1 LOCA Frequency Assuming no ISI....................................... D.57 

Figure D.37 Time-Dependent PWR-2 Cat1 LOCA Frequency Given ‘Comprehensive ISI’..................... D.57 

Figure D.38 Time-Dependent PWR-3 Cat1 LOCA Frequency Given ‘Cause-Based’ ISI ......................... D.58 

Figure D.39 Time-Dependent PWR-3 Cat1 LOCA Frequency Assuming no ISI....................................... D.58 

Figure D.40 Time-Dependent PWR-3 Cat1 LOCA Frequency Given ‘Comprehensive ISI’..................... D.59 

Figure D.41 Time-Dependent Cat1 LOCA Frequencies.............................................................................. D.64 

Figure D.42 Selected Base Case Weld Failure Rates................................................................................... D.65 

Figure D.43 BWR-1 LOCA Frequency ....................................................................................................... D.65 

Figure D.44 BWR-2 LOCA Frequency (Feedwater System Loop A & B)................................................. D.66 

Figure D.45 PWR-1 and PWR-2 LOCA Frequency.................................................................................... D.66 

Figure D.46 PWR-3 LOCA Frequency (ASME Code Class 1 HPI/NMU System).................................... D.67 

Figure D.47 Selected BWR Base Case Sensitivity Analysis Results – Cat1 LOCA................................... D.67 
D-7

Figure D.48 Influence of ISI on Time-Dependent BWR-1 Cat1 LOCA Frequency................................... D.68 

Figure D.49 Influence of ISI on Time-Dependent PWR-1 Cat1 LOCA Frequency ................................... D.68 

Figure D.50 Predicted (WinPRAISE) vs. Observed Weld Failure Rate (PIPExp)...................................... D.69 

Figure D.51 Comparison of Selected BWR Large LOCA Frequency Estimates........................................ D.71 

Figure D.52 Comparison of Selected PWR Large LOCA Frequency Estimates ........................................ D.71 
D-8

ABBREVIATIONS

ASME American Society of Mechanical 

Engineers 

ASTM American Society for Testing 

and Materials 

CL Cold Leg (of a PWR RCS) 

CV Chemical and Volume Control 

DN Nominal Pipe Size, metric [mm] 

DPD Discrete Probability Distribution 

EPRI Electric Power Research Institute 

FW Feedwater 

HAZ Heat Affected Zone (of weld) 

HL Hot Leg (of a PWR RCS) 

HPI High Pressure Injection 

HWC Hydrogen Water Chemistry 

IGSCC Intergranular Stress Corrosion 

Cracking 

IHSI Induction Heat Stress 

Improvement 

ISI Inservice Inspection 

LOCA Loss of Coolant Accident 

MSI Mechanical Stress Improvement 

MV Mid Value (50% percentile) 

NDE Nondestructive Examination 

NMU Normal Makeup 

NWC Normal Water Chemistry 

NPS Nominal Pipe Size, US [inch] 

PBF Pressure Boundary Failure 

PWSCC Primary Water Stress Corrosion 

Cracking 

RCPB Reactor Coolant Pressure 

Boundary 

RCS Reactor Coolant System 

RR Reactor Recirculation 

SC Sensitivity Case 

SI Safety Injection 

SS Stainless Steel 

TS Technical Specifications1

 

1

 For TS leak rate limits see for example NUREG-1431 (Vol 2, Rev. 

2, June 2001): Standard Technical Specifications Westinghouse Plants 

– Bases, Section B 3.4.13, RCS Operational Leakage. For unidentified 

leakage the Limiting Condition for Operation (LCO) is 3.8 lpm (1 

gpm), and for identified leakage the LCO is 38 lpm (10 gpm). See 

also NUREG-1433 (Vol. 2, Rev. 2, June 2001): Standard Technical 

Specifications General Electric Plants, BWR/4 – Bases, Section B 

3.4.4, RCS Operational Leakage. For unidentified leakage the LCO is 

19 lpm (5 gpm). Further, if an unidentified (BWR) leakage has been 

identified and quantified, it may be reclassified and considered as 

identified leakage; however, the total leakage limit would remain 

unchanged. 

UT Ultrasonic Testing 

Notation

A Attribute 

a/t Ratio of crack depth to pipe wall 

thickness 

C Conditional failure probability 

given a flawed weld and an 

unusual or severe loading 

condition 

F Failure (= through-wall flaw) 

L Large leak 

P Probability 

S Susceptibility (to degradation) 

W Weld count 

∅ Pipe diameter 

φ Crack occurrence rate 

λ Failure rate (frequency of PBF 

resulting in leak rate ≤ TS limit 

for unidentified leakage) 

ρ Rate of large leak event 

ν Leak/spill rate (gpm) 

σNO Normal operating weld stress 

(ksi) 

ω Repair rate 
D-9

D.1 Background

Limited to consideration of Code Class 1 piping failures, Base Case Report Number 2 documents an 

assessment of BWR and PWR loss of coolant accident (LOCA) frequency distributions. The assessment is a

demonstration of the role of statistical analysis of service experience data and Markov modeling in a “bottomup” approach to piping system reliability analysis.

D.1.1 Objectives 

Using primary coolant piping design information for three reference plants (one BWR plant and two PWR 

plants), the overall objective is to determine LOCA frequency distributions that are representative of currently 

operating U.S. nuclear power plants, including current in-service inspection (ISI) practices and degradation 

mitigation strategies. This determination is done analytically using a parametric model of piping reliability. 

The LOCA frequency distributions are determined for three time periods. To address today’s piping 

reliability state-of-knowledge the LOCA frequency is determined at T = 25 years. Next the LOCA frequency 

is extrapolated to T = 40 years to represent the primary system piping reliability status at the end of a 40-year 

operating license. Finally an extrapolation is made to T = 60 years to account for a possible license renewal. 

Analytically, this extrapolation is concerned with the potential impact on the structural integrity of the piping 

by material aging as well as by reliability improvement efforts. 

As implied by the report title, the objective is to develop LOCA frequency distributions. The report addresses 

two aspects of LOCA frequency distributions. It develops LOCA frequencies associated with a distribution of 

flow rate threshold values ranging from 380 lpm (100 gpm) at the low end to beyond 380,000 lpm (100,000 

gpm) at the high end. Additionally the study develops statistical uncertainty distributions for each set of 

LOCA frequencies to account for the uncertainty in the input parameters to this piping reliability analysis. 

D.1.2 Base Case Definition

During a meeting in Rockville (MD) in February 2003 [D.1], the Expert Elicitation Panel members defined 

five Base Cases that are denoted as BWR-1, BWR-2, PWR-1, PWR-2 and PWR-3, respectively. The five 

Base Cases are: 

BWR Base Case (Plant ‘B’)

• BWR-1; Reactor Recirculation (RR) System. This reference case includes one-of-two RR System 

loops. Each loop consists of one NPS28 recirculation pump loop with a NPS22 manifold with five 

NPS12 risers; NPS is nominal pipe size in inch. The reference case excludes any small-diameter 

piping or tubing attached to the main RR piping. With a few exceptions, the selected piping system 

layout is representative of a BWR/4 reference plant as described in NUREG/CR-6224 [D.2]. The 

Base Case RR System does not include the NPS4 bypass line, however. The RR piping is fabricated 

from austenitic Cr-Ni stainless steel of Type A-304 (≥ 0.035% carbon). 

• BWR-2; Feedwater (FW) System. As defined by isometric drawings, this reference case includes 

Loop B of the Class 1 portion of the FW System (i.e., the part of the FW System that is located in the

drywell containment structure). This system of two loops includes NPS12, NPS14 and NPS20 

piping. The FW piping is fabricated from carbon steel of Type A-333 Gr. 6. 

• Section D.1.3 includes additional information on the BWR Base Case system definitions. 

PWR Base Case (Plant ‘A.a/b’)

• PWR-1; Reactor Coolant (RC) System. As defined by an isometric drawing, this reference case 

includes one of the NPS30 hot leg (HL) in the RCS. 
D-10

• PWR-2; Pressurizer Surge Line. As defined by an isometric drawing, this reference case includes the 

NPS14 piping, which connects the pressurizer to the cold leg (CL). 

• PWR-3; High Pressure Injection/Normal Makeup (HPI/NMU) System. As defined by an isometric 

drawing, this reference case includes the 2-1⁄2 inch schedule 160 line between the containment 

isolation valve and the RCS cold leg (CL). 

• The PWR base cases associated with the RC hot leg and pressurizer surge line are typical of a 3-loop 

Westinghouse PWR (Plant A.a). The PWR base case associated with the HPI/NMU line is typical of 

a Babcock & Wilcox PWR (Plant A.b). 

• Section D.1.4 includes additional information on the PWR Base Case system definitions. 

D.1.3 BWR Base Case System Descriptions 

Plant B is a BWR/4 assumed to have been in commercial operation for at least 25 years. Similar to many

other operating BWR/4 plants in the USA, Plant B is also assumed to be operating with a combination of

IGSCC Category D and E welds, according to the nomenclature of U.S. NRC Generic Letter 88-01 [D.3, 

D.4]. In other words, the plant has experienced some IGSCC and the affected welds have been reinforced by 

weld overlays. It is further assumed that none of the IGSCC susceptible welds have been subjected to any 

stress improvement (SI) process such as induction heat stress improvement (IHSI) or mechanical stress 

improvement process (MSIP). It is also assumed that the weld overlay repairs (WOR) were all performed in 

the 1982-1988 timeframe. Finally, Plant B is assumed to have been operating with normal water chemistry 

(NWC) at all time. 

The system descriptions in this section are extracted from design information supplied by members of the 

Expert Elicitation Panel. The BWR-specific system information is included in the following documents and 

drawings: 

• Document No. EPRI-156-310: Degradation Mechanisms Evaluation for Class 1 Piping Welds at 

Plant B [D.5]. 

• Excel-file entitled “PlantBWelds.” This Excel-file includes weld lists with locations for the RR and 

FW ASME Section XI Code Class 1 piping. The lists are organized by weld identification numbers 

(as they appear on the isometric drawings identified below) nominal pipe size and pipe schedule. The 

Excel file forms the basis for the LOCA frequency model used to derive the LOCA frequency 

distributions. 

• Isometric drawing numbers 6M721-5358-5 (RR System Loop B Ring Header), 6M721-5359-5 (RR 

Loop B Suction & Discharge Piping), 6M721-2336-1 (FW System Inside Drywell), and 6M721-

3537-5 (FW System Inside Drywell). 

D.1.3.1 Reactor Recirculation (RR) System - The RR System evaluated in this study consists of two 

recirculation pump loops external to the reactor pressure vessel (RPV). These loops provide the piping path 

for the driving flow of water to the RPV jet pumps. Each loop contains a variable speed recirculation pump 

and two motor operated isolation valves (one on each side of each pump). The recirculation loops are part of 

the nuclear system process barrier and are located inside the drywell containment structure. The pipe 

segments that are subject to evaluation in this study consist of: 

Loop A: The Class 1 portion starts at the RPV nozzle N1A and is reconnected to the RPV at nozzles 

N2F, N2G, N2H, N2J, and N2K. Class 1 lines for the Residual Heat Removal (RHR) and 

Reactor Water Cleanup (RWCU) Systems are connected to this loop. These particular Class 

1 lines are excluded from the study scope, however. Loop A is excluded from the BWR 

Base Case. 
D-11

Loop B: The Class 1 portion starts at RPV nozzle N1B and is reconnected to the RPV at nozzles 

N2A, N2B, N2C, N2D, and N2E. Part the original design, a NPS4 bypass line at valve 

F031B has been removed from the system. Class 1 lines for the RHR and RWCU Systems 

are connected to this loop. These particular Class 1 lines are excluded from the study scope, 

however. 

D.1.3.2 Feedwater (FW) System - The FW System provides feedwater to maintain a pre-established 

water level in the RPV during normal plant operation. The Condensate and the FW Systems take water from 

the main condenser and deliver it to the RPV after passing it through the feedwater heaters and demineralizer 

system. The Class 1 portion of the FW System consists of two loops:

Loop A: Loop A starts at valve F076A and a connection to the High Pressure Coolant Injection 

(HPCI) discharge line (at valve F006), and connects to the RPV at nozzles N4A, N4B, and 

N4C. The HPCI discharge line is excluded from the study scope. Loop A is excluded from 

the BWR Base Case. 

Loop B: Loop B starts at valve F076B, connection to the Reactor Core Isolation Cooling (RCIC) 

discharge line at valve F013, and a discharge from the RWCU System (at valve F220), and 

connects to the RPV at nozzles N4D, N4E, and N4F. The RCIC and RWCU discharge lines 

are excluded from the study scope. 

D.1.4 PWR Base Case System Descriptions 

The system descriptions in this section are extracted from design information supplied by members of the 

Expert Elicitation Panel. The PWR-specific system information is included in the following documents and 

drawings: 

• Document No. EPRI-156-330: Degradation Mechanism Evaluation for Class 1 Piping Welds at Plant 

A.a [D.6]. This document summarizes the degradation mechanisms applicable to a Westinghouse 3-

loop PWR. 

• References [D.7-D.9] include design information as well as degradation mechanism information 

applicable to the HPI/NMU system of Plant A.b. 

• Excel file entitled “PlantAWelds.” This Excel file includes weld lists for the RC system of Plant A.a.

The lists are organized by weld identification numbers (as they appear on the isometric drawings 

identified below), nominal pipe size and pipe schedule. This Excel-file forms one of the bases for the

PWR LOCA frequency model used to derive the LOCA frequency distributions. 

• Isometric drawing numbers 1MS-22-2262 and CGE-1-4100A (RC Hot Leg), C-314-601 and CGE-1-

4500A (pressurizer surge line), and 17-MU-23 (HPI/NMU piping). 

D.1.4.1 Reactor Coolant (RC) System (Plant A.a) - The RC System evaluated in this study consists of 

three similar heat transfer loops connected in parallel to the RPV. Each loop contains a reactor coolant pump 

(RCP), steam generator, and associated piping and valves. In addition, the system includes a pressurizer, a 

pressurizer relief tank, interconnecting piping, and instrumentation necessary for operational control. The 

analysis in this report is concerned with a portion of one of the three RC loops; the portion from the RCP to 

the RPV (this is one of the hot legs). The pressurizer surge line connects the pressurizer to the RC cold leg 

Loop A. In summary, the piping sections that are subject to evaluation in this study consist of. 
D-12

RC-HL: The analysis is concerned with 1-of-3 hot legs. The Loop A HL starts at the RPV, includes 

an RCP and connects to the ‘A’ steam generator (S/G). The HL piping is fabricated from 

stainless steel piping. The section of the HL from the RPV to the RCP is of 31 inch inside 

diameter, while the section from the RCP to S/G is of 27.5 inch inside diameter piping. 

Surge Line: The single surge line is fabricated form NPS14 stainless steel piping and connects to the 29-

inch RC cold leg. 

D.1.4.2 High Pressure Injection (HPI)/Normal Makeup (NMU) Line (Plant A.b) - In Plant A.b, each of 

the four RCS cold legs is equipped with high-pressure injection piping. Two of these 2 1⁄2 inch (ID, or 

approximately NPS3-3⁄4) stainless steel piping lines provide the normal makeup flow to the RCS and they 

connect to the cold leg via nozzle assemblies. Each of the nozzles is comprised of a base nozzle and a safeend. To prevent thermal cycling of the base metal each nozzle is equipped with a 1.5-inch thermal sleeve. The 

analysis is concerned with one of the two HPI/NMU lines. 

D.1.5 Summary of Scope Limitations 

As outlined above, this LOCA frequency assessment is limited to specific portions of BWR and PWR Code 

Class 1 piping. The BWR base cases include contributions from potential pipe breaks in Loop B of the 

respective RR and FW System. Pipe break frequency contributions from normally pressurized sections of 

HPCI, RCIS, RHR or RWCU piping are not considered in this study. Piping system design information 

beyond that itemized above is not accounted for in this study. The PWR base cases include contributions 

from 3-of-3 RC hot legs and 2-of-2 HPI/NMU lines, respectively. 

Excluded from the analysis are LOCA frequency contributions due to degradation and failure of cast stainless 

steel components such as valve bodies. While there is some documented evidence of degradation of such 

components, (e.g., [D.10]) the frequency of a through-wall defect in valve bodies and pump casings is viewed 

as being considerably lower than for welds in Class 1 systems. 

D.1.6 Technical Approach to LOCA Frequency Estimation 

Existing service experience with piping systems shows a strong correlation between failures and presence of 

an active degradation mechanism in combination with service conditions and transient loading conditions. It 

is therefore possible to estimate piping reliability parameters through statistical analysis of service experience 

data. Such analysis includes data processing whereby the appropriate reliability attributes are correlated with 

influence factors as described in SKI Report 97:26 [D.11]. 

In this Base Case Report the technical approach to LOCA frequency estimation builds on statistical analysis 

of service data associated with ASME XI Class 1 piping in the BWR and PWR operating environments. The 

study accounts for two kinds of uncertainties in piping reliability analysis, namely data uncertainty and stateof-knowledge uncertainty. The pipe failure database on which this study is based is called PIPExp [D.12], 

which is the extended version of the OPDE pipe failure database [D.13]. A description of PIPExp is included 

in Appendix A. The uncertainty analysis is performed by using a Monte Carlo merge technique to develop

the LOCA frequency distributions. A commercial software package called Crystal Ball (Version 2000.2.2), 

which is an add-on for Microsoft Excel, is used to perform this Monte Carlo merge operation. Timedependent LOCA frequencies are developed using a Markov modeling approach [D.14]. 

The BWR Base Case analysis is based on the degradation mechanism analysis as documented in Reference 

[D.5], and it builds on insights from an earlier BWR LOCA frequency pilot study [D.15-D.16]. The PWR 

Base Case analysis is based on the degradation mechanism analysis as documented in References [D.6, D.8], 

and builds on insights and results from an earlier sensitivity analysis performed in support of a risk informed 
D-13

inservice inspection (RI-ISI) evaluation [D.12]. That sensitivity analysis addressed the impact of using a 

different pipe failure database on the RI-ISI weld selection. 

D.1.7 Study Conventions 

Throughout this report, pipe sizes are referenced by nominal pipe size (NPS), which indicates standard pipe 

size without an inch symbol. The smallest pipe size considered in this study is NPS3-3⁄4 (Plant A.b). All 

references to specific material types are made according to designations by the American Society for Testing 

and Materials (ASTM). The term “weld failure” is used to indicate a rejectable (non-through-wall or throughwall) flaw. 

During the NRC LOCA Elicitation Kick-off Meeting [D.1], a LOCA was defined as “a breach of the reactor

coolant pressure boundary which results in a leak rate greater than 380 lpm (100 gpm).” Instead of using the 

traditional (or historical) LOCA size classes (small – medium – large) that are based on break size, this study 

uses LOCA sizes that are based on leak rate threshold values as indicated in Table D.1 (adapted from [D.1]) 

and Table D.2. 

Table D.1 LOCA Size Classification Threshold Values 

LOCA 

Category 

Flow Rate (ν) Thresholds 

gpm (lpm) Comment 

0 ν > 10 (38) 

Cat0 corresponds to a pressure boundary failure (breach) resulting 

in a leakage exceeding the T.S. limit for identified leakage. 

1 ν > 100 (380) 

Breach in piping of up to 1.8-inch diameter (BWR), and 1.7-inch 

diameter (PWR); see Table D.3. 

2 ν > 1,500 (5,700) Breach in piping of up to 3.3-inch diameter (BWR), and 3-inch 

diameter (PWR) 

3 ν > 5,000 (19,000) Breach in piping of up to 7.3-inch diameter (BWR), and 6.8-inch 

diameter (PWR) 

4 ν > 25,000 (95,000) Breach in piping of up to 18.4-inch diameter (BWR), and 14-inch 

diameter (PWR) 

5 ν > 100,000 (380,000) 

Breach in NPS28 RR piping (BWR) yields on the order of 230,000 

gpm. Breach in RCS hot leg piping of up to 31-inch diameter. 

6 ν > 500,000 (1,900,000) Applies to PWR RCS-HL base case only, and only for a relatively 

short time following a postulated DEGB 

Table D.2 Estimated Flow Rates from Restrained Double-Ended Guillotine Break (DEGB)2

Pipe Size Restrained DEGB (Plant A – PWR) Restrained DEGB (Plant B – BWR)

[NPS] Break Size 

[sq. in.]

Press. 

[psig]

Max. Flow 

Rate [gpm]

Break Size 

[sq. in.]

Press. 

[psig]

Max. Flow 

Rate [gpm]

1 .41 2250 540 .41 1250 467 

2 1.65 2250 2158 1.65 1250 1869 

4 6.60 2250 8633 6.60 1250 7476 

6 14.84 2250 19424 14.84 1250 16823 

8 26.39 2250 32280 26.39 1250 29908 

12 59.37 2250 72495 59.37 1250 42411 

14 80.81 2250 98624 80.81 1250 57698 

22 199.54 2250 243542 199.54 1250 142478 

28 323.22 2250 394497 323.22 1250 230790 

30 371.05 2250 452867 N/A -- -- 

 

2

 Technical basis for leak rate calculation is documented in an attachment to Minutes of Meeting (2nd Elicitation Meeting), Bethesda 

(MD), June 4-5, 2003. 
D-14

The estimation of weld failure rates uses Bayesian reliability analysis methodology, and involves the 

development of prior and posterior failure rate distribution. In this study the term ‘prior’ refers to piping 

reliability characteristics before the implementation of industry programs to mitigate or eliminate 

susceptibilities to certain degradation mechanisms. The term ‘posterior’ refers to observed or expected 

reliability characteristics after reliability improvement actions have been implemented. 

D.1.8 Report Organization 

This report consists of eight sections and four appendices. Section D.2 is an overview of the analysis steps. 

Section D.3 summarizes the service experience applicable to the BWR and PWR Base Cases, respectively. 

Using the PIPExp database, Section D.4 includes a summary of the data interpretation and data processing 

steps necessary to derive piping reliability parameters that apply to the base case definitions. Section D.5 

documents the results of the pipe failure rate estimation while Section D.6 is a documentation of the models 

used for estimating LOCA frequency, while Section D.7 is a summary of results. Section D.8 is a list of 

references. Note that the Base Case results used in Table E.1 in the main body can be obtained from Tables 

16, 17, and 20 in this report. 

Appendix A summarizes the PIPExp database structure. Appendix B includes the Excel spreadsheets that are 

used as the basis for the LOCA frequency models, and Appendix C includes the Excel spreadsheets for the 

calculation of time-dependent LOCA frequencies. Finally, Appendix D is a summary of selected, significant 

Code Class 1 and 2 pipe failures in commercial nuclear power plants worldwide. 

D.2 Technical Approach

Base Case Report 2 develops BWR and PWR LOCA frequency distributions using a ‘bottom-up approach.’ 

Statistical analysis of relevant service experience data is used to quantify the weld failure rate and rupture 

frequency of individual welds. Next the failure rate and rupture frequency (= LOCA frequency) for an entire 

system is calculated by concatenating the individual weld failure rates and rupture frequencies. Markov 

model theory is used to evaluate the influence of alternate strategies for in-service inspection and leak 

detection on the frequency of leaks and ruptures. 

D.2.1 Overview of Analysis Steps 

Different approaches have been applied to estimating pipe failure rates and rupture frequencies; from 

probabilistic fracture mechanics, via direct statistical estimation to expert judgment. The most straightforward 

approach is to obtain statistical estimates of piping component failure rates based on data collected from field 

experience. A variation of this approach is to augment statistical estimates of pipe failure parameters with 

simple correlations that express the problem in terms of a failure rate and a conditional probability for each 

failure mode of interest such as the approach used in NUREG/CR-5750 [D.17]. 

A limitation of the statistical analysis approach is that attempts to segregate the service data to isolate the 

impact of key design parameters and properties of various degradation and damage mechanisms often leads 

to subdividing a database into very sparse data sets. If not optimized properly, this approach may introduce 

large uncertainties in the failure rate estimates. In addition, historical data may reflect the influence of no 

longer relevant inspection programs. If changes to these programs have been implemented, such changes may 

render the failure rate estimates no longer relevant. In risk-informed applications, the failure data and analysis 

methods need to provide future predictions of piping system reliability that can account for changes in the 

inspection strategy or improvement in the NDE technology. 
D-15

An objective of the work documented in this report is to demonstrate the utility of a pipe failure data 

collection. Time-dependent LOCA frequencies are calculated by making full use of the PIPExp database in 

combination with Markov model theory [D.14]. The LOCA frequency calculation in this report is structured 

to support the Expert Elicitation and consists of four steps; each step is addressed in a separate report section: 

• Section D.3. The service experience that is applicable to the five bases cases is summarized in 

this section. The data summaries correspond to queries in the PIPExp database. 

• Section D.4. The approach to calculating time-dependent LOCA frequencies is presented. A 

Bayesian update process is used to derive failure parameters that reflect the attributes of 

respective base case definition. The results of this analysis step are in the form of generic weld 

failure rate distributions. These distributions represent the industry-wide service experience 

prior to the implementation of the specific pipe failure mitigation programs that are currently 

in place. 

• Section D.5. In this section current state-of-knowledge (or base case specific) weld failure rate 

distributions are develop. The chosen estimation approach includes a formal uncertainty 

analysis that accounts for uncertainty in the failure data and exposure data. Engineering 

judgment and insights from the review of service data are used to address the conditional 

probability of pipe failure given presence of through-wall flaws. 

• Section D.6. An Excel spreadsheet format is used to develop LOCA frequency models 

corresponding to each of the five base cases. These models generate LOCA frequency 

distributions at T = 25 years. A Markov model is used to investigate the time-dependency of 

LOCA frequencies. The output of this model consists of LOCA frequencies at T = 40 years 

and T = 60 years. 

D.2.2 Sensitivity Analyses 

Two types of sensitivity analysis are included in this report. The first type addresses the impact on results by 

an assumed incompleteness of the failure data collection. The second type relates to the sensitivity of the 

time-dependent LOCA frequencies to different assumptions about leak detection and in-service inspection. 

The sensitivity analysis results are included in Section D.6. 

D.3 Service Experience Data Application to the Base Case Study

The PIPExp database documents service experience with Code Class 1, 2 and 3 and non-safety related (or

Class 4) piping in commercial nuclear power plants worldwide. For the time period 1970-2002, this database 

was queried for service experience data specific to the Base Case piping systems. The results of the database 

queries are summarized here, and they form the input to the data processing and failure parameter estimation 

in Sections D.4 and D.5. 

D.3.1 PIPExp Database, Revision 2003.1 

The pipe failure database utilized in the Base Case Study is called PIPExp. It is an ACCESS database and an 

extension of the OPDE database [D.12-D.13]. Since the conclusion of the original work in 1998 [D.11, 

D.17], the pipe failure database has been significantly expanded both in terms of the absolute number of 

event records and the depth of the database structure (Appendix A provides additional details). Lessons 

learned through database applications have been used to enhance the structure. In this study of HPI//NMU-, 

FW-, RC- and RR-piping reliability the statistical analysis is based on service data as recorded in PIPExp and 

with cutoff date of December 31, 2002. The analysis is inclusive of applicable worldwide BWR- and PWRspecific service experience with Code Class 1 piping. As of 12-31-2002 the database accounted for 
D-16

approximately 1,992 and 3,621 critical reactor-years of operating experience with commercial BWR and 

PWR plants, respectively. 

The database is actively maintained and periodically updated. The effort involved in populating the database 

while at the same time assuring data quality is not trivial. As an example, changing regulatory reporting 

thresholds imply that an ever increasing volume of raw data reside in restricted and proprietary database 

systems rather than in the public domain. For an event to be considered for inclusion in the database it 

undergoes screening for eligibility. For example: 

• The equipment failure must be positively identified as a piping component failure external to the 

reactor pressure vessel (RPV). A failure involves a pressure boundary degradation, which can be 

non-through-wall (crack with a/t-ratio ≥ 10%, where a = crack depth and t = wall thickness) or a 

through-wall leak. 

• There must exist documented evidence in the form of a hard copy (e.g., USNRC Inspection Report, 

Licensee Event Report, ISI Summary Report, Problem Identification Form, Condition Report, 

ASME Code Repair Relief Request, etc.) from which a sufficiently detailed case history is 

developed. The documented evidence of pipe degradation/failure must contain information on its 

location within a piping system (e.g., with reference to an isometric drawing and/or P&ID), 

metallurgy, operating conditions, impact on operation, method of discovery, failure history, etc. so 

that a data classification may be independently verified. 

• Where the documented evidence is deemed incomplete, additional information is solicited through 

direct contact with plant personnel or by accessing supplemental data. 

• There must be sufficient technical information available to fully address the complex relationships 

between piping reliability attributes (or design parameters) and influence factors (e.g., 

fabrication/welding techniques, environmental conditions such as water chemistry, flow conditions) 

on the one hand and degradation/failure mechanisms on the other. 

• Differentiation between UT indications versus confirmed crack indications. Only the latter are 

included in the database given an a/t-ratio ≥ 10%. 

Following on the initial data screening, each event selected for inclusion in the database is subjected to a 

classification so that the unique reliability attributes and influence factors are identified. Including memo 

fields, text fields, numerical fields and data filters, up to 114 database fields describe each record of the 

database. 

D.3.2 Review of BWR-Specific Piping Service Experience 

Limited to the BWR Base Case systems, this section summarizes the worldwide service experience with 

Code Class 1 piping. The results of this review are input to the pipe failure rate estimation. 

D.3.2.1 RR Piping Service Experience - The original piping material in BWR plants commissioned prior 

to mid-1980 is austenitic stainless steels that contain more than 0.03% carbon. During welding these steels 

are susceptible to sensitization that results in a loss of corrosion resistance. Intergranular stress corrosion 

cracking (IGSCC) occurs when the sensitized steel is subjected to stresses and corrosive environment. 

Sensitization can be avoided by controlling the carbon content to below 0.03%. Another approach to 

controlling sensitization is to add strong carbide formers such as titanium or niobium to the steel. Stainless 

steels with additions of titanium or niobium are called “stabilized.” It is noted that low-carbon content 

unstabilized stainless steel or stabilized stainless steels are not completely immune to IGSCC, however 

[D.18]. 
D-17

For Plant B, IGSCC is the predominant degradation mechanism acting on the RR piping welds, including 

heat-affected zones. During early plant life some weld reinforcements were performed where the inservice 

inspection revealed presence of surface penetrating, and subsurface cracking due to IGSCC. Since the 

analysis of LOCA frequency distributions is based on a degradation mechanism evaluation, the PIPExp 

database is queried for service data including IGSCC. The database queries are summarized in a set of charts 

and tables below. The database currently includes a total of about 1000 records on IGSCC in BWR piping. 

Figure D. 1 shows the number of weld failures due to IGSCC by calendar year and Figure D. 2 shows the 

number of weld failures by year of operation. Here a weld heat affected zone with a/t ≥ 10% is characterized 

as a “weld failure.” 

0

50

100

150

200

250

1970

1972

1974

1976

19781980

1982

19841986

1988

1990

1992

1994

1996

19982000

2002

No. db Records

BWR Worldw ide (Stabilized & Unstabilized SS)

BWR US

BWR (Unstabilized SS Only)

Figure D.1 Weld Failures Due to IGSCC in Code Class 1 & 2 Piping (1970-2002) 
D-18

0

20

40

60

80

100

120

140

160

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30

Plant Age [Years]

No. db Records

BWR Worldw ide (Stabilized & Unstabilized SS)

BWR US Plants

BWR Worldw ide (Unstabilized SS)

Figure D.2 IGSCC Experience by Year(s) of Operation 

In Figure D. 3 the IGSCC data is organized by mode of failure (crack – pinhole leak – leak) and pipe size. 

Figure D. 4 shows the IGSCC data by size and material type. 

86.56%

94.12%

89.36% 88.86%

13.44%

5.88%

10.64% 11.14%

0.00%

10.00%

20.00%

30.00%

40.00%

50.00%

60.00%

70.00%

80.00%

90.00%

100.00%

NPS12 NPS22 NPS28 All

% Crack

% Through Wall

Figure D.3 IGSCC Data by Failure Mode 
D-19

104

212

219

269

63

84

203

237

88

161

216

269

0 50 100 150 200 250 300

4 < ∅ ≤ 6

6 < ∅ ≤ 10

10 < ∅ ≤ 16

∅ > 16

Nominal Pipe Size (NPS)

Number of db Records

Worldwide (Unstabilized SS)

US Plants

Worldwide (Stabilized & Unstabilized SS)

Figure D.4 IGSCC Experience by Pipe Size and Material Type 

Figures D. 5 through D.7 include plots of crack depth versus crack length (L) in the circumferential (C) 

direction. An ‘a/t-ratio’ of 100% indicates a crack, which has penetrated the outside pipe wall. An ‘L/C-ratio’ 

of 100% indicates a crack, which spans the entire inside pipe circumference. Limited to part through-wall 

cracks, Figure D.8 summarizes the data by a/t-ratio. 

0.00%

10.00%

20.00%

30.00%

40.00%

50.00%

60.00%

70.00%

80.00%

90.00%

100.00%

0.0% 10.0% 20.0% 30.0% 40.0% 50.0% 60.0% 70.0% 80.0% 90.0% 100.0

%

L/C-ratio (RR Bypass, RR Riser, RR Manifold, RR Loop)

a/t-ratio

Figure D.5 Crack Depth Versus Crack Length in Austenitic Stainless Steel 
D-20

0.00%

10.00%

20.00%

30.00%

40.00%

50.00%

60.00%

70.00%

80.00%

90.00%

100.00%

0.0% 10.0% 20.0% 30.0% 40.0% 50.0% 60.0% 70.0% 80.0% 90.0% 100.0

%

L/C-ratio (RR Bypass, RR Riser)

a/t-ratio

Figure D.6 Crack Depth Versus Crack Length in Austenitic Stainless Steel (NPS4, NPS12) 

0.00%

20.00%

40.00%

60.00%

80.00%

100.00%

120.00%

0.0% 10.0% 20.0% 30.0% 40.0% 50.0% 60.0% 70.0% 80.0% 90.0% 100.0%

L/C-ratio (RR Manifold, RR Loop)

a/t-ratio

Figure D.7 Crack Depth Versus Crack Length in Austenitic Stainless Steel (NPS22, NPS28) 
D-21

477

376

248

183

139

35 18 11 4

0

100

200

300

400

500

600

> 10% > 20% > 30% > 40% > 50% > 60% > 70% > 80% > 90%

a/t-Ratio

No. db Records

Figure D.8 Summary of IGSCC Non-Through Wall Flaws by ‘a/t-Ratio’ 

In Figure D.5, the combination of a/t = 100% and L/C = 100% would indicate a case of DEGB where the 

pipe ends are separated from each other. As a rule-of-thumb, a through-wall crack (a/t = 100%) with L/C ≥

40% is unstable and may exhibit unstable crack growth if it were to be left in place.3

As seen from the above, there have been a limited number of cases of leaks in large-diameter Reactor 

Recirculation piping. Only a small fraction of the total number of through-wall flaws have been active leaks; 

i.e., leaks that have developed during routine power operation. The majority of the through-wall flaws have 

been “non-active leaks.” That is, leaks that have developed while shutting down for drywell inspection, 

during performance of weld crown grinding in preparation for ultrasonic examination (“ISI-leaks”), or during 

the performance of induction heat stress improvement (IHSI – “IHSI-leaks”). There are also some cases 

where leaks have been discovered during hydrostatic pressure testing to verify the integrity of weld repairs. 

Like Figure D.5, Figure D. 8 includes data on all IGSCC-susceptible, Code Class 1 and 2 piping systems in 

BWR plants. While Figure D.5 includes approximately 300 data points, Figure D.8 includes on the order of 

500 data points. This difference in the number of reports represented in respective chart is due to the fact that 

not all reports on IGSCC include complete details on the crack morphology (dimensions, orientation). 

Where through-wall flaws have been observed leak rates have been small. In terms of leak rate and 

operational impact, so far the two most significant instances of IGSCC occurred at Duane Arnold in 1978 and 

at the Spanish plant Santa Maria de Garona in 1980. In the former case the leak rate was about 11 lpm (3 

gpm) with L/C = 22%. In the latter case the observed leak rate was about 3.0 lpm (0.8 gpm) with L/C = 4.5%. 

 

3

 See for example the report EPRI NP-2472 (The Growth and Stability of Stress Corrosion Cracks in Large-Diameter BWR Piping, July 1982). 
D-22

D.3.2.2 FW Piping Service Experience - Figures D.9 and D.10 summarize the service experience with 

FW piping. With respect to plant designed by General Electric, the Code Class 1 portion of BWR carbon 

steel feedwater piping has performed well in the field. There are no reported leaks in medium-or largediameter RCPB piping. Foreign plants have experienced (and in some cases, continue to experience) thermal 

fatigue damage due to thermal mixing and stratification. In fact, 80% of the degradation of the RCPB 

portions of FW piping has occurred in foreign plants with a piping system design that differs from that of 

U.S. BWR plants. 

The U.S. service experience includes a few instances of non-through wall cracking of FW nozzle-to-safe-end 

(bimetallic) welds. The root cause of the cracking is attributed to weld defects from original construction. As 

documented in Information Notice 92-35 [D.19], Susquehanna Unit 1 has experienced flow-accelerated 

corrosion damage about 250 mm (10 inches) from a weld connecting NPS12 piping to a 20-inch by 12-inch 

reducing tee. There have been no reported flaws in any U.S. plant beyond T = 15 years of operation. 

0

2

4

6

8

10

12

14

1970

1972

1974197619781980

1982198419861988

1990

1992

19941996199820002002

No. db Records

FW Piping Inside Dryw ell

FW Piping Outside Dryw ell

Figure D.9 Service Experience with FW Piping (i) 
D-23

0

2

4

6

8

10

12

1

3579

111315

17

19

21

23

25

2729

Years of Operation

No. db Records

FW Piping Inside Dryw ell

FW Piping Outside Dryw ell

Figure D.10 Service Experience with FW Piping (ii)

D.3.3 Review of PWR-Specific Piping Service Experience 

Limited to the PWR Base Case systems, this section summarizes the service experience with Code Class 1

piping. The results of this review are input to the pipe failure rate estimation. 

D.3.3.1 RC & HPI/NMU Piping Service Experience - There have only been a limited number of events 

involving through-wall cracks in the large-diameter RC piping and the Class 1 portion of SI/CV piping.

Evidence of axial primary water stress corrosion cracking (PWSCC) in the bimetallic safe-end to RPV nozzle 

welds of the RC-HL piping has been reported at Ringhals [D.20] and V.C. Summer [D.21]. 

During an eight-year period, the now decommissioned Trojan nuclear power plant experienced pressurizer

surge line movement, which was attributed to thermal stratification [D.22]. In response, the NRC issued 

Bulletin 88-11 in December of 1988 [D.23] requesting that licensees perform visual inspections of the 

pressurizer surge line at the first available cold shutdown. Purpose of the inspections was to determine 

presence of any “gross discernible distress or structural damage in the entire pressurizer surge line, including 

piping, pipe supports, pipe whip restraints, and anchor bolts.” 

The current version (June 2004) of the PIPExp database includes four records associated with degradation of 

pressurizer surge lines: 

• Record # 19849; during the Three Mile Island-1 2003 Refueling Outage (18-Oct-2003 to 3-Dec2003), a UT examination found an axial flaw about 13 mm (0.5-inch) deep in the surge line nozzleto-safe end interface in dissimilar metal weld No. SR0010BM. This weld connects a 10-inch 

Schedule 140, carbon steel nozzle to stainless steel safe end. 

• Record # 19736; in November 2002 during UT examination of RC piping in the Belgian plant 

Tihange-2 (a 900 MWe series plant designed by Framatome), code rejectable indications were 
D-24

discovered in the 14-inch Inconel safe-end to nozzle weld. The flaw is believed to be an original 

construction defect. 

• Record # 1119; while in hot shutdown condition, a non-isolable weld leak developed in a 1-inch 

drain line off the pressurizer surge line of Oconee-1 (LER 50-269/1998-002-01). The through-wall 

crack had initiated by TGSCC and propagated through-wall by vibratory fatigue. Small-diameter 

piping connecting to a pressurizer surge line is not part of the PWR-2 Base Case definition. 

• Record # 420; during the 1988 annual refueling outage a pinhole leak was discovered in a 10-inch 

pressurizer surge line bi-metallic weld of Loviisa-1 (a Soviet designed WWER-440/213 plant located 

in Finland). The weld degradation was attributed to poor weld penetration and high residual stresses. 

This event was screened out from the data analysis.

Figures D.11 and D.12 summarize relevant service experience with medium- and large-diameter RC and 

safety injection (SI) and normal makeup (CV) piping. For comparison, Figure D.13 shows the service 

experience with small-diameter RC and SI/CV piping (≤ NPS2). Figure D.14 is a summary of the worldwide, 

PWR-specific data pipe failures that are attributed to thermal fatigue. In addition to RC-, SI- and CV-piping 

this figure includes failures in FW- and RHR-piping. 

0

2

4

6

8

10

12

1970197219741976

1978198019821984198619881990

19921994

19961998

20002002

No. db Records

2 < ∅ ≤ 4

4 < ∅ ≤ 10

∅ > 10

Figure D.11 Weld Failures in PWR RC-, CV- and SI-Piping (1970-2002) 
D-25

0

1

2

3

4

5

6

7

8

9

10

1

35

79

11

1315

1719

2123

25

2729

Years of Operation

No. db Records

2 < ∅ ≤ 4

4 < ∅ ≤ 10

∅ >10

Figure D.12 Weld Failures in PWR RC-, CV- and SI-Piping (1970-2002) 

0

20

40

60

80

100

120

1

3

5

7

9

11

1315

1719

21

232527

29

31

Years of Operation

No. db Records

∅ ≤ 2

Figure D.13 Weld Failures in Small-Diameter PWR Piping 
D-26

0

1

2

3

4

5

1

3

57

9

11

13151719

21

23252729

Year(s) of Operation

No. db Records

Large Leak (> 10 gpm)

Leak

Crack

Figure D.14 Pipe Failures Attributed to Thermal Fatigue in PWRs Worldwide 

Figure D.14 includes four significant (ν > 38 lpm [10 gpm]) events, three in foreign plants (Civaux-1 in 

France, Tsuruga-2 in Japan and Biblis-B in Germany) and one in a domestic plant (Oconee-2). The latter

event involved a failure of a weld between the HPI/NMU and the RCS cold leg (= PWR Base Case Plant 

A.b). The plant operators correctly diagnosed the leak and brought the plant to safe shutdown. Subsequent to 

the weld failure in Oconee-2, limited to small-diameter piping the Electric Power Research Institute issued 

the “Interim Thermal Fatigue Guideline” [D.9] for evaluating and inspecting regions where there might be 

high potential for thermal fatigue cracking. Additional perspectives on thermal fatigue mitigation are included 

in an OECD-NEA report [D.24]. The Babcock &Wilcox-designed plants now include a new design thermal 

sleeve to mitigate or prevent thermal fatigue cracking of welds. 

Prior to these ‘four significant events’, thermal fatigue damage occurred at Farley-2 and Tihange-1 (a Belgian 

plant) during 1988. At these plants, thermal fatigue initiated from cold water leaking through closed check or 

globe valves in safety injection lines. At Farley-2, the damage occurred in piping connected to the RCS cold 

leg, and at Tihange-1 in piping connected to the RCS hot leg. In these events the leak rates were 2.6 lpm (0.7 

gpm) and 23 lpm (6 gpm), respectively. The U.S. NRC issued Bulletin 88-08 in response to these events.

D.4 Data Processing and Data Reduction

The objective of data processing is to extract from a pipe failure data collection relevant case histories that 

reflect specific combinations of reliability attributes and influence factors. Next, the data reduction prepares 

the input to the statistical parameter estimation in the form of event counts and exposure terms to develop 

Bayesian prior and posterior distributions. 
D-27

D.4.1 Strategy for Data Processing and Data Reduction 

Shown in Figure D.15 is a general four-state Markov model of piping reliability. All failure processes of this 

model can be evaluated using service data, assuming that such a data collection is of sufficient technical 

detail and completeness. This model is used in Section D.6 to develop time-dependent LOCA frequencies.

Piping Reliability States: 

S = Success (or undamaged state); 

C = Crack (non-through wall flaw); 

F = Leaking through-wall flaw (leak rate is within Technical 

Specification limit); 

L = Large leak (leak rate in excess, or well in excess of 

Technical Specification limit). 

State Transitions: 

φ Occurrence of non-through wall flaw 

λC Occurrence of small leak given a flaw (‘C’) 

λF Occurrence of large leak given a through-wall 

flaw (‘F’) 

ρS Occurrence of large leak given no flaw 

ρC Occurrence of large leak given a non-through 

wall flaw 

ρF Occurrence of large leak given a small leak 

μ Detect and repair a through-wall flaw 

ω Inspect and repair a non-through wall flaw 

Figure D. 15 Four-State Markov Model of Piping Reliability4

In this and subsequent report sections, a pipe failure (F) is defined as a through-wall defect resulting in a nonactive leak or small, active leak. The frequency of a large leak (L) in excess of Technical Specification limits 

is estimated using the following simple model: 

FL = λ × pLF (D.1) 

 

4

 The figure is reproduced courtesy of K.N. Fleming (Technology Insights, Inc., San Diego, California).

S

C

F

L

ρF

φ ω

μ

ρC

ρS

λC

λS
D-28

λ = [No. Failures] / [Exposure (= T × No. Components)] (D.2) 

Where: 

FL = Frequency of a large leak [1/Reactor-year]. 

λ = Failure frequency [1/Reactor-year.Extension]; where ‘extension’ refers to the piping 

component boundary definition. Depending on the intended application and type(s) of 

degradation mechanism, the extension could be formed by counts of bends, pipes, tees, 

welds or length of piping. In Equation (4.2), the exposure term reflects the total component 

population in the data survey. 

T = Exposure time (or reactor operating years) 

pLF = Conditional probability of a large leak given a through-wall defect. Section D.5 includes a 

technical basis for estimating conditional failure probabilities. 

The parameter estimation uses a Bayes’ update process that begins with the development of prior 

distributions for each of the terms in equation (D.1). These prior distributions are shaped by our knowledge 

about the susceptibility of different piping systems to degradation. The input to the Bayes’ update process 

comes from a small subset of PIPExp after it has been subjected to screening for pipe failures that meet 

certain selection criteria. A software tool (Bayesian Analysis Reliability Tool -BART

) is used to perform the 

updates.5

Piping reliability is a function of pipe size (diameter and wall thickness), and metallurgy, process medium, 

environment and design requirements; or attributes and influences, respectively. The purpose of data 

processing and data reduction is to extract from the total PIPExp database those subsets of service data that 

correspond to the attributes and influences of the Base Case definitions. 

D.4.1.1 Informative versus Noninformative Prior Distributions - The type, extent and quality of 

applicable service data will determine the actual implementation of the Bayesian update process. Where

sufficient service data is available an empirical Bayes approach is used. In this case classical estimation 

techniques are used to fit a prior distribution to the available data. When no or sparse service data is available 

a non-informative prior is defined. Relative to the five Base Cases the following approaches are used to 

determine the prior failure rates distributions: 

• BWR Base Case – RR Loop B. There is ample service data on IGSCC. Our prior state-of-knowledge 

consists of service data before implementation of IGSCC mitigation strategies (mid-1980s). A prior 

failure rate is derived through classical statistical estimation. 

• BWR Base Case – FW Loop B. Given the scarce service data, a lognormal distribution with a mean 

value of 1.0E-06 per weld-year and range factor (RF) equal to 100 is used. This is a noninfomative 

prior distribution. 

• PWR Base Case – RC Hot Leg. The only available service data involves axial cracks in RPV 

nozzle-to-safe-end welds at three PWR units. A point estimate for the failure rates is calculated for 

 

5

 Details on Bayesian reliability analysis is found in text books on statistical analysis of reliability data; e.g., Martz and Waller (1991): 

Bayesian Reliability Analysis, Krieger Publishing Company, Malabar (FL), ISBN 0-89464-395-9. For conjugate functions like the 

gamma and beta distributions a Bayesian point estimator for the failure rate is the mean of respective posterior probability density 

function, or: 

λ = (δ + r)/(ρ + T) – gamma 

λ = (δ + r)/(δ + ρ + n) – beta 

Where, (δ, ρ) are the parameters of respective distribution and (r, T, n) correspond to new evidence (i.e., ‘r’ failures in ‘T’ 

hours, or ‘r’ failures in ‘n’ tests). 
D-29

the period 1970 through 2000. This point estimate is approximated by a lognormal distribution with 

range factor of 100; i.e., essentially a noninformative prior. 

• PWR Base Case – RC Surge Line. For the pressurizer surge line there is no service data including 

non-through wall or through-wall cracking. Again, a lognormal distribution with a mean value of 

1.0E-06 per weld-year and RF = 100 is used. 

• PWR Base Case – HPI/NMU Line. Service data exists, which is directly applicable to this base case. 

To account for design changes that have been implemented post-1997, a non-informative prior is 

combined with B&W-specific failure data and exposure data through end of calendar year 1997. The 

resulting failure rate represents a prior distribution, which is applicable to this Base Case. 

D.4.2 BWR-Specific Apriori Pipe Failure Rates 

The failure rate development consists of determining an industry generic pipe failure rates for RR and FW 

piping, respectively. Next a Bayesian update is performed to generate failure rates that best represent the 

design and operating conditions of Plant B. Rather than taking an apriori failure from some published source, 

the approach in this study is to derive apriori failure information from the PIPExp database. The information 

that is summarized in Section D.3 provides some insights into the time-dependency of failure rates. These 

insights are explored further below. 

D.4.2.1 RR Pipe Failure Data - Programs to mitigate the effects of certain degradation mechanisms 

strongly influence the achieved piping reliability. As an example, all BWR plants commissioned prior to the 

early to mid-1980s have experienced IGSCC. Industry initiatives to mitigate or eliminate the influence by 

IGSCC were implemented by the mid-1980s, and thereafter the rate of IGSCC has dropped sharply. The 

trend in the IGSCC rate is established by normalizing the data displayed in Figure D.2 (IGSCC by Years of 

Operation). Calculating the rate of IGSCC per weld-year for a given system and pipe size performs the 

normalization. Before performing this normalization the database is subjected to additional processing to 

exclude from further consideration any IGSCC data not directly applicable to the RR System that is 

representative of the Base Case. Similarly the part of the database including plant population and weld 

population data must be processed in such a way that an appropriate exposure term is developed 

commensurate with the failure data. In developing the RR-specific exposure term the following exclusion 

criteria were applied: 

Plant Population Exclusion Criteria Applicable to RR Piping

• BWR plants without external recirculation loops; 

• BWR plants in which the RR piping is fabricated from IGSCC resistant material e.g., Nuclear Grade 

stainless steel. 

Table D.3 includes selected weld counts used to derive an exposure term according to Equation (D.2). 

Organized by pipe size and years of operation, Table D.4 is a summary of weld failures in RR piping. 

Noteworthy is the observation that there have been no reported through-wall defects in any plant beyond T = 

15 years of operation. Using the information in Tables D.3 and D.4, Figure D.16 shows the calculated rate of 

RR pipe failure per weld-year. 

The failure rates in Figure D.16 assume that all RR welds of a certain size to be equally susceptible to 

IGSCC. As was shown in SKI Report 98:30 [D.15], a correlation exists between weld failure rate and weld 

configuration. This correlation is assumed to be attributed to the piping layout, complexity of welding 

operation, and the associated weld residual stresses. The chart in Figure D.17 shows the weld configuration 

versus fraction of weld failures. 
D-30

Table D.3 Selected Weld Counts in RR Piping 

Plant ID Weld Count by Pipe Size (NPS) 

(NSSS Type) 4 6 8 12 14 16 22 24 28 

1 (AA/3)6 30 42 23 -- -- -- -- 54 -- 

2 (BWR/4) 8 -- -- 58 -- -- 12 -- 33 

3 (BWR/4) -- -- -- 34 -- 2 -- 4 24 

4 (BWR/5) 36 -- -- 51 10 16 -- 50 -- 

5 (BWR/5) 39 -- -- 63 -- 16 -- 45 -- 

6 (BWR/5) -- -- -- 130 -- -- 24 -- 97 

7 (BWR/5) -- -- -- 138 -- -- 16 -- 97 

8 (BWR/4) -- -- -- 24 -- -- 4 -- 28 

9 (BWR/4) -- -- -- 25 -- -- 4 -- 38 

10 (BWR/4) 12 -- -- 59 -- -- 10 -- 36 

11 (BWR/4) 12 -- -- 62 -- -- 12 -- 36 

Plant B (BWR/4) -- -- -- 50 -- -- 16 -- 56 

Mean: 23 -- -- 63 -- -- 12 -- 49 

Table D.4 Number of Through-Wall Flaws in RR Piping Attributed to IGSCC7

Pipe Diameter (∅) Years of Operation 

[NPS] Total 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 

3 < ∅ ≤ 6 15 0 1 1 2 4 1 0 2 2 0 2 0 0 0 0 

6 < ∅ ≤ 12 9 0 0 0 0 1 0 1 0 1 1 2 1 1 0 1 

12 < ∅ ≤ 22 4 0 0 0 0 0 0 1 0 0 0 2 1 0 0 0 

∅ > 22 16 0 0 0 0 0 0 0 0 5 0 6 4 0 1 0 

 44 0 1 1 2 5 1 2 2 8 1 12 6 1 1 1 

1.00E-05

1.00E-04

1.00E-03

1.00E-02

0 2 4 6 8 10 12 14

Year(s) of Operation

λ [1/weld-yr.]

3 < ∅ ≤ 6

6 < ∅ ≤ 12

12 < ∅ ≤ 22

∅ > 22

Figure D.16 Rate of IGSCC-Induced RR Pipe Failure (‘Prior State-of-Knowledge’) 

 

6

 See SKI 98:30 [D.15] for details. 

7

 This table includes active leaks (= leaks detected during routine power operation) and ‘non-active’ leaks (= leaks discovered during 

change of plant mode of operation), but it excludes ‘ISI-leaks.’ Appendix A, Table A-5 includes details on the through-wall cracks in 

NPS12, NPS22 and NPS28 Reactor Recirculation piping as included in Table D.4 above. 
D-31

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7

Elbow-to-pipe

Nozzle-to-safe-end

Pipe-to-pipe

Pipe-to-safe-end

Pipe-to-sweepolet

Pipe-to-reducer

Pipe-to-tee

Pipe-to-end-cap

Elbow-to-valve

Pipe-to-valve

Elbow-to-pump

Pipe-to-cross

Pipe-to-pump

Reducer-to-cross

Weld Configuration

Fraction of Weld Failure Rate

∅ > 22

12 < ∅ ≤ 22

6 < ∅ ≤ 12

Figure D.17 RR Weld Failures as a Function Weld Configuration 

In addition to weld configuration, the likelihood of failure is also a function of pipe size. For a weld of type 

“i” and size “j” the failure rate is expressed as follows: 

λij = Fij/(Wij × T) (D.3)

and with 

Sij = Fij / Fj

 (D.4) 

Aij = Wj/Wij (D.5) 

the failure rate of weld of type “i” and size “j” is expressed as 

λij = (F × Sij)/(Wij × T) (D.6) 

λij = Sij × Aij× λj (D.7) 
D-32

Where: 

λij = Failure rate of an IGSCC-susceptible weld of type “i”, size “j” 

λj = Failure rate of an IGSCC susceptible weld of size ‘j’ 

Fj = Number of size “j” weld failures 

Fij = Number of type “i” and size “j” weld failures 

Wj = Size “j” weld count 

Wij = Type “i” and size “j” weld count 

Susceptibility 

(Sij) 

= The service experience shows the failure susceptibility to be correlated with the 

location of a weld relative to pipe fittings and other in-line components (flanges, 

pump casings, valve bodies). For a given pipe size and system, the susceptibility 

is expressed as the fraction of welds of type “ij” that failed due to a certain 

degradation mechanism). This fraction is established from the PIPExp database; 

Table D.5. 

Attribute 

(Aij) 

= In the above expressions the attribute (A) is defined as the ratio of the total 

number of welds of size “j” to the number of welds of type “i”. In expression 

(4.7), Aij is a correction factor and accounts for the fact that piping system design 

& layout constraints impose limits on the number of welds of a certain type. For 

example, in a given system there tends to be more elbow-to-pipe welds than, say, 

pipe-to-tee welds. 

Combining the global (or averaged) failure rates in Figure D.16 with the information summarized in Figure 

D.17 and Table D.5 provides the apriori failure rates that are input to Equation (D.1). The results are 

summarized in Section D.4.3. 

Table D.5 IGSCC Susceptibility by Weld configuration – Selected Parameter Values 

RR System Configuration Dependent Parameters 

[NPS] Weld Configuration Susceptibility (Sij) Attribute (Aij) 

12 Elbow-to-pipe 6.03E-01 2.8 

 Nozzle-to-safe-end 1.35E-01 5.0 

 Pipe-to-reducer 2.38E-02 25.0 

22 Pipe-to-end-cap 2.71E-01 4.0 

 Pipe-to-sweepolet 8.33E-02 2.0 

 Pipe-to-cross 6.25E-02 4.0 

28 Elbow-to-pipe 4.62E-02 5.6 

 Pipe-to-pipe 5.77E-02 3.1 

 Cross-to-reducer 9.60E-03 28.8 

D.4.2.2 FW Pipe Failure Data - The estimation of failure rates for FW piping uses a non-informative prior 

distribution together with the weld population data in Table D.6. This approach is selected based on the 

available, limited service experience with ASME XI Class 1 FW piping; Table D.7. In developing the data 

summary in Table D.7 the following FW exclusion criteria were used to develop a point estimate of the 

failure rate: 

Failure Data Exclusion Criteria Applicable to FW Piping

• Piping external to the drywell containment structure; 

• Non-US data. 
D-33

Table D.6 Selected Weld Counts in ASME XI Class 1 FW Piping 

Plant ID Weld Count by Pipe Size (NPS) 

(NSSS Type) 8 10 12 14 16 18 20 22 24 

1 (AA/3) 1 61 -- 14 -- -- -- -- -- 

2 (BWR/4) -- -- 28 -- -- 25 -- -- -- 

3 (BWR/4) -- -- 28 -- -- 26 -- -- -- 

4 (BWR/5) -- -- 40 12 -- 6 -- -- 38 

5 (BWR/5) -- -- 41 6 6 -- -- 42 

6 (BWR/5) 13 -- 66 -- 8 -- 7 -- 22 

7 (BWR/5) 9 -- 68 -- 8 -- 7 6 19 

9 (BWR/4) -- -- 50 -- -- -- 6 -- 45 

10 (BWR/4) -- -- 32 1 -- 30 -- -- -- 

11 (BWR/4) -- -- 30 1 -- 29 -- -- -- 

Plant B 

(BWR/4) -- -- 63 5 -- -- 53 -- -- 

Mean: -- -- 42 -- -- -- 418 -- -- 

Table D.7 Summary of FW Pipe Failure Data 

Location Data Origin Failure Mode 

Pipe Size (∅) 

[NPS] Drywell Ex-drywell US NonUS 

Crack / 

Wall 

Thinning 

P/H-leak Leak Rupture 

3 < ∅ ≤ 6 YES YES 0 0 0 0 

6 < ∅ ≤ 12 YES YES 4 0 0 0 

∅ > 12 YES YES 1 0 0 0 

3 < ∅ ≤ 6 YES YES 5 0 0 0 

6 < ∅ ≤ 12 YES YES 9 0 0 0 

∅ > 12 YES YES 5 0 0 0 

3 < ∅ ≤ 6 YES YES 1 2 19 3 

6 < ∅ ≤ 12 YES YES 1 1 7 0 

∅ > 12 YES YES 2 0 3 0 

3 < ∅ ≤ 6 YES YES 3 0 1 0 

6 < ∅ ≤ 12 YES YES 2 0 2 0 

∅ > 12 YES YES 0 1 1 0 

D.4.3 PWR-Specific Apriori Pipe Failure Rates 

As summarized in Section D.3.3, there have been only a few through-wall defects in Class 1 PWR piping.

For the RC-HL, the rate of PWSCC per weld-year is established using the normalization process discussed in 

Section D.4.2.1 and with the following specializations: 

 RC-HL Apriori Failure Rate

 

8

 The mean of weld count in NPS20-, 22- and 24-piping. 
D-34

• The apriori failure rate is derived using the PIPExp for the time-period 1970 through 2000 to 

include the consideration of the through-wall defect at V.C. Summer. 

• Table D.8 includes the weld population data used to calculate λNPS30 = 8.12E-05 per weld-year. 

• Failure rate “post-processing” to account for different weld configuration susceptibilities to 

PWSCC is done consistent with Section D.4.2 and with the Sij and Aij assumed values shown in 

Table D.9. 

Table D.8 Selected Weld Counts in Code Class 1 PWR Piping 

Plant ID Weld Count by Pipe Size [NPS] 

(NSSS Type) 3-3⁄4 10 12 14 309

1 (WEST/4) -- 24 18 1 84 

2 (WEST/4 -- 24 18 1 52 

3 (WEST/4) -- 24 18 -- 92 

4 (WEST/4) -- 24 14 -- 68 

Plant A.a (WEST/3) -- -- 5 14 50 

Plant A.b (B&W; 

HPI/NMU system only) 9 -- -- -- -- 

Table D.9 Degradation Susceptibility by Weld Configuration 

RC System Configuration Dependent Parameters

(NPS) Weld Configuration Susceptibility (Sij) Attribute (Aij)

30 Nozzle-to-safe-end 8.00E-01 12.5 

(RC Hot Leg) Elbow-to-safe-end 8.00E-02 12.5 

 Elbow-to-pump 5.00E-02 12.5 

 Pipe-to-pump 4.00E-02 12.5 

 Elbow-to-pipe 3.00E-02 1.5 

14 Nozzle-to-safe-end 5.00E-01 14.0 

(Surge Line) Pipe-to-safe-end 2.50E-01 14.0 

 Branch-to-pipe 5.00E-02 14.0 

 Branch-to-HL 1.50E-01 14.0 

 Elbow-to-pipe- 5.00E-02 1.40 

3-3⁄4 Elbow-to-nozzle 8.50E-01 9.0 

(HPI/NMU) Elbow-to-pipe 1.00E-01 2.25 

 Elbow-to-valve 4.50E-02 3.0 

 Pipe-to-pipe 5.00E-03 9.0 

In contrast to the BWR weld susceptibility factors in Table D.5, the weld susceptibility factors in Table D.10 

are assumed values that reflect the applicable service experience. As an example, for the RC Hot Leg the 

nozzle-to-safe-end weld is assigned the highest value in view of the available service experience; i.e., the 

Ringhals and V.C. Summer hot leg cracking as described in Section D.3.3.1. As another example, for the RC 

Surge Line, relatively high weld susceptibility factors are assigned the safe-end welds and Hot Leg branch 

connection. In view of the recent experience at TMI-1, the nozzle-to-safe-end weld is given a greater weight 

than other weld configurations. The uncertainty in the PWR weld susceptibility factors is not evaluated 

 

9

 NPS30 is used to characterize the CL- and HL-piping. 
D-35

further in this study, however. EPRI TR-11188010 is used for characterizing the prior knowledge about pipe 

failure due to thermal fatigue. 

D.4.4 Prior Distributions for Bayesian Updating 

Included for comparison, Figure D.18 shows the calculated flaw rates (non-through wall). Listed in Table 

D.10 are the RR and FW weld failure rates that represent the state-of-knowledge at T = 15 years of operation. 

Listed in Table D.11 are the RC and HPI/NMU prior weld failure rates. The failure rates represent the 

frequency per weld-year of a through-wall flaw resulting in a leakage of less than or equal to the Technical 

Specification limit for unidentified leakage. In summary, the derivation of prior weld failure rates includes the 

following steps: 

• Determine the number of through-wall leaks from PIPExp database. Includes performing a trend 

analysis. 

• From the PIPExp database, determine the appropriate exposure terms. 

• Establish the susceptibility of different weld locations to degradation. 

• Combine the output from the previous steps to determine prior failure rates applicable to welds in the

RR and FW systems. 

1.00E-06

1.00E-05

1.00E-04

1.00E-03

1.00E-02

> 10% > 20% > 30% > 40% > 50% > 60% > 70% > 80% > 90%

a/t-Ratio

λNTWC [1/Weld.Yr]

NPS12

NPS22

NPS28

Figure D.18 Prior Frequency of Non-Through Wall IGSCC in RR Piping 

 

10 Table 2-3, page 2-10; λj

 = 1.34E-05 (RF = 100). TR-111880: Piping System Failure Rates and Rupture Frequencies for Use in Risk

Informed In-Service Inspection (September 1999). 
D-36

Table D.10 Prior RR and FW Weld Failure Rates 

Pipe Size Lognormal Distribution Parameters

System [NPS] Weld Configuration Mean [1/Weld-yr] Range Factor 

(RF)

RR 12 Elbow-to-pipe 3.95E-04 10 

 Nozzle-to-safe-end 1.59E-04 10 

 Pipe-to-safe-end 7.49E-05 15 

 Pipe-to-sweepolet 7.02E-05 15 

 Pipe-to-reducer 1.40E-04 15 

RR 22 Pipe-to-end-cap 7.67E-04 10 

RR 22 Pipe-to-sweepolet 1.18E-04 15 

 Pipe-to-cross 1.77E-04 15 

RR 28 Pipe-to-elbow 7.19E-04 10 

 Nozzle-to-safe-end 1.50E-04 10 

 Pipe-to-safe-end 6.74E-04 10 

 Pipe-to-valve 2.25E-04 10 

 Pipe-to-pump 2.25E-04 10 

 Pipe-to-tee 1.25E-04 10 

 Pipe-to-pipe 4.99E-05 20 

 Pipe-to-cross 7.49E-05 20 

 Pipe-to-reducer 7.49E-05 20 

FW 12 Nozzle-to-safe-end 6.20E-06 100 

 Elbow-to-pipe 3.32E-07 100 

 Pipe-to-pipe 5.17E-07 100 

 Pipe-to-reducer 1.55E-06 100 

 Pipe-to-reducing-tee 4.65E-07 100 

FW 20 Elbow-to-pipe 4.00E-06 100 

 Elbow-to-valve 6.00E-07 100 

 Pipe-to-reducing-tee 4.80E-07 100 

 Pipe-to-reducer 7.20E-07 100 

No failure rates derived for welds in NPS14 piping; it is assumed that the data on welds in NPS12 

piping is also representative of welds in NPS14 piping. 
D-37

Table D.11 Prior RC and HPI/NMU Weld Failure Rates

Pipe Size Lognormal Distribution Parameters

System [NPS] Weld Configuration Mean [1/Weld-yr] RF

RC 30 Nozzle-to-safe-end 8.12E-04 100 

(Hot Leg) Elbow-to-pump 5.07E-05 100 

 Pipe-to-pump 4.06E-05 100 

 Elbow-to-pipe 3.65E-06 100 

RC 14 Branch-to-HL 2.10E-06 100 

(Surge Nozzle-to-safe-end 7.00E-06 100 

Line) Pipe-to-safe-end 3.50E-06 100 

 Branch-to-pipe 7.00E-07 100 

 Elbow-to-pipe 7.00E-08 100 

HPI/NMU 3-3⁄4 Elbow-to-nozzle 9.86E-04 16 

 Pipe-to-pipe 6.48E-06 100 

 Elbow-to-pipe 1.90E-06 100 

 Elbow-to-valve 1.31E-06 100 

Note to Data on Pressurizer Surge Line: 

• Susceptibility factors from Table D.9 applied directly to prior knowledge. 

Notes to Data on HPI/NMU Line: 

• Susceptibility factors from Table D.9 applied to prior knowledge. 

• To develop a B&W-specific weld failure rate, the prior knowledge (data from TR-111880; see 

footnote #18) was combined with B&W-specific service data and exposure data from PIPExp 

through end of calendar year 1997. The failure rate for weld type ‘elbow-to-nozzle’ accounts for 

the weld failure at Oconee-2 in 1997. 

D.5 Data for LOCA Frequency Estimation 

Using the information in Section D.4, this section documents the input data to the LOCA frequency model. A 

Bayesian update is performed to develop posterior weld failure rates. The frequency of leaks exceeding

Technical Specification limits are developed through estimates of the conditional probability of a large leak 

given a small through-wall flaw. 

D.5.1 Posterior Weld Failure Rates

The failure rate calculation involves two factors, the number of applicable failures and the exposure data. To 

account for uncertainty in the exposure data, which could influence the failure rate calculation the following 

process is used. First, a best estimate update is performed using the appropriate number of failure events and 

the number of welds of exposure. To account for plant-to-plant variability in the weld exposure term, a 

second update is performed using the same failure data but an exposure estimate that is 50% higher, and a 

third update using an exposure estimate that is 50% lower. Each of the three updates is combined in a 

posterior weighting process using the following weights: 50% for the best estimate, 25% for the high 

exposure case and 25% for the low exposure case. The result is an uncertainty distribution for each failure 

rate, which reflects greater uncertainty than the best estimate data would imply alone. The results are given in 

Tables D.12 and D.13; Attachment B includes the input to the failure rate calculation. Figure D.19 displays 

posterior IGSCC flaw frequencies. Figures D.20-D.23 compare the prior and posterior non-through wall 

crack frequencies. 
D-38

Table D.12 Posterior RR and FW Weld Failure Rate Distributions – BWR Base Cases 

Pipe Size 

Failure Rate Uncertainty Distribution Parameters 

[(≤TS Leak)/Weld-yr] 

System [NPS] Weld Configuration Mean 5%-tile 50%-tile 95%-tile 

RR 12 Elbow-to-pipe 4.32E-05 8.48E-06 3.17E-05 1.16E-04 

 Nozzle-to-safe-end 4.38E-05 5.52E-06 2.72E-05 1.36E-04 

 Pipe-to-safe-end 2.99E-05 2.98E-06 1.70E-05 9.64E-05 

 Pipe-to-sweepolet 3.14E-05 2.80E-06 1.71E-05 1.06E-04 

 Pipe-to-reducer 7.82E-05 5.71E-06 3.97E-05 2.77E-04 

RR 22 Pipe-to-end-cap 1.54E-04 2.28E-05 1.01E-04 4.52E-04 

 Pipe-to-cross 4.24E-05 4.38E-06 2.47E-05 1.37E-04

 Pipe-to-sweepolet 7.37E-05 7.02E-06 4.09E-05 2.40E-04 

RR 28 Pipe-to-elbow 8.52E-05 1.59E-05 6.07E-05 2.33E-04 

 Nozzle-to-safe-end 6.55E-05 5.95E-06 3.61E-05 2.15E-04 

 Pipe-to-safe-end 1.44E-04 2.11E-05 9.36E-05 4.28E-04 

 Pipe-to-valve 5.96E-05 7.68E-06 3.75E-05 1.84E-04

 Pipe-to-pump 8.36E-05 8.68E-06 4.85E-05 2.71E-04 

 Pipe-to-tee 5.78E-05 5.06E-06 3.13E-05 1.96E-04 

 Pipe-to-pipe 1.29E-05 5.74E-07 5.25E-06 4.78E-05 

 Pipe-to-cross 3.86E-05 7.89E-07 1.08E-05 1.50E-04

 Reducer-to-cross 3.86E-05 7.89E-07 1.08E-05 1.50E-04 

FW 12 Nozzle-to-safe-end 2.29E-06 8.61E-10 6.88E-08 5.29E-06 

 Elbow-to-pipe 1.75E-07 4.61E-11 4.28E-09 3.74E-07

 Pipe-to-pipe 2.78E-07 7.39E-11 6.89E-09 6.33E-07 

 Pipe-to-safe-end 2.43E-07 6.20E-11 5.97E-09 5.50E-07 

 Pipe-to-reducer 9.73E-07 2.38E-10 2.24E-08 2.12E-06 

FW 12 Elbow-to-reducing-tee 3.33E-07 7.46E-11 7.11E-09 6.94E-07 

FW 20 Pipe-to-elbow 1.62E-06 5.57E-10 4.61E-08 3.71E-06 

 Pipe-to-pipe 4.10E-07 9.57E-11 8.77E-09 8.00E-07 

 Pipe-to-valve 3.38E-07 7.09E-11 6.78E-09 6.53E-07

 Elbow-to-valve 3.54E-07 9.22E-11 8.80E-09 8.39E-07 

 Pipe-to-tee 4.50E-07 7.33E-11 7.20E-09 7.25E-07 

 Pipe-to-reducer 5.68E-07 1.17E-10 1.14E-08 1.14E-06 

 Tee-to-valve 2.18E-07 4.06E-11 4.12E-09 4.06E-07 

Table D.13 Posterior RC and HPI/NMU Weld Failure Rate Distributions – PWR Base Cases 

Pipe Size 

Failure Rate Uncertainty Distribution Parameters 

[(≤TS Leak)/Weld-yr] 

System [NPS] Weld Configuration Mean 5%-tile 50%-tile 95%-tile 

RC 30 Nozzle-to-safe-end 7.64E-05 2.12E-07 7.34E-06 2.61E-04 

(Hot Leg) Elbow-to-pump 1.96E-05 7.59E-09 5.36E-07 4.02E-05 

 Pipe-to-pump 1.24E-05 5.78E-09 4.47E-07 3.17E-05 

 Elbow-to-pipe 1.05E-06 5.38E-10 3.94E-08 3.04E-06

RC 14 Branch-to-CL 1.14E-06 2.90E-10 2.69E-08 2.42E-06 

(Surge Nozzle-to-safe-end 2.95E-06 9.94E-10 7.99E-08 6.49E-06 

Line) Pipe-to-safe-end 1.75E-06 4.72E-10 4.00E-08 3.64E-06 

 Branch-to-pipe 4.76E-07 1.04E-10 1.04E-08 9.87E-07 

 Elbow-to-pipe 4.60E-08 1.06E-11 1.04E-09 9.66E-08

HPI/ 3-3⁄4 Pipe-to-safe-end 6.56E-04 1.45E-05 1.99E-04 2.53E-03 

NMU Elbow-to-pipe 1.58E-06 3.30E-10 3.43E-08 3.39E-06 

 Pipe-to-valve 1.96E-06 2.36E-10 2.39E-08 2.35E-06

 Pipe-to-pipe 4.55E-06 1.13E-09 1.13E-07 1.11E-05 
D-39

1.00E-06

1.00E-05

1.00E-04

1.00E-03

1.00E-02

> 10% > 20% > 30% > 40% > 50% > 60% > 70% > 80% > 90%

a/t-Ratio

λNTWC [1/Weld.Yr]

NPS12

NPS22

NPS28

Figure D.19 Posterior IGSCC Frequency (Non-Through Wall) 

1.00E-06

1.00E-05

1.00E-04

1.00E-03

1.00E-02

> 10% > 20% > 30% > 40% > 50% > 60% > 70% > 80% > 90%

a/t-Ratio

λNTWC [1/Weld.Yr]

NPS12 - Prior

NPS12 - Posterior

Figure D.20 Prior and Posterior IGSCC Frequency (Non-Through Wall) for NPS12 Welds 

1.00E-06

1.00E-05

1.00E-04

1.00E-03

1.00E-02

> 10% > 20% > 30% > 40% > 50% > 60% > 70% > 80% > 90%

a/t-ratio

λNTWC [1/Weld.Yr]

NPS22 - Prior

NPS22 - Posterior

Figure D.21 Prior and Posterior IGSCC Frequency (Non-Through Wall) for NPS22 Welds 
D-40

1.00E-06

1.00E-05

1.00E-04

1.00E-03

1.00E-02

> 10% > 20% > 30% > 40% > 50% > 60% > 70% > 80% > 90%

a/t-Ratio

λNTWC [1/Weld.Yr]

NPS28 - Prior

NPS28 - Posterior

Figure D.22 Prior and Posterior IGSCC Frequency (Non-Through Wall) for NPS28 Welds 

D.5.2 Conditional Weld Failure Probability 

This section develops a basis for calculating conditional weld failure probabilities. In the model of piping 

reliability, the conditional weld failure probability, pLF, represents the likelihood of a weld flaw propagating 

to a significant structural failure. However, for Code Class 1 piping there is no service experience data 

available to support a direct estimation of pLF. Therefore, the options for calculating the conditional failure 

probability when the service experience data consists of zero events include applications of (1) probabilistic 

fracture mechanics modeling and (2) Bayesian modeling. Since the objective of this Base Case was to 

directly utilize insights from service experience data reviews, the latter approach was selected. Before 

defining the input to the Bayesian modeling, it is useful to organize the available service experience

according to piping classification and severity of observed failures. Figure D.23 shows the conditional pipe 

failure probability as a function of observed through-wall flow rate for reactor coolant pressure boundary 

(RCPB) piping (Code Class 1), Code Class 2 and 3 piping, and ASME B31.1 (non-Code) piping. 
D-41

1.0E-4

1.0E-3

1.0E-2

1.0E-1

1.0E+0

> 0> 1> 5

> 10> 50

> 100 (Cat1)

> 200> 300

> 1500 (Cat2)> 5000 (Cat3)

> 25000 (Cat4)

> 100000 (Cat5)

> 200000 (Cat5+)

Through-wall Leak/Flow Rate [gpm]

Conditional Failure Probability

LWR Class 1 (1632 records)

Class 1 & 2 Small-Diameter (1" or Less;

1471 records) 

Class 2 (1818 records)

Class 3 (1471 records)

ASME B31.1 Piping (2259 records)

Figure D.23 Likelihood of Structural Failure According to Service Experience with 

Light Water Reactor Piping System 

The empirical data that is used to construct the chart in Figure D.23 represents approximately 7,200 recorded 

pipe failure events representing almost 9,000 reactor-years of commercial nuclear power plant operation. It is 

important to note that the chart covers a wide variety of piping systems, from Reactor Coolant Pressure 

Boundary (RCPB) piping, Safety Injection and Recirculation piping, and Auxiliary Cooling piping to ASME 

B31.1 piping (or non-Code piping). Figure D.23 represents our current state-of-knowledge with respect to the 

probability of pipe failure and it provides an upper bound for estimates of conditional pipe failure probability. 

Noteworthy is the fact that only for ASME B31.1 (non-Code) piping systems have significant structural 

failures been observed (that is, failures equivalent to double-ended guillotine breaks). For the other classes of 

piping it is necessary to do data extrapolations beyond the Category 1 through-wall flow rates. In this Base 

Case study, a Bayesian approach is used for this extrapolation and by acknowledging service data insights. 

For the Code Class 1 piping, it is also important to recognize that according to the available service

experience data, through-wall flow rates greater than 38 lpm (10 gpm) have been primarily attributed to 

failure of small bore piping. It is unlikely that the structural integrity of the Class 1 piping is equal to or less 

that that of Code Class 2 and 3 piping for which Category 1 LOCA events have been observed. 

The Beta Distribution has some convenient and useful properties for use in Bayes’ updating. The analysis 

starts by defining a prior distribution that represents the analyst’s understanding of piping performance given 

the presence of some sort of degraded condition. The prior distribution is defined by selecting an appropriate 

set of initial values for parameters A(α) and B(β), denoted as APrior and BPrior. Then, when looking at the 

relevant service experience data, if there are “N” structural failures of a certain magnitude and “M” successes 
D-42

(or degraded conditions that were repaired before progressing to a structural failure), the Bayes’ updated, or 

posterior distribution is also a Beta Distribution with the following parameters: 

APosterior = APrior + N 

 BPosterior = BPrior + M 

The above explains how the Beta Distribution can be used to estimate conditional weld failure probabilities. 

The challenge is to justify the selected parameters when the evidence is zero structural failures. Certainly, it 

can be argued the ASME B31.1 service experience data represents a very conservative upper bound for the 

conditional weld failure probability. 

Selecting a well justified set of “A” and “B” parameters is not a trivial task. One basic ground rule should be 

for the “weight” of the field experience data to determine the shape of the posterior Beta Distribution. 

However, many different parameter combinations will produce the same predicted mean value. Where very 

little evidence is available about the parameters, constrained non-informative priors may be selected. For such 

a case, one can say that the “A” parameter has to be a small number. 

In this Base Case study, the prior “A” and “B” parameters are defined by first deriving a constraint for the 

prior mean value of the conditional failure probability and then fixing the “A” parameter at 1.0 for stress 

corrosion cracking and 2.0 for thermal fatigue to account for the fact that according to available service 

experience data, thermal fatigue cracks propagate in the through-wall direction considerably faster than flaws 

caused by stress corrosion cracking. The process for developing conditional failure probabilities starts by 

deriving a point-estimate of pL|F for small-diameter piping given susceptibility to stress corrosion cracking 

(SCC). This point estimate is based on Jeffrey’s non-informative prior and service experience data. There 

have been 42 through-wall flaws and zero large leaks in small-bore BWR piping. This gives a point estimate 

of 1.2×10-2, which is used as a “fix point” for determining conditional weld failure probabilities for other pipe

sizes. The relationship between pipe size (diameter and wall thickness) and the conditional failure probability 

is assumed to follow a power law of the form: 

pL|F = a × DNb (D.8) 

Where, “DN” is the nominal pipe size in [mm]. Decreasing trends correspond to negative values of b. 

Parameters a and b are determined for pL|F = 1.2×10-2 and DN = 25. Point estimates of pL|F for other pipe sizes 

are derived using the power law for conditional failure probabilities and assuming that the general shape of 

the curve is similar to that of piping susceptible to vibratory fatigue for which pL|F = 2.5/DN. Next the 

predicted conditional weld failure probability using the power law approach is used to determine the Beta 

Distribution parameter “B.” For Class 1 piping, engineering judgment, as portrayed by Figure D.23 is used to 

assign values to the prior Beta Distribution parameters. The proposed Beta Distribution posterior parameters 

for this Base Case study are summarized in Table D.14. 

Table D.14 Proposed Beta Parameters for Code Class 1 Piping 

Degradation Pipe Size 

Mechanism DN NPS 

Parameter B in 

Beta Posterior (“Large Leak”) 

300 12 1,262 (APost = 1; M = 0) 

550 22 1,496 (APost = 1; M = 0) SCC 

(APrior = 1) 

700 28 1,700 (APost = 1; M = 0) 

90 3-3⁄4 227 (APost Thermal Fatigue = 2; M = 0) 

(APrior = 2) 350 14 592 (APost = 2; M = 0) 
D-43

D.5.3 Conditional Failure Probability and Flow Rate

The conditional failure probabilities derived in the previous section are assumed applicable to Cat0 LOCA. It 

is furthermore assumed that for a significant primary piping breach to occur there has to be a through-wall 

flaw coinciding with a plant operational mode change or an unusual or severe loading condition such that the 

leakage exceeds a Cat0 LOCA. The service data collection (e.g., PIPExp) includes numerous examples where 

pressure pulses or spikes caused by changing flow conditions following a plant operational mode change

have resulted in non-active leaks11 becoming active leaks. The physics of such transitions from non-active to 

active leaks are complex and location-dependent (e.g., function of flaw size and pipe stresses). Some 

published work exists on the correlation between crack propagation and plant transient history [D.25]. Using 

available empirical data, the uncertainties in such crack growth assessments are considerable, however. 

In this analysis a simple parametric approach is applied to the estimation of weighted conditional failure 

probabilities (CL) of a pressure boundary breach that exceeds a Cat0 flow rate threshold value. This approach 

is described through the event tree in Figure D.24. An undetected, or detected but monitored through-wall is 

exposed to a pressure pulse or unusual loading condition before a decision to perform manual, controlled 

reactor shutdown. The pressure pulse or unusual loading condition is characterized as a subjectively defined 

probability distribution. 

Through-wall

Defect

(Cat0)

Unusual

Load

Unstable

Crack

Growth

Extreme

High

Very High

Severe

Moderate

Moderate-to-high

Cat0

Cat3

Cat1

Cat6

Cat5

Cat4

Cat2

LOCA

Category

Cat0,Cat1

Cat0,Cat1

Cat0,Cat1

Cat0,Cat1

Cat0,Cat1

Figure D.24 Event Tree for Definition of LOCA Categories 

In the cases of “moderate-to-high” to “extreme”, the term “unusual” implies a loading condition beyond that 

resulting from anticipated transients including manual and automatic reactor/turbine trips. The conditional 

probability of an unusual or severe loading condition is described by five sets of subjective 3-bin discrete and 

overlapping probability distributions as summarized in Table D.15. These DPDs are combined with the weld 

failure rate distributions and conditional weld failure probability distributions by using a Monte Carlo merge 

technique. 

 

11 The term ‘non-active leak’ is taken to mean a through-wall flaw without visible leakage or with a small, detectable leakage that stays 

relatively constant over time. 
D-44

Table D.15 Probability of LOCA Given Severe Overloading 

Flow Rate (ν) Intervals DPD for Severe Loading

Category [gpm] CL-High CL-Med CL-Low pHigh pMed pLow 

0 10 < ν ≤ 100 N/A N/A N/A N/A N/A N/A 

1 100 < ν ≤ 1500 .80 .50 .20 .2 .6 .2 

2 1500 < ν ≤ 5000 .32 .20 .08 .2 .6 .2 

3 5000 < ν ≤ 25,000 .13 .08 .03 .2 .6 .2 

4 25,000 < ν ≤ 100,000 .05 .03 .01 .2 .6 .2 

5 100,000 < ν ≤ 500,000 .02 .01 .005 .2 .6 .2 

6 ν > 500,000 .01 .005 .002 .2 .6 .2 

Service data on water hammer events provides a justification for the chosen DPDs. From PIPExp, a point 

estimate for CL-WH-Cat6 is approximately 4.9E-03, which is based on two events involving severe overloading 

(including plastic deformation) of a pipe section in 411 recorded water hammer events. This is taken as a best 

estimate CL-value for calculating a Cat6 LOCA. Figure D.24 includes the rules for how the DPDs are applied 

to the LOCA frequency calculation. The Cat0 and Cat1 LOCAs include contributions from each loading 

condition associated with Cat2 or larger pressure boundary breach. In other words, the calculation accounts 

for the possibility that an ‘unusual’ loading condition may not result in a global or catastrophic pressure 

boundary breach. Given a through-wall flaw and severe overload, Figure D.25 shows the conditional failure 

probability as a function of pipe size. 

1.00E-05

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1 3 5 7 9 11 13 15

NPS [inch]

Conditional Failure Probability

Cat0|Through-wall Flaw

LOCA @ Upper Flow Rate Limit

Power (LOCA @ Upper Flow Rate Limit)

Figure D.25 Conditional Probability of Weld Failure Given Through-Wall Flaw and Severe 

Overloading 
D-45

D.6 BWR and PWR Base Case LOCA Frequencies 

The Base Case LOCA frequency models are based on three Excel files entitled “PlantBWelds” (BWR Base 

Case), “PlantA.aWelds” (PWR RC-HL and Surge Line), and “PlantA.bWelds” (PWR HPI/NMU). These 

Excel files are part of the plant design information supplied by members of the Expert Elicitation Panel. 

Adding information on weld type by location and weld failure parameters to respective Excel files provides 

the basis for calculating LOCA frequencies including uncertainty distributions. 

D.6.1 LOCA Frequency Models 

For the BWR Base Case, the original “PlantBWelds” includes two spreadsheets, one for the feedwater system 

and one for the reactor recirculation system. In a first step to create a LOCA frequency model, each 

spreadsheet was split into two, one for Loop A and one for Loop B of respective system. Next a new column 

was added to each new spreadsheet to include information on weld type by location. A review of isometric 

drawings provided the input to the new columns. 

The statistical information that is summarized in Section D.5 of this report is included in five separate 

spreadsheets of the modified Excel file. The posterior weld failure rates are included on Tab “Parameters,” 

each parameter assigned a unique variable name. The calculation of LOCA frequency, including Monte Carlo 

merge operations are performed on Tabs “Intermediate FW_A”, “Intermediate FW_B”, “Intermediate 

RR_A”, and “Intermediate RR_B.” These intermediate calculation sheets are linked to the design information 

for FW Loop A, FW Loop B, RR Loop A and RR Loop B, respectively. Using the variable names as defined 

on Tab “Parameters,” each weld is assigned an appropriate failure rate (including uncertainty distribution as 

defined using a Crystal Ball “assumption”). Finally, an integrated calculation of LOCA frequency by leak 

threshold category (0 through 5) is performed on a separate spreadsheet, which is linked to the intermediate 

calculations. Each integrated LOCA frequency calculation is defined as a Crystal Ball “forecast.” 

For the PWR Base Case, the original “PlantAWelds” includes a single spreadsheet with the ASME XI Class

1 Category B-F/B-J welds. In a first step to create a LOCA frequency model, each spreadsheet was split into 

two, one for the RC-HL and one for the pressurizer surge line. A third spreadsheet with the HPI/NMU weld 

listed was added to form a new Excel-file corresponding to the PWR Base Case LOCA frequency model. 

D.6.2 BWR Base Case LOCA Frequencies at T = 25 

Summarized in Table D.16 are the LOCA frequency uncertainty distributions that are derived for BWR Base 

Case Cat0 through Cat6 LOCA. The results are representative of Plant B after 25 years of operation (T = 25). 
D-46

Table D.16 Plant B LOCA Frequencies at T = 25 Years 

LOCA Uncertainty Distribution

System Category Mean 5%-tile 50%-tile 95%-tile

RR Loop A Cat0 2.06E-06 9.72E-08 1.38E-06 6.28E-06 

NPS12 Cat1 1.33E-06 6.29E-08 8.95E-07 4.09E-06 

Cat2 1.44E-07 6.04E-09 8.66E-08 4.73E-07 

Cat3 5.81E-08 2.34E-09 3.51E-08 1.91E-07 

Cat4 2.31E-08 9.08E-10 1.39E-08 7.54E-08 

Cat5 N/A N/A N/A N/A 

RR Loop B Cat0 2.07E-06 1.00E-07 1.36E-06 6.41E-06 

NPS12 Cat1 1.34E-06 6.45E-08 8.81E-07 4.15E-06 

Cat2 1.44E-07 5.85E-09 8.65E-08 4.72E-07 

Cat3 5.84E-08 2.26E-09 3.47E-08 1.95E-07 

Cat4 2.31E-08 9.15E-10 1.40E-08 7.48E-08 

Cat5 N/A N/A N/A N/A 

RR Loop A Cat0 1.23E-06 4.88E-08 7.30E-07 4.07E-06 

NPS22 Cat1 8.58E-07 3.37E-08 5.12E-07 2.89E-06 

Cat2 7.36E-07 2.42E-09 3.94E-08 2.57E-07 

Cat3 2.95E-08 1.01E-09 1.60E-08 1.01E-07 

Cat4 1.16E-08 3.94E-10 6.33E-09 4.04E-08 

Cat5 4.72E-09 1.57E-10 2.52E-09 1.63E-08 

RR Loop B Cat0 1.24E-06 4.97E-08 7.26E-07 4.06E-06 

NPS22 Cat1 8.66E-07 3.39E-08 5.10E-07 2.88E-06 

 Cat2 7.37E-08 2.48E-09 3.95E-08 2.59E-07 

 Cat3 2.91E-08 9.77E-10 1.59E-08 1.01E-07 

 Cat4 1.18E-08 3.86E-10 6.36E-09 3.97E-08 

 Cat5 4.73E-09 1.56E-10 2.55E-09 1.67E-08 

RR Loop A Cat0 2.73E-06 1.31E-07 1.80E-06 8.41E-06 

NPS28 Cat1 1.91E-06 9.05E-08 1.24E-06 5.91E-06 

 Cat2 1.64E-07 6.27E-09 9.69E-08 5.38E-07 

 Cat3 6.54E-08 2.49E-09 3.92E-08 2.16E-07 

 Cat4 2.61E-08 1.03E-09 1.55E-08 8.65E-08 

 Cat5 1.05E-08 4.24E-10 6.20E-09 3.50E-08 

RR Loop B Cat0 2.76E-06 1.35E-07 1.82E-06 8.44E-06 

NPS28 Cat1 1.93E-06 9.31E-08 1.27E-06 5.92E-06 

 Cat2 1.65E-07 6.32E-09 9.73E-08 5.42E-07 

 Cat3 6.62E-08 2.57E-09 3.93E-08 2.18E-07 

 Cat4 2.64E-08 1.10E-09 1.56E-08 8.74E-08 

 Cat5 1.06E-08 4.40E-10 6.33E-09 3.40E-08 

FW Loop A Cat0 6.69E-07 2.78E-08 4.05E-07 2.20E-06 

NPS12 Cat1 4.34E-07 1.80E-08 2.61E-07 1.42E-06 

 Cat2 4.75E-08 1.61E-09 2.58E-08 1.61E-07 

 Cat3 1.89E-08 6.41E-10 1.03E-08 6.49E-08 

 Cat4 7.53E-09 2.62E-10 4.07E-09 2.53E-08 

 Cat5 N/A N/A N/A N/A 

FW Loop B Cat0 6.83E-07 2.89E-08 4.16E-07 2.24E-06 

NPS12 Cat1 4.44E-07 1.90E-08 2.68E-07 1.45E-06 

 Cat2 4.86E-08 1.70E-09 2.64E-08 1.67E-07 

 Cat3 1.93E-08 6.58E-10 1.06E-08 6.71E-08 

 Cat4 7.75E-09 2.80E-10 4.20E-09 2.62E-08 

 Cat5 N/A N/A N/A N/A 

FW Loop B Cat0 6.02E-08 1.52E-09 2.61E-08 2.26E-07 

NPS14 Cat1 4.19E-08 1.06E-09 1.84E-08 1.55E-07 

 Cat2 3.59E-09 7.64E-11 1.42E-09 1.37E-08 

 Cat3 1.45E-09 3.12E-11 5.69E-10 5.51E-09 
D-47

Table D.16 Plant B LOCA Frequencies at T = 25 Years 

LOCA Uncertainty Distribution

System Category Mean 5%-tile 50%-tile 95%-tile

 Cat4 5.80E-10 1.19E-11 2.30E-10 2.21E-09 

 Cat5 2.32E-10 4.74E-12 9.11E-11 8.83E-10 

FW Loop A Cat0 8.43E-07 3.47E-08 5.08E-07 2.74E-06 

NPS20 Cat1 5.92E-07 2.37E-08 3.55E-07 1.95E-06 

 Cat2 5.08E-08 1.75E-09 2.75E-08 1.74E-07 

 Cat3 2.03E-08 7.15E-10 1.10E-08 6.83E-08 

 Cat4 8.10E-09 2.77E-10 4.33E-09 2.81E-08 

 Cat5 3.26E-09 1.12E-10 1.74E-09 1.13E-08 

FW Loop B Cat0 1.00E-06 4.40E-08 6.13E-07 3.24E-06 

NPS20 Cat1 7.014-07 3.01E-08 4.28E-07 2.12E-06 

 Cat2 6.03E-08 1.03E-09 3.32E-08 2.04E-07 

 Cat3 2.41E-08 8.45E-10 1.33E-08 8.00E-08 

 Cat4 9.64E-09 3.44E-10 5.27E-09 3.32E-09 

 Cat5 3.84E-09 1.38E-10 2.14E-09 1.31E-08 

RR Total Cat0 1.21E-05 3.02E-06 1.03E-05 2.70E-05 

Loops A & B Cat1 8.24E-06 2.02E-06 6.98E-06 1.86E-05 

 Cat2 7.64E-07 1.40E-07 6.07E-07 1.92E-06 

 Cat3 3.07E-07 5.43E-08 2.44E-07 7.79E-07 

 Cat4 1.22E-07 2.22E-08 9.73E-08 3.06E-07 

 Cat5 3.05E-08 3.59E-09 2.19E-08 8.52E-08 

FW Total Cat0 3.26E-06 5.55E-07 2.60E-06 8.13E-06 

Loops A & B Cat1 2.21E-06 2.72E-07 1.75E-06 5.56E-06 

 Cat2 2.10E-07 2.70E-08 1.54E-07 5.85E-07 

 Cat3 8.40E-08 1.10E-08 6.18E-08 2.30E-07 

 Cat4 3.36E-08 4.40E-09 2.47E-08 9.35E-08 

 Cat5 7.33E-09 3.95E-10 4.33E-09 2.44E-08 

RR + FW Total Cat0 1.53E-05 5.24E-06 1.37E-05 3.10E-05 

 Cat1 1.05E-05 3.51E-06 9.30E-06 2.14E-05 

 Cat2 9.75E-07 2.24E-07 8.24E-07 2.26E-06 

 Cat3 3.90E-07 9.00E-08 3.32E-07 9.05E-07 

 Cat4 1.56E-07 3.64E-08 1.31E-07 3.57E-07 

 Cat5 3.78E-08 6.43E-09 2.93E-08 9.73E-08 

D.6.3 PWR Base Case LOCA Frequencies at T = 25 

The PWR Base Case Cat0 through Cat6 LOCA frequencies including uncertainty distributions, are 

summarized in Table D.17. These results are representative of Plant A.a/A.b after 25 years of operation (T = 

25). 
D-48

Table D.17 Plant A.a/A.b LOCA Frequencies at T = 25 Years 

LOCA Uncertainty Distribution

System Category Mean 5%-tile 50%-tile 95%-tile

RC Hot Leg Cat0 8.94E-07 4.84E-09 1.27E-07 2.88E-06

(3-of-3); Plant A.a Cat1 6.65E-07 3.55E-09 9.39E-08 2.14E-06 

 Cat2 4.87E-08 2.10E-10 6.15E-09 1.49E-07 

 Cat3 1.83E-08 8.33E-11 2.42E-09 5.95E-08 

 Cat4 6.99E-09 3.03E-11 8.93E-10 2.21E-08 

 Cat5 2.55E-09 1.16E-11 3.29E-10 8.29E-09 

 Cat6 1.26E-09 5.44E-12 1.58E-10 4.04E-09 

RC Surge Line Cat0 1.44E-07 2.65E-09 2.98E-08 5.02E-07 

Plant A.a Cat1 1.14E-07 2.13E-09 2.36E-08 3.94E-07 

 Cat2 9.60E-09 1.48E-10 1.88E-09 3.46E-08 

 Cat3 3.84E-09 5.78E-11 3.50E-10 1.35E-08 

 Cat4 1.44E-09 2.01E-11 2.77E-10 5.06E-09 

 Cat5 5.31E-10 8.23E-12 1.01E-10 1.87E-09 

 Cat6 N/A N/A N/A N/A 

HPI/NMU (2-of-2) Cat0 2.72E-05 4.64E-07 6.90E-06 1.07E-04 

Plant A.b Cat1 1.60E-05 2.62E-07 3.93E-06 6.09E-05 

 Cat2 2.33E-06 3.30E-08 5.40E-07 9.02E-06 

 Cat3 9.22E-07 1.28E-08 2.14E-07 3.59E-06 

 Cat4 N/A N/A N/A N/A 

 Cat5 N/A N/A N/A N/A 

 Cat6 N/A N/A N/A N/A 

D.6.4 Time-Dependency of LOCA Frequency Results 

For respective Base Case Plant, the LOCA frequencies are determined for three time periods: T= = 25 years 

after plant startup (corresponding to today’s state-of-knowledge), T = 40 years after plant startup 

(corresponding to original design life), and T = 60 years after plant startup (corresponding to end-of-life 

extension). The time-dependent analysis is performed in two different ways. First a ‘prospective analysis’ is 

performed based on a Markov model of piping reliability (Figure D.15). Second, a ‘retrospective analysis’ is 

performed by using Bayesian statistics. 

D.6.4.1 Use of Markov Model to Determine Time-Dependency - According to the Markov model 

diagram in Figure D.15, a piping component can be in four mutually exclusive states: S (= Success), C (= 

Cracked), F (= Leaking, non-active leakage, or active leakage with leak rate within Technical Specification 

Limit) or L (= Leaking, with leak rate in excess of Technical Specification Limit). The time-dependent

probability that a piping component is in each state S, C, F, or L is described by a differential equation. Under 

the assumption that all the state transition rates are constant the Markov model equations will consist of a set 

of coupled linear differential equations with constant coefficients. The reliability term needed to represent 

LOCA frequency is the system failure rate or hazard rate h{t}, which is time-dependent. The hazard rate is 

defined as: 

h{t} = (1/(1- L{t})) × dL{t}/dt (D.9) 

Where: 

1 – L{t} = S{t} + C{t} + F{t} (D.10) 
D-49

The hazard rate is a function of time and the parameters of the Markov model; h{t} is the time-dependent 

frequency of pipe rupture. Reference [D.14] provides solutions to the Markov model and derives an 

expression for h{t} as a function of the six parameters associated with the 4-state Markov model: An 

occurrence rate for detectable flaws (φ), a failure rate for leaks given the existence of a flaw (λF), two rupture 

frequencies including one from the initial state of a flaw (ρF) and one from the initial state of a leak (ρL), a 

repair rate for detectable flaws (ω), and a repair rate for leaks (μ). The latter two parameters dealing with 

repair are further developed by the following simple models. 

( ) FI R

FI FD

T T

P P

+

ω = (D.11) 

Where: 

PFI = probability that a piping element with a flaw will be inspected per inspection interval. 

This parameter has a value of 0 if it is not in the inspection program and 1 if it is in the 

inspection program. For the inspected elements, a value of 1 is used for any ISI inspection case 

and 0 for the case of no ISI. The element may be selected for inspection directly by being 

included in the sections sampled for ISI inspection, or indirectly by having a rule such that if 

degradation is detected anywhere in the system, the search will be expanded to include 

examination of that element. 

PFD = probability that a flaw will be detected given this element is inspected. This is the 

reliability of the inspection program and is equivalent to the term used by NDE experts, 

“Probability of detection (POD).” This probability is conditioned on the occurrence of one or 

more detectable flaws in the segment according to the assumptions of the model. Also note 

that 

TFI = mean time between inspections for flaws, (inspection interval). 

TR = mean time to repair once detected. Depending on the location of the weld to be repaired, 

the actual weld repair could take on the order of several days to much more than a week. 

Accounting for time to prepare for repair, NDE, root cause evaluation, etc., the total outage 

time attributed to the repair of a Class 1 weld is on the order of 1 month or more. However, 

since this term is always combined with TFI, and TFI could be 10 years, in practice the results 

are insensitive to assumptions regarding TR

Similarly, estimates of the repair rate for leaks can be estimated according to: 

μ =

+

P

T T

LD

LI R ( )

 (D.12) 

Where: 

PLD = probability that the leak in the element will be detected per leak inspection or detection 

period 

TLI = mean time between inspections for leaks. For RCPB piping the time interval between 

leaks can be essentially instantaneous if the leak is picked up by radiation alarms, to as long as 

the time period between leak tests performed on the system. 
D-50

TR = as defined above but for full power applications, this time should be the minimum of the 

actual repair time and the time associated with cooldown to enable repair and any waiting time 

for replacement piping. 

A summary of the root input parameters of the Markov model and the general strategy for estimation of each 

parameter is presented in Table D.18. 

Table D.18 Four-State Markov Model Root Input Parameters 

Parameter Assumed or Estimated Value Basis 

ω 2.1 x 10-2/year 

{=(.25) × (.90)/(10+(200/8760))} 

Element assumed to have a 25% chance of being 

inspected for flaws every 10 years with a 90% 

detection probability. In the given example 

detected flaws will be repaired in 200 hours 

μ 7.92 x 10-1/ year 

{=(.90) × (.90)/(1+(200/8760))} 

Element is assumed to have a 90% chance of being 

inspected for leaks once a year with a 90% leak 

detection probability 

ρC Table D.13, D.14 and D.15 The basis is developed in Sections D.4 and D.5. 

λC Table D.13 and D.14 The basis is developed in Sections D.4 and D.5. 

ρF 2.0 x 10-2/year 

If the element is already leaking, the conditional 

frequency of ruptures is assumed to be determined 

by the frequency of severe overloading events; the 

given value is equal to the frequency of severe 

water hammer (from PIPExp database). 

φ Variable 

(for IGSCC φ = 7.58 × (λC + ρC)) 

The occurrence rate of a flaw is estimated from 

service data. As an example, IGSCC in the BWR 

operating environment will create ca. 7.58 flaws 

for every through-wall leak that is observed. 

PFI 1 or 0 Probability per inspection interval that the pipe 

element will be included in the inspection 

program. 

PFD Variable 

(see text above for details) 

Probability per inspection interval that an existing 

flaw will be detected. A chosen estimate is based 

on NDE reliability performance demonstration 

results and difficulty and accessibility of 

inspection for particular weld. 

PLD

Variable 

(0 – no leak detection to 0.9 for leak 

detection using current 

methods/technology) 

Probability per detection interval that an existing

leak will be detected. Estimate based on system, 

presence and type of leak detection system, and 

locations and accessibility. 

TFI 10 years (per ASME XI) Flaw inspection interval, mean time between inservice inspections. 

TLD Variable 

(1.5 – once per refueling outage / 

1.92E-2 – weekly / 9.13E-4 – each 

shift) 

Leak detection interval, mean time between leak 

detections. Estimate based on method of leak 

detection; ranges from immediate/ continuous to 

frequency of routine inspections for leaks (incl. 

hydrostatic pressure testing). 

TR Variable 

(see text above for details) 

Mean time to repair the affected piping element 

given detection of a critical flaw or leak. Estimate 

of time to tag out, isolate, prepare, repair, leak test 

and tag into service. 
D-51

In addition to generating a time-dependent LOCA frequency, the Markov model provides a basis for 

investigating the sensitivity of LOCA frequency to different in-service inspection and leak detection 

strategies. The Markov model determines the inspection effectiveness factor, I, which is the ratio of the 

LOCA frequency with credit for inspections to that given no credit for inspections: 

{ }

{ }''

25

25

h noinsp

h inspprog j

I

j = (D.13) 

Where: 

h25{inspprog ‘j’} = hazard rate at T = 25 given inspection strategy ‘j.’ 

h25{noinsp} = hazard rate given no inspections. 

The solutions to the Markov model for time dependent hazard rates are developed in terms of closed form 

analytical solutions using an Excel spreadsheet. In this study the time-dependent LOCA frequencies are

determined for twelve cases that are defined by varying the following parameters (Table D.19): 

• Whether or not the piping segment is subjected to any ISI program; 

• The extent of the ISI program (‘Caused-Based’ vs. ‘Extensive’, all encompassing); 

• The inspection interval of the ISI program; 

• Type and frequency of leak detection. The different leak detection methods include primary system 

mass balance calculations, visual observation (through video monitor), (PWR) containment sump 

level and flow rate monitors, airborne particulate radioactivity and gaseous radioactivity monitors, 

and different main control room monitors for primary system temperature, pressure, etc. 

Table D.19 Inspection Cases Evaluated for Selected Pipe Segments 

 In-Service Inspection Strategy 

Leak Inspection 

Strategy None Cause-Based [PFD = 0.50] 

Comprehensive 

[PFD = 0.90] 

None Case 1 Case 5 Case 9 

Refueling Cycle (Hydro Test) Case 2 Case 6 Case 10 

Weekly Case 3 Case 7 Case 11 

8 Hour Shift Case 4 Case 8 Case 12 

The time-dependent LOCA frequencies associated with the five Base Cases are summarized in Figures D.26

through D.40. Figure D.26 is assumed to be representative of Base Case 1; the LOCA frequency at T = 25

years is equal to the calculated point estimate of 8.24E-06 per reactor-year under an assumption of “causedbased” ISI with POD = 0.5 and leak detection (e.g., hydrostatic pressure testing prior to exiting a refueling 

outage). This assumption is applied to the other base cases as well (Figures D.29, D.32, D.35, and D.38). 

It is noted that the service data input to the calculation is associated with piping that has been subjected to 

different inspection strategies. In some cases flaws have been detected fortuitously and in other cases the flaw 

detection has resulted from augmented IGSCC inspection programs. The results in Figure D.26 are based on 

an assumed ‘cause-based’ inspection strategy whereby the inspection sample is determined by an initial

discovery of a flaw. If a flaw is found, the inspection is immediately expanded to cover other similar

locations. The combination of inspection sample and rules for expanded search for flaws are sufficient to 

result in an average probability of detection (POD) of 0.50. The analysis also considers what in this study is 
D-52

termed “comprehensive” ISI, which implies 100% ISI coverage using state-of-the-art NDE technology. Such 

a program is assumed to result in an average probability of detection of 0.90.

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

0 10 20 30 40 50 60 70

Plant Age [Years]

BWR-1 Cat1 LOCA Frequency [1/Reactor.Yr]

Hydro Test

Weekly

Shift

None

Figure D.26 Time-Dependent BWR-1 Cat 1 LOCA Frequency Given ‘Cause-Based’ ISI 

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

0 10 20 30 40 50 60 70

Plant Age [Years]

BWR-1 Cat1 LOCA Frequency [1/Reactor.Yr]

None

Hydro Test

Weekly

Shift

Figure D.27 Time-Dependent BWR-1 Cat 1 LOCA Frequency Assuming no ISI 
D-53

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

0 10 20 30 40 50 60 70

Plant Age [Years]

BWR-1 Cat1 LOCA Frequency [1/Reactor.Yr.]

None

Hydro Test

Weekly

Shift

Figure D.28 Time-Dependent BWR-1 Cat 1 LOCA Frequency Given ‘Comprehensive ISI’ 

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

0 10 20 30 40 50 60 70

Plant Age [Years]

BWR-2 Cat1 LOCA Frequency [1/Reactor.Yr]

Hydro Test

Weekly

Shift

None

Figure D.29 Time-Dependent BWR-2 Cat 1 LOCA Frequency Given ‘Cause-Based’ ISI 
D-54

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

0 10 20 30 40 50 60 70

Plant Age [Years]

BWR-2 Cat1 LOCA Frequency [1/Reactor.Yr]

None

Hydro Test

Weekly

Shift

Figure D.30 Time-Dependent BWR-2 Cat 1 LOCA Frequency Assuming no ISI 

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

0 10 20 30 40 50 60 70

Plant Age [Years]

BWR-2 Cat1 LOCA Frequency [1/Reactor.Yr.]

None

Hydro Test

Weekly

Shift

Figure D.31 Time-Dependent BWR-2 Cat 1 LOCA Frequency Given ‘Comprehensive ISI’ 
D-55

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

0 10 20 30 40 50 60 70

Plant Age [Years]

PWR-1 Cat1 LOCA Frequency [1/Reactor.Yr]

Hydro Test

Weekly

Shift

None

Figure D.32 Time-Dependent PWR-1 Cat 1 LOCA Frequency Given ‘Cause-Based’ ISI 

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

0 10 20 30 40 50 60 70

Plant Age [Years]

PWR-1 Cat1 LOCA Frequency [1/Reactor.Yr]

None

Hydro Test

Weekly

Shift

Figure D.33 Time-Dependent PWR-1 Cat 1 LOCA Frequency Assuming no ISI 
D-56

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

0 10 20 30 40 50 60 70

Plant Age [Years]

PWR-1 Cat1 LOCA Frequency [1/Reactor.Yr.]

None

Hydro Test

Weekly

Shift

Figure D.34 Time-Dependent PWR-1 Cat 1 LOCA Frequency Given ‘Comprehensive ISI’ 

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

0 10 20 30 40 50 60 70

Plant Age [Years]

PWR-2 Cat1 LOCA Frequency [1/Reactor.Yr]

Hydro Test

Weekly

Shift

None

Figure D.35 Time-Dependent PWR-2 Cat 1 LOCA Frequency Given ‘Cause-Based’ ISI 
D-57

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

0 10 20 30 40 50 60 70

Plant Age [Years]

PWR-2 Cat1 LOCA Frequency [1/Reactor.Yr]

None

Hydro Test

Weekly

Shift

Figure D.36 Time-Dependent PWR-2 Cat 1 LOCA Frequency Assuming no ISI 

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

0 10 20 30 40 50 60 70

Plant Age [Years]

PWR-2 Cat1 LOCA Frequency [1/Reactor.Yr.]

None

Hydro Test

Weekly

Shift

Figure D.37 Time-Dependent PWR-2 Cat 1 LOCA Frequency Given ‘Comprehensive ISI’ 
D-58

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

1.0E-02

0 10 20 30 40 50 60 70

Plant Age [Years]

PWR-3 Cat1 LOCA Frequency [1/Reactor.Yr]

Hydro Test

Weekly

Shift

None

Figure D.38 Time-Dependent PWR-3 Cat 1 LOCA Frequency Given ‘Cause-Based’ ISI 

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

1.0E-02

0 10 20 30 40 50 60 70

Plant Age [Years]

PWR-3 Cat1 LOCA Frequency [1/Reactor.Yr]

None

Hydro Test

Weekly

Shift

Figure D.39 Time-Dependent PWR-3 Cat 1 LOCA Frequency Assuming no ISI 
D-59

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

1.0E-02

0 10 20 30 40 50 60 70

Plant Age [Years]

PER-3 Cat1 LOCA Frequency [1/Reactor.Yr.]

None

Hydro Test

Weekly

Shift

Figure D.40 Time-Dependent PWR-3 Cat 1 LOCA Frequency Given ‘Comprehensive ISI’ 

D.6.4.2 Speculative LOCA Frequency at T = 40 & T = 60 - A retrospective evaluation is performed 

through a Bayesian update process whereby the exposure term in Equation 4.1 is modified to account for the 

longer exposure time. The analysis is performed by assuming that the service history at T = 40 and T = 60 

years is known; zero (0) weld failures during the intervals ∆T = 15 years (T40 to T25) and ∆T = 35 years 

(T60 to T25). This is a purely speculative assumption implying that the ISI/NDE technologies and other

piping reliability management programs remain at least as effective as at the present and that no unexpected 

material aging occurs. The extrapolated LOCA frequencies are summarized in Table D.20. Under the given

assumptions the LOCA frequency would be expected to decrease with time. 
D-60

Table D.20 Base Case LOCA Frequency Results (T = 25, 40 & 60 Years) 

LOCA Frequency – Statistical Mean [per Reactor-year] 

Base Flow Rate Interval [gpm] 

Case 100 < ν ≤ 1500 1500 < ν ≤ 5000 5000 < ν ≤ 25,000 

25,000 < ν ≤

100,000 

100,000 < ν ≤ 500,000 

BWR-1, T = 25 8.24E-06 7.64E-07 3.07E-07 1.22E-07 3.05E-08 

BWR-1; T = 40 2.67E-06 2.29E-07 9.14E-08 3.64E-08 1.45E-08 

BWR-1; T = 60 2.44E-06 2.08E-07 8.38E-08 3.34E-08 1.34E-08 

BWR-2, T = 25 2.21E-06 2.11E-07 8.40E-08 3.36E-08 7.33E-09 

BWR-2, T = 40 2.07E-06 2.03E-07 8.05E-08 3.13E-08 6.61E-09 

BWR-2, T = 60 1.87E-06 1.85E-07 7.35E-08 2.97E-08 6.09E-09 

PWR-1, T = 25 6.65E-07 4.87E-08 1.83E-08 6.99E-09 2.55E-09 

PWR-1, T = 40 2.14E-07 1.49E-08 6.10E-09 2.24E-09 8.14E-10 

PWR-1, T = 60 1.19E-07 8.34E-09 3.38E-09 1.26E-09 4.62E-10 

PWR-2, T = 25 1.14E-07 9.60E-09 3.84E-09 1.44E-09 5.31E-10 

PWR-2, T = 40 1.07E-07 9.22E-09 3.68E-09 1.34E-09 4.79E-10 

PWR-2, T = 60 9.67E-08 8.31E-09 3.36E-09 1.27E-09 4.41E-10 

PWR-3, T = 25 1.60E-05 2.33E-06 9.22E-07 N/A N/A 

PWR-3, T = 40 1.08E-05 1.58E-06 6.31E-07 N/A N/A 

PWR-3, T = 60 8.23E-06 1.20E-06 4.81E-07 N/A N/A 

Note 1: PWR-1 in this table accounts for 3-of-3 hot legs. 

Note 2: PWR-3 in this table accounts for 2-of-2 HPI/NMU lines. 

D.6.5 Influence of Service Data on LOCA Frequency 

The LOCA frequencies in this Base Case Report are derived from service data on Code Class 1 piping. In this 

section we investigate how the LOCA frequencies relate to two data issues: 1) completeness of the pipe

failure data collection, and 2) data interpretations. The former remains an ever-present issue in probabilistic 

safety assessment. Completeness is addressed by having in place an active and rigorous data collection

process (c.f. Appendix A). Two sensitivity cases (SC:s) are defined to demonstrate how changes in the input 

to the failure rate calculations affect the estimated LOCA frequency. The sensitivity cases are defined as: 

1. SC1: A small leak (≤ T.S. limit for unidentified RCPB-leakage) is assumed to have occurred in a 

pipe-to-safe-end weld in a BWR NPS28 reactor recirculation pipe during the time period 1988 – 

2002. This evidence is used to modify the posterior weld failure rates. 

2. SC2: This sensitivity case is concerned with an assumed large leak (= Cat0 LOCA) in a NPS28 

BWR reactor recirculation pipe. Again, the large leak is assumed to have occurred in the time period 

1988 – 2002. This evidence is used to modify the posterior weld failure rates and the conditional 

failure probability. 

The results of the sensitivity analysis are summarized in Table D.21. These sensitivity cases are hypothetical 

in that they do not account for effects on piping reliability by the anticipated industry and regulatory actions 

that invariably would arise in response to the results of root cause analysis to determine the reasons behind a 

significant RCPB degradation such as defined by SC1 or SC2. 
D-61

Table D.21 BWR LOCA Frequency Sensitivity Analysis Results 

LOCA Frequency – Statistical Mean [per Reactor-year] 

Base Flow Rate Interval [gpm] 

Case Cat1

100 < ν ≤ 1500 

Cat2

1500 < ν ≤ 5000 

Cat3

5000 < ν ≤ 25,000 

Cat4: 25,000 < ν ≤

100,000 

Cat5: 100,000 < ν

≤ 500,000 

Base-1 8.24E-06 7.64E-07 3.07E-07 1.22E-07 3.05E-08

Base-1 – SC1 8.70E-06 8.07E-07 3.27E-07 1.29E-07 3.29E-08 

Base-1 – SC2 1.30E-05 1.17E-06 4.77E-07 1.87E-07 5.63E-07 

D.6.6 Service Data and Conditional Failure Probabilities 

There is no service data associated with Cat0 LOCA events. Therefore, the estimation of conditional failure 

probabilities is based on zero-failure statistics. Since not all flaws propagate through-wall if left unattended, 

an alternative to the approach in Section D.5.2 (constrained noninformative prior) would be to use Jeffrey’s 

noninformative prior and to assume all flaws (non-through wall and through wall) as pressure boundary 

integrity challenges. The result would be conditional failure probabilities that are closely approximated by the 

power law (Equation D.8), however. It is acknowledged that this is just one way of representing the current 

state-of-knowledge with respect to gross Code Class 1 pipe failure. It is not a physical model of flaw

propagation given its interactions with certain loading conditions and pipe stresses. 

D.7 Summary of Results 

An application of a parametric attribute/influence method has yielded results as summarized in this section. 

Central to the method is the processing and interpretation of service data on Code Class 1 piping. A Markov 

model of piping reliability is used to develop time-dependent LOCA frequencies. 

D.7.1 Discussion of Assumptions 

A parametric attribute/influence method is applied to five base cases. Three types of assumptions are made in 

the analysis; global assumptions (applicable to all five base cases), BWR-specific assumptions and PWRspecific assumptions: 

Global Assumptions

• Pipe failure results from observable degradation mechanisms and loading conditions. A statistical 

evaluation of service experience data therefore provides a sufficiently accurate basis for piping 

reliability analysis. 

• The PIPExp database is of sufficient completeness and depth to support an application of the 

parametric attribute/influence methodology. This database addresses piping performance in response 

to both anticipated and unanticipated loading conditions. 

• The effect on piping reliability from pressure, deadweight, weld residual stresses, thermal loading, 

and thermal stratification is implicitly accounted for in the PIPExp database. This database also 

accounts for the effects from inadvertent over-pressurization and relief valve actuation, water 

hammer and seismic12 events. 

 

12 The database includes a single event involving the fracture of a small-diameter steam line due to seismic event (Fukushima-Daiichi Unit 6 on 07-21-

2000). 
D-62

BWR-Specific Assumptions

• The BWR-specific LOCA frequencies are assumed to be representative of a plant with IGSCC 

Category D and E welds operating with normal water chemistry (NWC). The pipe failure database 

includes plants with hydrogen water chemistry (HWC) and NWC. This study did not differentiate 

between plants with weld overlays and HWC versus plants with weld overlays and NWC, however. 

This study shows improved water chemistry together with weld reinforcements to lower the weld 

failure rates by about a factor of ten (10). 

• Because of service conditions and piping arrangements, flow accelerated corrosion (FAC) is not 

viewed as a significant degradation mechanism affecting Code Class 1 feedwater piping. 

Degradation involving wall thinning is therefore not viewed as having an effect on the timedependent LOCA frequency. 

PWR-Specific Assumptions

• The estimation of RC-HL weld failure rates is based on the assumption that the observed (in 

4

th quarter 2000) weld degradation at V.C. Summer is a circumferential flaw in the RPV 

nozzle-to-safe-end weld. This assumption is believed to result in an over-estimation of the 

actual weld failure rate. 

• Relative to PWRs of Westinghouse design, the pipe failure database includes no records on 

through-wall flaws in large-diameter pressurizer surge line welds. The analysis assumes that 

the piping is susceptible to thermal fatigue of sufficient magnitude to potentially cause a flaw 

in the through-wall direction. 

D.7.2 Summary of Input Data and Results 

Tables D.22 and D.23 summarize the input data to the LOCA frequency calculation. Tables D.24 through 

D.26 give the results at T = 25, 40 and 60 years, respectively. Consistent with the LOCA frequency elicitation 

structure, Table D.27 is summary of mid values (MV, or 50th percentiles) at T = 25 years rather than mean 

values, however. Figure D.41 shows the time-dependent LOCA frequencies. Figure D.42 shows selected 

weld failure rates. Figures D.43 through D.46 show the contribution to LOCA frequency by respective Base 

Case. Note that Figure D.45 includes the contribution to LOCA frequency by PWR-1 (Reactor Coolant 

System Hot Legs; all 3 loops are accounted for in this figure) and PWR-2 (Pressurizer Surge Line). Note that 

the Base Case results used in Table 4.1 in the main body can be obtained from Tables D.16, D.17, and D.20 

in this report. 

Table D.22 Summary of Key BWR Base Case Input Data

Base Case 

Input BWR-1 BWR-2 

Data NPS12 NPS22 NPS28 NPS12 NPS14 NPS20 

Weld count 50 16 56 63 5 53 

Weld failure rate 

Dominant [1/Reactor-yr.] 6.50E-05 1.54E-04 1.44E-04 2.20E-06 2.20E-06 1.58E-06 

Weld failure rate 

Minimum [1/Reactor-yr.] 2.37E-05 3.32E-05 1.29E-05 1.77E-07 1.77E-07 1.73E-07 
D-63

Table D.23 Summary of Key PWR Base Case Input Data

Base Case 

Input Data PWR-1 PWR-2 PWR-3 

NPS30 NPS14 NPS3-3⁄4 

Weld count 50 14 9 

Weld failure rate 

Dominant [1/Reactor-yr.] 7.64E-05 1.56E-06 6.56E-04 

Weld failure rate 

Minimum [1/Reactor-yr.] 1.05E-06 4.60E-08 1.58E-06 

Table D.24 Calculated LOCA Frequencies (T = 25 Years) 

LOCA Frequency – Statistical Mean [per Reactor-year] 

Base Flow Rate Threshold Value [gpm] 

Case Cat1

ν > 100 

Cat2

ν > 1,500 

Cat3

ν > 5,000 

Cat4

ν > 25,000 

Cat5

ν > 100,000 

Cat6

ν > 500,000 

BWR-113 9.46E-06 1.22E-06 4.60E-07 1.53E-07 3.05E-08 N/A14

BWR-215 2.54E-06 3.36E-07 1.25E-07 4.09E-08 7.33E-09 N/A 

PWR-116 7.42E-07 7.62E-08 2.93E-08 1.09E-08 3.77E-09 1.26E-09 

PWR-2 1.29E-07 1.50E-08 5.40E-09 1.56E-09 5.31E-10 N/A 

PWR-317 1.60E-05 2.32E-06 9.22E-07 N/A N/A N/A 

Table D.25 Calculated LOCA Frequencies (T = 40 Years) 

LOCA Frequency – Statistical Mean [per Reactor-year] 

Base Flow Rate Threshold Value [gpm] 

Case Cat1

ν > 100 

Cat2

ν > 1,500 

Cat3

ν > 5,000 

Cat4

ν > 25,000 

Cat5

ν > 100,000 

Cat6

ν > 500,000 

BWR-1 1.14E-05 1.47E-06 5.54E-07 1.84E-07 3.78E-08 N/A 

BWR-2 2.56E-06 3.39E-07 1.26E-07 4.13E-08 7.40E-09 N/A 

PWR-1 8.96E-07 9.20E-08 3.54E-08 1.32E-08 4.55E-09 1.45E-09 

PWR-2 1.60E-07 1.86E-08 6.70E-09 1.93E-09 6.59E-10 N/A 

PWR-3 1.95E-05 3.30E-06 9.44E-07 N/A N/A N/A 

Table D.26 Calculated LOCA Frequencies (T = 60 Years) 

LOCA Frequency – Statistical Mean [per Reactor-year] 

Base Flow Rate Threshold Value [gpm] 

Case Cat1

ν > 100 

Cat2

ν > 1,500 

Cat3

ν > 5,000 

Cat4

ν > 25,000 

Cat5

ν > 100,000 

Cat6

ν > 500,000 

BWR-1 1.88E-05 2.43E-06 9.16E-07 3.05E-07 6.07E-08 N/A 

BWR-2 2.56E-06 3.39E-07 1.26E-07 4.13E-08 7.40E-09 N/A 

PWR-1 9.74E-07 1.00E-07 3.85E-08 1.43E-08 4.95E-09 1.57E-09 

PWR-2 1.77E-07 2.06E-08 7.41E-09 2.14E-09 7.29E-10 N/A 

PWR-3 1.96E-05 3.32E-06 9.50E-07 N/A N/A N/A 

 

13 BWR-1 is the combination of RR Loop A and B. 

14 N/A = not applicable. 

15 The results are for FW Loop A and B. 

16 The results are for 3-of-3 RC hot legs. 

17 The results are for 2-of-2 HPI/NMU lines.
D-64

Table D.27 BWR and PWR LOCA Frequency Elicitation Anchor (MV) Values (T = 25 Years) 

Median (MV) LOCA Frequency [per Reactor-year] 

Base Flow Rate Threshold Value [gpm] 

Case Cat1

ν > 100 

Cat2

ν > 1,500 

Cat3

ν > 5,000 

Cat4

ν > 25,000 

Cat5

ν > 100,000 

Cat6

ν > 500,000 

BWR-1 8.23E-06 1.08E-06 4.03E-07 1.29E-07 2.19E-08 N/A 

BWR-2 1.09E-06 1.35E-07 5.03E-08 1.65E-08 2.10E-09 N/A 

PWR-1 1.54E-07 2.25E-08 8.33E-09 2.85E-09 8.53E-10 1.58E-10 

PWR-2 1.37E-08 1.39E-09 5.15E-10 1.54E-10 5.46E-11 N/A 

PWR-3 6.87E-06 1.15E-06 2.14E-07 N/A N/A N/A 

1.00E-09

1.00E-08

1.00E-07

1.00E-06

1.00E-05

1.00E-04

0 10 20 30 40 50 60 70

Years

LOCA Frequency [1/Reactor.Yr]

BWR-1 Cat1 LOCA

BWR-2 Cat1 LOCA

PWR-1 Cat1 LOCA

PWR-2 Cat1 LOCA

PWR-3 Cat1 LOCA

Figure D.41 Time-Dependent Cat 1 LOCA Frequencies 
D-65

1.44E-04

8.52E-05 7.82E-05 7.64E-05

1.96E-05

2.95E-06

6.56E-04

1.0E-06

1.0E-05

1.0E-04

1.0E-03

RR NPS28 Pipe-toSafe-endRR NPS28 Elbow-toPipeRR NPS12 Pipe-toReducerRC-HL NPS30 Nozzleto-Safe-endRC-HL NPS30 Elbowto-PumpRC Surge Line

(NPS14) Nozzle-toSafe-end

HPI/NMU (NPS 3-3/4)

Pipe-to-Nozzle Safeend

Base Case Weld Failure Rates (T = 25 Years) - Key LOCA Frequency Contributors

Weld Failure Rate [1/Reactor-yr.]

 Weld failure is defined as through-wall leak. Leak rate is < T.S. limit.

Figure D.42 Selected Base Case Weld Failure Rates 

4.60E-07

1.53E-07

3.05E-08

1.22E-06

9.46E-06

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

Leak Rate > 100 gpm Leak Rate > 1500 gpm Leak Rate > 5000 gpm Leak Rate > 25,000 gpm Leak Rate > 100,000 gpm

BWR Reactor Recirculation System - T = 25

LOCA Frequency [1/Reactor-yr.]

Figure D.43 BWR-1 LOCA Frequency 
D-66

4.70E-09

1.52E-09

2.42E-10

1.26E-08

9.59E-08

1.0E-10

1.0E-09

1.0E-08

1.0E-07

1.0E-06

Leak Rate > 100 gpm Leak Rate > 1500 gpm Leak Rate > 5000 gpm Leak Rate > 25,000 gpm Leak Rate > 100,000

gpm

Leak Rate > 500,000

gpm

BWR-2 (Feedwater System) - T = 25

LOCA Frequency [1/Reactor-yr.]

Figure D.44 BWR-2 LOCA Frequency (Feedwater System Loop A & B) 

3.19E-08

1.17E-08

3.99E-09

1.20E-09

8.33E-08

7.99E-07

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

Leak Rate > 100 gpm Leak Rate > 1500 gpm Leak Rate > 5000 gpm Leak Rate > 25,000 gpm Leak Rate > 100,000

gpm

Leak Rate > 500,000

gpm

PWR Reactor Coolant System - T = 25 (Hot Legs + Surge Line)

LOCA Frequency [1/Reactor-yr.]

Figure D.45 PWR-1 and PWR-2 LOCA Frequency 
D-67

9.21E-07

3.22E-06

1.90E-05

1.0E-07

1.0E-06

1.0E-05

1.0E-04

Leak Rate > 100 gpm Leak Rate > 1500 gpm Leak Rate > 5000 gpm Leak Rate > 25,000 gpm Leak Rate > 100,000

gpm

Leak Rate > 500,000

gpm

PWR HPI/NMU System - T = 25

LOCA Frequency [1/Reactor-yr.]

Figure D.46 PWR-3 LOCA Frequency (ASME Code Class 1 HPI/NMU System) 

Figure D.47 displays the results of a sensitivity analysis associated with the BWR Base Cases. It is concerned 

with the influence by service data on the calculated LOCA frequency. 

1.30E-05

8.70E-06 8.24E-06

1.0E-06

1.0E-05

1.0E-04

BWR LOCA (RR System - Base Case - Cat1) BWR LOCA (RR System - Sensitivity Case 1 - Cat1) BWR LOCA (RR System - Sensitivity Case 2 - Cat1)

BWR Base Case Sensitivity Analysis Results

LOCA Frequency [1/Reactor-yr.]

Sensitivity Case 1: Posterior w eld failure rate assumes 1 small leak in RR28 

pipe-to-safe-end w eld.

Sensitivity Case 2: Posterior w eld failure rate assumes 1 large leak in RR28 

pipe-to-safe-end w eld.

Figure D.47 Selected BWR Base Case Sensitivity Analysis Results – Cat 1 LOCA 
D-68

Figures D.48 (BWR) and D.49 (PWR) show the influence of in-service inspection on the time-dependent 

LOCA frequency; no ISI and ISI with POD = 0.5 and 0.9, respectively. 

1.00E-07

1.00E-06

1.00E-05

1.00E-04

0 10 20 30 40 50 60 70

Years

Cat1 LOCA Frequency [1/Reactor.Yr]

BWR-1 Cat1 Base Case (POD = 0.5)

BWR-1 Cat1 - No ISI

BWR-1 Cat1 - ISI w/POD = 0.9

Figure D.48 Influence of ISI on Time-Dependent BWR-1 Cat 1 LOCA Frequency 

1.00E-08

1.00E-07

1.00E-06

1.00E-05

0 10 20 30 40 50 60 70

Years

Cat1 LOCA Frequency [1/Reactor.Yr]

PWR-1 Cat1 Base Case (POD = 0.5)

PWR-1 Cat1 - No ISI

PWR-1 Cat1 - ISI w/POD = 0.9

Figure D.49 Influence of ISI on Time-Dependent PWR-1 Cat 1 LOCA Frequency 
D-69

D.7.3 Benchmarking 

A limited scope benchmarking exercise was performed to compare predicted weld failure rates with the 

reported service experience. The benchmarking was limited to NPS12 BWR reactor recirculation welds 

susceptible to IGSCC. Probabilistic fracture mechanics (PFM) calculations using the WinPRAISE computer

code generated predictions about the weld failure rate for different assumptions about the normal operating 

stresses (σNO).18 Bayesian reliability analysis was used to derive weld failure rates from service experience 

data. Figure D.50 shows the results of the benchmarking exercise. Table D.28 includes a description of the 

different cases of the benchmarking exercise. 

Table D.28 Benchmarking of WinPRAISE Versus Service Experience 

Case Definition

BOYL (PIPExp) 

Table D.13 (this report). NPS12 Reactor Recirculation pipe-to-reducer weld with 

weld overlay. T = 25 years. This weld configuration has the highest predicted 

failure rate. 

DOH-1 (D.O. Harris) NPS12 reactor recirculation system weld with normal operating stress, σNO = 10 

ksi19

DOH-2 NPS12 reactor recirculation system weld; σNO = 12 ksi 

DOH-3 NPS12 reactor recirculation system weld; σNO = 15 ksi 

DOH-4 NPS12 reactor recirculation system weld; σNO = 20 ksi 

7.82E-05

5.68E-05

6.15E-04

3.74E-03

1.19E-02

1.00E-05

1.00E-04

1.00E-03

1.00E-02

1.00E-01

BOYL

(PIPExp)

DOH-1 DOH-2 DOH-3 DOH-4 

Weld Failure Rate [1/Reactor-yr.]

Figure D.50 Predicted (WinPRAISE) Versus Observed Weld Failure Rate (PIPExp) 

 

18 D.O. Harris, “Progress in Benchmarking SCC for 12 Inch Recirculation Line, July 1, 2003. 

19 1 ksi = 6.9 MPa 
D-70

D.7.4 Comparison to Historical LOCA Frequency Estimates 

Figures D.51 and D.52 compare the Base Case results to historical LOCA frequency estimates. Direct (oneto-one) comparisons are not feasible due to different LOCA definitions and estimation approaches. Listed 

below are the selected BWR and PWR LOCA frequency references. 

BWR Large (≥ Cat3) LOCA Frequency Estimates (Figure D. 51)

• SKI 98:30 (FW/RR); the displayed value range is taken from Reference [D.18]. It excludes 

contribution from thermal fatigue in Code Class 1 feedwater system piping. The feedwater system 

design is unique to the pilot plant in SKI Report 98:30 and it is therefore not applicable to BWR-2. 

• NUREG/CR-5750 (Appendix J) provides recommended pipe LOCA frequencies. The given value 

range accounts for all Code Class 1 pipe failure contributions. 

• GRS-98 is a probabilistic safety assessment of the German plant Gundremmingen; a BWR plant 

designed and built by Kraftwerk Union. This reactor design has no external recirculation loops; the 

given LOCA frequency value range accounts only for contributions from Code Class 1 feedwater 

pipe failure. 

• BFN-1 (NUREG/CR-2802) is the 1982 probabilistic safety assessment of Browns Ferry Unit 1 

performed as part of the NRC-sponsored Interim Reliability Evaluation Program. The given LOCA 

frequency value range accounts for Reactor Recirculation pump suction piping failure. 

PWR Large (≥ Cat3) LOCA Frequency Estimates (Figure D. 52)

• NUREG/CR-5750 (Appendix J) provides recommended pipe LOCA frequencies. The given value 

range accounts for all Code Class 1 pipe failure contributions. 

• Surry-1 (1990 Expert Elicitation). Surry-1 is a 3-loop Westinghouse reactor, similar to the PWR1/PWR-2 reference design. The given LOCA frequency value range applies to RCS pipe failure and 

resulted from a NRC-sponsored expert elicitation.20

• EPRI TR-100380 (Piping Failures in U.S. Commercial Nuclear Power Plants, 1992) includes 

recommended BWR and PWR LOCA frequencies that are based on statistical analysis of service 

data. The given LOCA frequency value range accounts for all Code Class 1 pipe failure 

contributions. 

 

20 See for example T.V. Vo et al (1991). “Estimates of Rupture Probabilities for Nuclear Power Plant Components: Expert Judgment Elicitation,” 

Nuclear Technology, 96:259-270. 
D-71

1.5E-06

5.2E-06

9.0E-05

4.5E-07

4.5E-05

5.4E-08

1.9E-07

9.0E-07

4.5E-09

4.6E-07 4.5E-07

1.6E-06

2.0E-05

1.0E-07

9.9E-06

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

Base Case Report No. 2 SKI 98:30 (FW+RR) NUREG/CR-5750 (App

J)

GRS-98 (FW System

Only)

BFN-1 (NUREG/CR2802)

Large LOCA Frequency [Yr-1

]

Upper Bound

Lower Bound

Mean

 

Figure D.51 Comparison of Selected BWR Large LOCA Frequency Estimates 

1.1E-07

1.0E-05

8.0E-05

7.0E-04

2.8E-09

1.0E-07

8.0E-07

1.2E-05

3.2E-08

4.0E-06

2.1E-05

1.4E-04

1.0E-09

1.0E-08

1.0E-07

1.0E-06

1.0E-05

1.0E-04

1.0E-03

Base Case Report No. 2 NUREG/CR-5750 (App J) Surry-1 (1990 Expert Elicitation) EPRI TR-100380

Large LOCA Frequency [Yr-1

]

Upper Bound

Lower Bound

Mean

Figure D.52 Comparison of Selected PWR Large LOCA Frequency Estimates 
D-72

D.8 References

D.1 Scott, P., 2003. Meeting Notes from U.S. NRC LOCA Elicitation Kick-Off Meeting, Rockville (MD), 

February 4-6, 2003. 

D.2 Zigler, G. et al, 1995. Parametric Study of the Potential for BWR ECCS Strainer Blockage Due to 

LOCA Generated Debris, NUREG/CR-6224, U.S. Nuclear Regulatory Commission, Washington (DC). 

D.3 U.S. Nuclear Regulatory Commission, 1988. NRC Position on IGSCC in BWR Austenitic Stainless 

Steel Piping, Generic Letter 88-01 (January 25, 1988), Washington (DC). 

D.4 U.S. Nuclear Regulatory Commission, 1992. NRC Position on IGSCC in BWR Austenitic Stainless 

Steel Piping, Generic Letter 88-01, Supplement 1 (February 4, 1992), Washington (DC). 

D.5 Structural Integrity Associates, Inc., 2001. Degradation Mechanisms Evaluation for Class 1 Piping at 

Plant B Nuclear Station, File No. EPRI-156-310. 

D.6 Structural Integrity Associates, Inc., 2001. Degradation Mechanisms Evaluation for the Class 1 

(Category B-J/B-F) Piping at Plant A Nuclear Station, File No. EPRI-156-310. 

D.7 Shah, V.N. et al, 1998. Assessment of Pressurized Water Reactor Primary System Leaks, NUREG/CR6582, U.S. Nuclear Regulatory Commission, Washington (DC). 

D.8 Boman, B.L. et al, 2000. “Evaluation of Oconee-2 High Pressure Injection/Normal Makeup (HPI/NMU) 

Line Weld Failure,” Assessment Methodologies for Preventing Failure: Service Experience and Environment 

Considerations, PVP-Vol. 410-2, American Society of Mechanical Engineers, New York (NY), ISBN: 0-

7918-1891-8, pp 111-117. 

D.9 Deardorff, A.F., 2001. Interim Thermal Fatigue Management Guideline (MRP-24), 1000701, Electric 

Power Research Institute, Palo Alto (CA). 

D.10 Trolle, M., 1996. Status för gjutet rostfritt stål i äldre svenska kärnkraftverk, mars 1996 (Status of Cast 

Austenitic Stainless Steel in Older Swedish Nuclear Power Plants), SKI Report 96:26, Swedish Nuclear 

Power Inspectorate, Stockholm (Sweden). 

D.11 Nyman, R. et al, 2003. Reliability of Piping System Components. Framework for Estimating Failure

Parameters from Service Data, SKI Report 97:26 (2nd Edition), Swedish Nuclear Power Inspectorate, 

Stockholm (Sweden). 

D.12 Fleming, K.N. and B.O.Y. Lydell, 2004. “Database Development and Uncertainty Treatment for 

Estimating Pipe Failure Rates and Rupture Frequencies,” Reliability Engineering and System Safety, article 

in press. 

D.13 Lydell, B.O.Y., E. Mathet and K. Gott, 2004. ”Piping Service Life Experience in Commercial Nuclear 

Power Plants: Progress with the OECD Pipe Failure Data Exchange Project,” Proc. ASME PVP-2004 Conf.: 

2004 ASME Pressure Vessels and Piping Conference, American Society of Mechanical Engineers, New 

York (NY). 

D.14 Fleming, K.N., 2004. “Markov Models for Evaluating Risk-informed In-service Inspection Strategies 

for Nuclear Power Plant Piping Systems,” Reliability Engineering & System Safety, 83:27-45. 

D.15 Lydell, B.O.Y., 1999. Failure Rates in Barsebäck-1 Reactor Coolant Pressure Boundary Piping. An 

Application of a Piping Failure Database, SKI Report 98:30, Swedish Nuclear Power Inspectorate, 

Stockholm (Sweden). 

D.16 Lydell, B.O.Y., 1999. Failure Rates in Barsebäck-1 Reactor Coolant Pressure Boundary Piping. SKI

98:30 Appendix H: Barsebäck-1 Piping Reliability Database, RSA-R-99-01P (Proprietary), Prepared for 

Barsebäck Kraft AB (Sweden), RSA Technologies, Fallbrook (CA). 

D.17 Lydell, B.O.Y., 1997. Strategies for Reactor Safety: Preventing Loss of Coolant Accidents, Nordic 

Nuclear Safety Research, Risø (Denmark), NKS/RAK-1(97)R10, ISBN: 87-7893-046-4. 
D-73

D.18 Wachter, O. and G. Brümmer, 1997. “Experiences with Austenitic Steels in Boiling Water Reactors,” 

Nuclear Engineering and Design, 168:35-52. 

D.19 U.S. Nuclear Regulatory Commission, 1992. Higher Than Predicted Erosion/Corrosion in Unisolable 

Reactor Coolant Pressure Boundary Piping Inside Containment at a Boiling Water Reactor, Information 

Notice 92-35 (May 6, 1992), Washington (DC). 

D.20 International Incident Reporting System, 2000. Cracks in Weld Area of Safe-end to the Reactor 

Pressure Vessel, IRS No. 7397, International Atomic Energy Agency, Vienna (Austria). 

D.21 Crowley, B. et al, 2001. Virgil C. Summer Nuclear Station – NRC Special Inspection Report No. 50-

395/00-08, Exercise of Enforcement Discretion, U.S. Nuclear Regulatory Commission, Washington (DC). 

D.22 U.S. Nuclear Regulatory Commission, 1988. Thermal Stresses in Piping Connected to Reactor Coolant 

Systems, Bulletin 88-08 (June 22, 1988), Washington (DC). 

D.23 U.S. Nuclear Regulatory Commission, 1988. Pressurizer Surge Line Thermal Stratification, Bulletin 

88-11 (December 20, 1988), Washington (DC). 

D.24 OECD Nuclear Energy Agency, 1998. Experience with Thermal Fatigue in LWR Piping Caused by 

Mixing and Stratification, NEA/CSNI/R(98)8, Issy-les-Moulineaux (France). 

D.25 Aaltonen, P., K. Saarinen and K. Simola, 1993. “The Correlation of IGSCC Propagation with the 

Power Plant Trainsient History,” The International Journal of Pressure Vessels and Piping, 55:149-162. 
D-74

ATTACHMENT A TO APPENDIX D 

SUMMARY OF PIPEXP DATABASE 

The PIPExp database has evolved over a period of about ten years. A first database version was developed 

with financial support from the Swedish Nuclear Power Inspectorate (SKI). Since the conclusion of the initial 

R&D effort in 1998 an active maintenance program has supported the database. 

D.A.1 Database Structure

Designed in Access, the database consists of searchable free-format text fields and a large number of data 

fields that are used as data filters in support of a range of data processing needs. Table D.A.1 is a summary of 

text and data fields. 

D.A.2 Completeness and Quality Management

The completeness of the pipe failure data is addressed through a continuous database management program. 

Extracted from Monthly Summary Reports, Table D.A.2 provides snapshots of the database evolution from 

1998 to the present. In PIPExp, each record is assigned a ‘Quality Index’ (Table D.A.1, item #4, and Table 

D.A.3) as one means of monitoring the completeness and technical accuracy of source information as well as 

the process of classifying and coding of the source information. 

Table D.A.1 Description of Data Fields in PIPExp 

Item No. Field Name Type Description 

1 UPDATE Date Date of the most recent update. 

2 MER Yes/No21

Multiple Events Report; some reports include information on more than 

one crack/leak in one system. Used to identify events where a discovery 

resulted in an investigation (e.g., augmented ISI) to identify further 

piping degradation due to a common cause. A new record is added if 

additional degradation is positively identified (by component socket). 

3 DDA Text Data filter used to classify a record as either ‘public’ (= Licensee Event 

Report), ‘restricted’ or ‘proprietary.’ 

4 QA-Index Number QA-Index of ‘1’ signifies a data entry determined to be ‘complete.’ By 

contrast, a QA-Index of ‘6’ signifies a database entry for which only a 

LER (or equivalent) abstract was available. 

5 EVENT DATE Date Event date (MM/DD/YY); date of discovery (in case of ISI). 

6 PLANT TYPE Text Plant type; e.g., BWR, PWR, used as data filter. 

7 DESIGN Text NSSS design/design generation; keyword using generally accepted or 

standard nomenclature. This field is used as a data filter. 

8 NSSS-VENDOR Text Reactor vendor; e.g., ABB-Atom, KWU/Siemens, Westinghouse; used as 

data filter. 

9 PLANT NAME Text Plant name 

10 COUNTRY Text Two-letter code based on the ISO 3166-1-alpha-2 code elements. 

11 CONSTRUCTOR Text Name of company responsible for the original piping system design. The 

default name is the architect engineering firm. Used as data filter. 

12 COD Date Date (MM/DD/YY) of commercial operation as default. If known, date 

of initial criticality. For U.S. data, based on NUREG-0020 

13 PLANT Text 

Plant operational state (at the time of discovery); keyword using 

generally accepted or standard nomenclature. This field is used as a data 

 

21 A check box without check mark implies ‘No’ or ‘Unknown/Pending.’
D-75

Table D.A.1 Description of Data Fields in PIPExp 

Item No. Field Name Type Description 

OPERATIONAL 

STATE - POS 

filter. Pulldown menu with the following options: 

§ CSD – Cold Shutdown 

§ HSD – Hot Shutdown 

§ HSB – Hot Standby 

§ Refueling 

§ Shutting Down 

§ Starting Up 

§ Power Operation 

14 REFERENCE-1 Text Primary reference 

15 REFERENCE-2 Text Secondary (or supplemental) reference 

16 REFERENCE-3 Text Tertiary (or supplemental) reference 

17 LER-RO? Yes/No Check if the information source is a Licensee Event Report (or 

equivalent); i.e., from a regulatory reporting system. 

18 EVENT TYPE Text Event type; ‘Crack’, ‘Wall Thinning’, ‘P/H-leak’ (P/H = pinhole), ‘Leak’, 

‘Severance’, ‘Rupture.’ Used as data filter. 

19 FAILURE-ONDEMAND Yes/No Check if pipe failure occurred when a demand was placed on the affected 

system (e.g., standby system). Used as data filter.

20 SYNERGY Yes/No 

Check if the pipe failure was caused by multiple degradation 

mechanisms; e.g., crack initiation through IGSCC and crack propagation 

through thermal fatigue. Used as data filter. 

21 DEGRADATION+ 

LOADING Yes/No 

Check if the pipe failure resulted from the combined effect of a 

degradation mechanism (e.g., flow-accelerated corrosion, FAC) and a 

severe (or unusual) loading condition. Used as data filter. 

22 ECA Text Event Category. Used as data filter. This database field is used to 

characterize actual or potential impact on plant risk by a degradation or 

failure. The following options are available: 

§ S-/M-/L-LOCA (implies that a pressure boundary failure resulted in 

ESF actuation); 

§ S-/M-/L-LOCA Precursor (implies mitigation of a pressure 

boundary failure through prompt operator response; e.g., plant 

shutdown prior to reaching ESF actuation setpoint);

§ Internal Flooding (spill rate in excess of room/compartment floor 

drain capacity); 

§ Internal Flooding Precursor (accumulation of large water volumes 

prevented through prompt operator response); 

§ Common Cause Initiating (CCI) Event (pressure boundary failure 

results in spatial effects through spraying or steaming of safety 

equipment); 

§ CCI Precursor (pressure boundary failure results in spraying or 

steaming but prompt operator action prevents safety equipment from 

being affected); 

§ System Disabled (pressure boundary failure is large enough to 

incapacitate a system function); 

§ System Degraded (default used for at-power events that result in an 

entry into a Technical Specification Action Statement). 

23 CCC Yes/No Check if event is considered to be a ‘common cause candidate’ (CCC) 

event. Used as data filter. 

24 CA Text Corrective Action. Used as data filter. The following types of corrective 

action are defined: 

§ REPAIR (used in a generic sense); 

§ REPLACEMENT 

§ REPLACEMENT – IN-KIND 

§ REPLACEMENT – NEW MATERIAL 

§ TEMP. REPAIR (temporary repair to allow continued operation 

until next refueling outage or major maintenance outage at which 

time a Code-repair (which would require system isolation and 
D-76

Table D.A.1 Description of Data Fields in PIPExp 

Item No. Field Name Type Description 

draining) or a replacement is performed. 

§ WOR (= weld overlay repair); primarily applies to ASME Section 

XI Class 1 or 2 (or equivalent) piping. 

25 ISS Yes/No 

Safety system actuation; check if pipe failure resulted automatic actuation 

of a make-up system or other safety system. Used as data filter. 

26 IRT Yes/No Automatic reactor trip; check if pipe failure resulted in automatic reactor 

trip/turbine trip. Used as data filter. 

27 IPO Text Impact of pipe failure on plant operation; e.g., power reduction, manual 

reactor trip. Used as data filter. 

28 TTR Number Repair time in hours. 

29 TTR-Class Number A data filter: 1: TTR ≤ 8 hours; 2: 8 < TTR ≤ 24 hours; 3: 24 < TTR ≤ 96 

hours; 4: 96 < TTR ≤168 hours; 5: TTR > 168 hours. 

30 NARRATIVE Memo 

Event narrative; includes details on plant condition prior to event and 

plant response during event, method of detection, corrective action plan. 

This field should include sufficient information to support independent 

verification of the event data classification. 

31 LQT Number Quantity of process medium released [kg] 

32 DOL Text Duration of release 

33 LRT Number Leak rate [kg/s] 

34 gpm Number Leak rate [U.S. gallons/minute] 

35 LEAK CLASS Number A data filter: 1: Leak Rate (LR) ≤ 1 gpm; 2: 1 < LR ≤ 5 gpm; 3: 5 < LR ≤

10 gpm; 4: 10 < LR ≤ 50 gpm; 5: LR > 50 gpm. 

36 FLO Text 

Location of crack/leak/rupture; description of where in the piping system 

a degradation or failure occurred. Include sufficient detail to support the 

consequence evaluation/classification. 

37 K1 Yes/No Data filter; steamline break outside containment. 

38 K2 Yes/No Data filter; feedwater line break 

39 K3 Yes/No Data filter; steamline break inside containment. 

40 IMPULSE-LINE Yes/No Check (= ‘Yes’) if affected line is a valve impulse line. 

41 INSTR. LINE Yes/No Check (= ‘Yes’) if affected line is an instrument sensing line. 

42 ISOMETRIC 

DRAWING # 

Text Isometric drawing number 

43 P&ID # Text Piping and instrument drawing number. 

44 MSA Text Name of the affected plant system 

45 SHARED Yes/No Check (= ‘Yes’) if affected piping is shared by two reactor units. Mainly 

applies to support systems (e.g., Service Water, Instrument Air) where 

sections of a piping system may be shared by two reactor units; this is 

relatively common in the U.S. 

46 OSA Text Name of other systems affected by the degradation or failure. Secondary 

effects of piping failure 

47 S-TYPE Text 

Category of system affected by the degradation or failure. Used as data 

filter. The following types are used: 

• RCPB (Reactor Coolant Pressure Boundary); 

• SIR (Safety Injection & Recirculation); includes emergency core 

cooling systems & decay heat removal). 

• CS (Containment Spray) 

• AUX (Reactor Auxiliary Systems); includes component cooling 

water, chemical & volume control, reactor water cleanup, control 

rod drive, containment heat removal, standby liquid control, 

radwaste control, spent fuel pool cooling. 

• FWC (Feedwater & Condensate Systems) 

• STEAM (Main Steam System) 

• SUPPORT (Service Water & Instrument Air systems) 

• PCS (turbine generator) 

• FIRE (Fire Protection). 
D-77

Table D.A.1 Description of Data Fields in PIPExp 

Item No. Field Name Type Description 

48 ISO Yes/No Check if the affected pipe section can be isolated to prevent or mitigate 

direct/indirect impacts. 

49 DET Text 

Method of detection; e.g., ISI, WT = walk-through inspection, leak 

detection system in combination with control room indication and/or 

alarm. Used as data filter. Pulldown menu with the following options: 

§ Walk-through 

§ UT-examination 

§ Liquid penetrant testing 

§ Hydrotesting 

§ Leak detection 

§ Containment/drywell inspection 

§ Control Room Indication 

50 DRYWELL 

ENTRY Yes/No 

BWR-specific data field. Checked for ‘at-power’, unidentified P/H-leak 

or leak requiring power reduction or reactor shutdown for containment 

drywell entry to determine leak source. Used as a data filter. Also, this 

data could be input to plant availability models. 

51 CONTAINMENT 

ENTRY Yes/No Check if power reduction initiated to allow for containment entry to identify source of leakage. Used for other than BWR plants. 

52 CRS Text Verbal description of crack morphology; orientation and size/ geometry 

of crack or fracture 

53 CRACK-DEPTH Number Crack depth in percent of wall thickness (a/t-ratio) 

54 AXIAL-LENGTH Number Axial crack length in [mm]. 

55 CRACK-LENGTH Number Circumferential crack length as percent of inside diameter 

56 ASPECT-RATIO Number Ratio of crack depth (a) to flaw length (L) 

57 WELD-CONFIG Text Configuration of the affected weld in a piping system; e.g., BP = bend-topipe weld, PP = pipe-to-pipe weld, etc. 

58 INSIDE 

CONTAINMENT Yes/No Check if pipe failure located inside containment 

59 AUXILIARY 

BUILDING Yes/No Check if pipe failure located in Auxiliary Building (PWR) 

60 REACTOR 

BUILDING Yes/No Check if pipe failure located in Reactor Building (BWR) 

61 TURBINE 

BUILDING Yes/No Check if pipe failure located in Turbine Building 

62 Not used N/A N/A 

63 Not used N/A N/A 

64 CTA Text Component Type; pulldown menu with the following options: 

§ Bend 

§ Elbow 

§ Elbow – 45-degree 

§ Elbow – 90-degree 

§ Elbow – LR (Long Radius) 

§ Pipe 

§ Reducer 

§ Tee 

§ Weld 

§ Socket weld 

65 ASME Class Number Differentiate between 1, 2, 3 and 4 (= non-Code Class) 

66 BELOW-GRADE Yes/No Check if ‘Yes’; Below Grade / Underground Piping. Used as data filter. 

67 FIELD-WELD Yes/No Check if ‘Yes’. Used as data filter. 

68 SHOP-WELD Yes/No Check if ‘Yes’. Used as data filter. 

69 CONCRETE-LINED Yes/No Check if ‘Yes’; could apply to essential or non-essential service water (or 

equivalent) system piping. Used as data filter. 

70 REPLACEMENT Yes/No Check if piping replaced using new material. 

71 REPL-DATE Date/Time Date of component (e.g., weld and spool piece) replacement. Used in 

hazard plotting. 
D-78

Table D.A.1 Description of Data Fields in PIPExp 

Item No. Field Name Type Description 

72 YOO Number Years of commercial operation when failure occurred. Used in aging 

analysis. 

73 AGE Number Age of component socket [hours]. Used in hazard plotting. For additional 

information. 

74 CLASS Number 

Based on diameter; events grouped in six diameter classes; 1 = (≤ DN15), 

2 = (15 < DN ≤ 25), 3 = (25 < DN ≤ 50), 4 = (50 < DN ≤ 100), 5 = (100 < 

DN ≤ 250), 6 = (> DN250). DN = nominal diameter in [mm]. This field 

is used as a data filter. 

75 THOMAS Number 

Ratio of diameter and pipe wall thickness ([CSI/WTK]); for details, see 

the paper by H.M. Thomas (1981): “Pipe and Vessel Failure Probability,” 

Reliability Engineering, 2:83-124. This field is used as a data filter. 

76 CSI Number Nominal diameter [DN] in [mm]. Used as data filter. 

77 WTK Number Wall thickness [mm] 

78 SCHEDULE Number Pipe schedule number 

79 DIS-MET Yes/No Dissimilar metal weld; check if ‘yes’. Used as data filter. 

80 MTR Text Material; e.g., carbon steel, stainless steel, etc. Used as data filter. 

81 MTR-DES Text Material designation according to national standard; e.g., AISI 304, 

SS2343, etc. Used as data filter. 

82 PMD Text Process medium. Used as data filter. 

83 RAW WATER Text Source of raw water (applies to Fire Protection and Service Water 

piping); differentiate between LAKE – RIVER – SEA-BRACKISH. 

Used as data filter. 

84 STG Yes/No Normally stagnant process medium? Used as data filter. 

85 HWC Yes/No For BWRs; hydrogen water chemistry; check if ‘Yes’. Used as data filter 

(e.g., in factor-of-influence assessments). 

86 HWC-START Date/Time Date when HWC was introduced

87 NMCA Yes/No Check if Noble Metal Chemical Addition. Used as data filter. 

88 NMCA-Start Date/Time Date when NMCA started. 

89 IHSI Yes/No Induction heat stress improvement; check if ‘Yes’. Used as data filter. 

90 IHSI-DATE Date/Time Date when IHSI was performed

91 MSIP Yes/No Check if Mechanical Stress Improvement Process applied to weld. Used 

as data filter. 

92 MSIP-Date Date/Time Date of MSIP application 

93 S-A Number Stress intensity allowance; ratio of the critical stress intensity factor to the 

assessed stress intensity factor given a flaw. This information is extracted 

from fracture mechanics evaluations. 

94 OPA Number Operating temperature [°C] 

95 DPA Number Design temperature [°C] 

96 OPB Number Operating pressure [MPa] 

97 DPB Number Design pressure [MPa] 

98 OPC Text Process medium chemistry (for primary system); e.g., NWC = normal 

water chemistry, HWC = hydrogen water chemistry. Used as data filter. 

99 MPR Text Method of fabrication; e.g., cold formed, hot formed. Used as data filter. 

100 SYS Yes/No 

Systematic failure? Used as data filter to enable queries that address the 

effectiveness of remedial actions (e.g., preventing recurring failures). 

101 RFL Text Description of the extent and nature of a systematic failure 

102 REST Yes/No 

Failure due to deficient system restoration?; e.g., no venting prior to fill 

procedure, etc. Used as data filter. 

103 CEA Text Apparent cause of failure; e.g., IGSCC, PWSCC, TGSCC, etc. Used as 

data filter. Pull down menu with the following options: 

• B/A SCC (B/A = boric acid) 

• Corrosion (general, pitting or crevice corrosion) 

• Corrosion-fatigue 

• Erosion 
D-79

Table D.A.1 Description of Data Fields in PIPExp 

Item No. Field Name Type Description 

• Erosion-cavitation 

• Flow accelerated corrosion (FAC) 

• Fretting 

• HF: Construction/installation error 

• HF: Human error 

• HF: Repair/maintenance error 

• HF: Welding error 

• HPSCC (High Potential SCC) 

• IGSCC 

• MIC (Microbiologically induced corrosion) 

• Overpressurization 

• PWSCC 

• Severe overloading (other than water hammer) 

• SICC (Strain-rate induced SCC) 

• TGSCC 

• Thermal fatigue 

• Unknown 

• Unreported 

• Vibration 

• Vibration-fatigue 

• Water hammer 

104 EPRI-CODE Text 

Failure code as used in the EPRI ’97 / EPRI ’98 databases (see for 

example EPRI TR-110161 (Piping System Reliability and Failure Rate 

Estimation Models for Use in Risk-Informed In-Service Inspection 

Applications, December 1998). Used as data filter. Pull down menu with 

the following options: 

§ CF – Corrosion-fatigue 

§ COR – General corrosion, microbiologically induced corrosion 

(MIC), pitting corrosion 

§ COR-EXT – external corrosion 

§ D&C – Design & Construction errors 

§ E-C – Erosion-cavitation 

§ E/C – Erosion-corrosion 

§ FRET – Fretting 

§ HE – Human Error 

§ OVP – Overpressurization 

§ SCC – Stress corrosion cracking 

§ TF – Thermal Fatigue 

§ UNK – Unknown 

§ VF – Vibration-fatigue 

§ WH – Water hammer 

105 RC1 Text Contributing factor number 1 

106 RC2 Text Contributing factor number 2 

107 CEC Memo 

Description of events and causal factors. Include sufficient technical 

detail from the root cause analysis process so that recurrence may be 

prevented. 

108 CMT Memo 

Any other information of relevance to understanding of underlying causal 

factors. Also, information on the type and extent of repair/replacement. 

The purpose of this free-format database field is to facilitate future 

applications, for example, by codifying the information on piping 

replacements. 

109 ISI Yes/No 

Deficient ISI; e.g., ISI not performed, or ISI failed to detect a flaw. Used 

as data filter to identify events caused by ISI program deficiencies (e.g., 

affected component should have been included in program) or an 

inspection prior to failure missed a degradation that propagated in the 
D-80

Table D.A.1 Description of Data Fields in PIPExp 

Item No. Field Name Type Description 

through-wall direction. 

110 ISI-CMT Memo Comments on ISI history; e.g., date of last inspection, details on 

examination technique(s). 

111 mCF Yes/No Check if there are multiple circumferential flaws in a weld. 

112 Number of Flaws Number 

113 D0-1 Number Distance from 12 o’clock position to the first circumferential flaw; this 

field is repeated for up to nine flaws. 

114 CF-1 Number Length of the first circumferential flaw (counted from the 12 o’clock 

position 
Table D.A.2 Summary of PIPExp Database Development

Database as of 12-31-1998 

No. Damage Reports by QA-Index22

Database as of 12-31-1999 

No. Damage Reports by QA-Index 

Totals 1 2 3 4 5 6 Totals 1 2 3 4 5 6 

673 210 66 3 74 7 277 1595 1000 168 2 146 53 226 

100 30 3 -- 56 1 10 100 30 3 -- 56 1 10 

1376 386 123 6 152 84 746 1656 645 176 5 211 208 411 

57 3 6 -- 19 28 1 66 7 6 -- 22 31 -- 

2291 629 198 9 301 120 1034 3417 1682 253 7 435 243 647 

Database as of 12-31-2000 

No. Damage Reports by QA-Index 

Database as of 12-31-2001 

No. Damage Reports by QA-Index 

Totals 1 2 3 4 5 6 Totals 1 2 3 4 5 6 

1711 1111 164 2 175 58 201 1784 1172 166 2 197 63 184 

95 43 1 -- 41 10 -- 96 44 1 -- 41 10 -- 

1748 696 181 5 260 209 397 1952 811 194 5 329 221 392 

125 12 5 1 77 30 -- 125 12 5 1 77 30 -- 

3679 1862 351 8 553 307 598 3957 2039 366 8 644 324 576 

Database as of 12-31-2002 

No. Damage Reports by QA-Index 

Database as of 06-01-2004 

No. Damage Reports by QA-Index 

Totals 1 2 3 4 5 6 Totals 1 2 3 4 5 6 

1872 1216 174 12 219 75 176 2033 1370 189 77 233 93 71 

-- -- -- -- -- -- -- 12 12 -- -- -- -- -- 

106 51 2 -- 42 11 -- 101 47 1 -- 41 12 -- 

2077 1011 198 6 351 233 278 2280 1213 218 103 344 270 132 

160 48 -- -- 18 81 -- 160 12 5 4 109 30 -- 

4215 2290 379 22 721 349 454 4586 2654 413 184 727 405 203 

 QA-Index. 
D-82

Table D.A.3 Definition of QA Index for Database Management 

QA-Index Definition 

1 Validated – all source data has been accessed & 

reviewed – no further action required 

2 Validated – source data may be missing some, noncritical information – no further action anticipated 

3 Validated – incomplete source data – assumptions made 

about material grade and/or exact flaw location – no 

further action anticipated 

4 Validation based on incomplete information – depending 

on application requirements, further action may be 

necessary 

5 Validation based on available, incomplete information – 

further action expected (e.g., retrieval of additional 

source data) 

6 Not validated – validation is pending, or record is 

subject to deletion from database 

Table D.A.4 Summary of Through-Wall Cracks in BWR Reactor Recirculation Piping23

EID Date of 

Detection Plant NPS Comment 

1978 06-14-1978 Duane Arnold 10 LER 50-331/78-030. Active, at power leak 

1067 01-26-1983 Brunswick-1 12 Weld B32-RR-12-BR-H4

2419 02-12-1980 Santa Maria de Garona 10 Active, 0.8 gpm leak 

1397 02-21-1985 Browns Ferry-2 12 Riser-to-manifold weld 

404 11-02-1982 Monticello 12 LER 50-220/82-013; Riser-to-safe-end weld 

7414 01-15-1986 Hatch-1 12 Weld 1B31-1RC-12BR-B-3 

3231 11-22-1982 Monticello 12 Leak detected during hydrostatic testing 

2681 04-01-1985 Quad Cities-2 12 Weld 02M-F7 

1848 11-05-1986 Quad Cities-2 12 Weld 02K-S3 

2085 10-20-1982 Monticello 22 Pipe-to-safe-end weld

443 10-31-1982 Monticello 22 Pipe-to-safe-end weld 

2850 11-06-1982 Hatch-1 22 LER 50-321/82-089; End cap-to-manifold 

3217 03-10-1985 Duane Arnold 22 Weld RHB-J1 

7543 03-23-1982 Nine Mile Point-1 28 LER 50-220/82-009; Weld P32-FW-17-W 

7542 03-23-1982 Nine Mile Point-1 28 Weld P32-FW-22-W 

7538 04-15-1982 Nine Mile Point-1 28 Weld P32-FW-27-W 

437 04-15-1985 Nine Mile Point-1 28 Weld P32-FW-42-W 

3224 03-10-1985 Duane Arnold 28 Weld RCB-J36 

2839 07-18-1985 Brunswick-1 28 Weld 1B32-RR-28-A-4 

2838 07-18-1985 Brunswick-1 28 Weld 1B32-RR-28-A-15

2837 07-18-1985 Brunswick-1 28 Weld 1B32-RR-28-B-8 

2836 07-18-1985 Brunswick-1 28 Weld 1B32-RR-28-B-4 

1711 07-18-1985 Brunswick-1 28 Weld 1B32-RR-28-A-14

3183 01-09-1986 Brunswick-2 28 Weld 2B32-RR-28-B-3 

3182 01-09-1986 Brunswick-2 28 Weld 2B32-RR-28-B-4 

3181 01-09-1986 Brunswick-2 28 Weld 2B32-RR-28-B-5 

3180 01-09-1986 Brunswick-2 28 Weld 2B32-RR-28-B-11

1723 01-09-1986 Brunswick-2 28 Weld 2B32-RR-28-A-4 

 

23 Basis for Table D.5 (Section D.4). 
D-83

ATTACHMENT B TO APPENDIX D 

BASIS FOR LOCA FREQUENCY MODELS

Attached as Tables D.B.1 (RR System Loop B) and D.B.2 (FW System Loop B) are the Excel spreadsheets 

on which the BWR LOCA frequency model is based. Attached as Tables D.B.3 (RC-HL), D.B.4 (RC Surge 

Line), and D.B.5 (HPI/NMU Line) are the spreadsheets on which the PWR LOCA frequency model is based. 

The input to the calculation of a selection of posterior weld failure rates is summarized in Tables D.B.6 

(BWR-1, NPS28), D.B.7 (BWR-2, NPS12), D.B.8 (PWR-1), D.B.9 (PWR-2), and D.B.10 (PWR-3). 
Table D.B.1 BWR-1 – RR System Loop B Weld List 

Category

Item 

Line 

Number 

Weld

Order

Component ID NPS 

(in) 

Wall Thk (in) or 

Schedule 

Configuration Description 

B5.10 5358-5(2) 5 101-304E 12 Schd. 80 Nozzle-safe-end RRI Nozzle-to-Safe End Butt Weld 

(N2E) 

B5.10 5358-5(5) 5 2-303A 12 Schd. 80 Nozzle-safe-end RRI Nozzle-to-Safe End Butt Weld 

(N2A) 

B5.10 5358-5(4) 5 2-303B 12 Schd. 80 Nozzle-safe-end RRI Nozzle-to-Safe End Butt Weld 

(N2B) 

B5.10 5358-5(6) 4 2-303C 12 Schd. 80 Nozzle-safe-end RRI Nozzle-to-Safe End Butt Weld 

(N2C) 

B5.10 5358-5(3) 5 2-303D 12 Schd. 80 Nozzle-safe-end RRI Nozzle-to-Safe End Butt Weld 

(N2D) 

B9.11 5358-5(5) 1 FW-RD-2-B10 12 Schd. 80 Pipe-sweepolet B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(4) 1 FW-RD-2-B11 12 Schd. 80 Pipe-sweepolet B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(6) 0.5 FW-RD-2-B12 12 Schd. 80 Pipe-reducer B9.11 5358-5(3) 1 FW-RD-2-B13 12 Schd. 80 Pipe-sweepolet B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(2) 1 FW-RD-2-B14 12 Schd. 80 Pipe-sweepolet B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(5) 4 FW-RD-2-B15 12 Schd. 80 Pipe-safe-end B9.11 5358-5(4) 4 FW-RD-2-B16 12 Schd. 80 Pipe-safe-end B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(6) 3 FW-RD-2-B17 12 Schd. 80 Pipe-safe-end B9.11 5358-5(3) 4 FW-RD-2-B18 12 Schd. 80 Pipe-safe-end B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(2) 4 FW-RD-2-B19 12 Schd. 80 Pipe-safe-end B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(5) 2 SW-RD-2-B4-W1 12 Schd. 80 Pipe-elbow B9.11 5358-5(5) 3 SW-RD-2-B4-W2 12 Schd. 80 Elbow-pipe B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(4) 2 SW-RD-2-B5-W1 12 Schd. 80 Pipe-elbow B9.11 5358-5(4) 3 SW-RD-2-B5-W2 12 Schd. 80 Elbow-pipe B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(3) 2 SW-RD-2-B6-W1 12 Schd. 80 Pipe-elbow B9.11 5358-5(3) 3 SW-RD-2-B6-W2 12 Schd. 80 Elbow-pipe B9.11 5358-5(2) 2 SW-RD-2-B7-W1 12 Schd. 80 Pipe-elbow B9.11 5358-5(2) 3 SW-RD-2-B7-W2 12 Schd. 80 Elbow-pipe B9.11 5358-5(6) 1 SW-RD-2-B8-W1 12 Schd. 80 Pipe-elbow B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(6) 2 SW-RD-2-B8-W2 12 Schd. 80 Elbow-safe-end B31-Reactor Recirc - Loop B Circ WeldB9.31 5358-5(1) 2 SW-RD-2-B3-W1 22 Schd. 80 Pipe-sweepolet B9.31 5358-5(1) 3 SW-RD-2-B3-W2 22 Schd. 80 Pipe-sweepolet B9.31 5358-5(1) 6 SW-RD-2-B3-W3 22 Schd. 80 Pipe-sweepolet B9.31 5358-5(1) 7 SW-RD-2-B3-W4 22 Schd. 80 Pipe-sweepolet B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(1) 4 SW-RD-2-B3-W6 22 Schd. 80 Pipe-cross B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(1) 5 SW-RD-2-B3-W7 22 Schd. 80 Pipe-cross B31-Reactor Recirc - Loop B Circ Weld
Table D.B.1 BWR-1 – RR System Loop B Weld List 

Category

Item 

Line 

Number 

Weld

Order

Component ID NPS 

(in) 

Wall Thk (in) or 

Schedule 

Configuration Description 

B9.11 5358-5(1) 1 SW-RD-2-B3-W8 22 Schd. 80 Pipe-end-cap B31-Reactor Recirc - Loop B Circ WeldB9.11 5358-5(1) 8 SW-RD-2-B3-W9 22 Schd. 80 Pipe-end-cap B5.10 5359-5(S) 1 4-303B 28 Schd. 80 Nozzle-safe-end RRS Nozzle-to-Safe End Butt Welds 

(N1B) 

B9.11 5359-5(D)-

5358-5(6) 

3 FW-RD-2-B1-W1 28 Schd. 80 Pipe-pipe B31-Reactor Recirc - Loop B Circ Weld

B9.11 5359-5(D)-

5358-5(6) 

8 FW-RD-2-B2-W2 28 Schd. 80 Pipe-sweepolet B31-Reactor Recirc - Loop B Circ Weld 

- 28 x 4 inch SWOL 

B9.11 5359-5(D)-

5358-5(6) 

1 FW-RD-2-B6 28 Schd. 80 Pipe-pump RR Pump discharge 

B9.11 5359-5(D)-

5358-5(6) 

4 FW-RD-2-B7 28 Schd. 80 Pipe-valve B31-Reactor Recirc - Loop B Circ Weld

B9.11 5359-5(D)-

5358-5(6) 

5 FW-RD-2-B8 28 Schd. 80 Pipe-valve 

B9.11 5359-5(D)-

5358-5(6) 

11 FW-RD-2-B9 28 Schd. 80 Cross-tee 

B9.11 5359-5(S) 2 FW-RS-2-B1 28 Schd. 80 Pipe-safe-end RPV Nozzle area B9.11 5359-5(S) 5 FW-RS-2-B2 28 Schd. 80 Pipe-pipe B31-Reactor Recirc - Loop B Circ WeldB9.11 5359-5(S) 9 FW-RS-2-B3 28 Schd. 80 Pipe-valve B9.11 5359-5(S) 10 FW-RS-2-B4 28 Schd. 80 Pipe-valve B31-Reactor Recirc - Loop B Circ WeldB9.11 5359-5(S) 14 FW-RS-2-B5 28 Schd. 80 Pipe-pump RR Pump suction B9.31 5359-5(D)-

5358-5(6) 

2 SW-RD-2-B1-W1 28 Schd. 80 Pipe-pipe RPV Nozzle area

B9.11 5359-5(D)-

5358-5(6) 

6 SW-RD-2-B2-W1 28 Schd. 80 Elbow-pipe 

B9.11 5359-5(D)-

5358-5(6) 

10 SW-RD-2-B2-W2 28 Schd. 80 Pipe-tee B31-Reactor Recirc - Loop B Circ Weld

B9.11 5359-5(D)-

5358-5(6) 

9 SW-RD-2-B2-W2O 28 Schd. 80 Pipe-pipe B31-Reactor Recirc - Loop B Circ Weld

B9.31 5359-5(D)-

5358-5(6) 

7 SW-RD-2-B2-W3 28 Schd. 80 Pipe-pipe 

B9.11 5359-5(D)-

5358-5(6) 

99 SW-RD-2-B3-W5 28 Schd. 80 Cross-reducer B31-Reactor Recirc - Loop B Circ Weld

B9.11 5359-5(S) 3 SW-RS-2-B1-W1 28 Schd. 80 Elbow-pipe B31-Reactor Recirc - Loop B Circ WeldB9.11 5359-5(S) 4 SW-RS-2-B1-W2 28 Schd. 80 Elbow-pipe B9.11 5359-5(S) 8 SW-RS-2-B2-W1 28 Schd. 80 Elbow-pipe B9.11 5359-5(S) 7 SW-RS-2-B2-W10A 28 Schd. 80 Pipe-tee B31-Reactor Recirc - Loop B Circ WeldB9.11 5359-5(S) 6 SW-RS-2-B2-W2 28 Schd. 80 Pipe-tee 
Table D.B.1 BWR-1 – RR System Loop B Weld List 

Category

Item 

Line 

Number 

Weld

Order

Component ID NPS 

(in) 

Wall Thk (in) or 

Schedule 

Configuration Description 

B9.11 5359-5(S) 13 SW-RS-2-B3-W1 28 Schd. 80 Elbow-pipe B31-Reactor Recirc - Loop B Circ WeldB9.31 5359-5(S) 12 SW-RS-2-B3-W2 28 Schd. 80 Pipe-pipe B31-Reactor Recirc - Loop B Circ WeldB9.31 5359-5(S) 11 SW-RS-2-B3-W3 28 Schd. 80 Pipe-pipe B9.11 5359-5(S) 11.1 SW-RS-2-B3-W4 28 Schd. 80 Pipe-pipe B9.11 5359-5(S) 11.2 SW-RS-2-B3-W5 28 Schd. 80 Pipe-pipe 
Table D.B.2 BWR-2 – FW System Loop B Weld List 

ategory 

Item

Line 

Number

Weld 

Order

Component ID NPS 

(in)

Wall Thk (in) or 

Schedule

Configuration Description

.11 3537-5(4) 2 SW-N21-2336-19WF 12 Schd. 100 Elbow-pipe .11 3537-5(4) 3 SW-N21-2336-19WG 12 Schd. 100 Elbow-pipe .11 3537-5(4) 4 FW-N21-2336-19W20 12 Schd. 100 Elbow-pipe .11 3537-5(4) 7 FW-N21-2336-20WF4 12 Schd. 100 Elbow-pipe .11 3537-5(4) 8 SW-N21-2336-20WM 12 Schd. 100 Elbow-pipe .11 3537-5(5) 2 SW-N21-2336-17WB 12 Schd. 100 Elbow-pipe .11 3537-5(5) 3 SW-N21-2336-17WD 12 Schd. 100 Elbow-pipe .11 3537-5(5) 4 FW-N21-2336-17W18 12 Schd. 100 Elbow-pipe .11 3537-5(5) 6 SW-N21-2336-18WP 12 Schd. 100 Elbow-pipe .11 3537-5(5) 7 SW-N21-2336-18WQ 12 Schd. 100 Elbow-pipe .11 3537-5(5) 8 FW-N21-2336-18W0 12 Schd. 100 Elbow-pipe Transition piece .11 3537-5(6) 2 SW-N21-2336-14WB 12 Schd. 100 Elbow-pipe .11 3537-5(6) 6 SW-N21-2336-15WP 12 Schd. 100 Elbow-pipe N21-2336-Feedwater Loop A Circ 

Weld 

.11 3537-5(6) 7 FW-N21-2336-15WF2 12 Schd. 100 Elbow-pipe .11 3537-5(5) 1 FW-N21-2336-16W17 12 Schd. 100 Elbow-reducingtee .11 3537-5(6) 1 FW-N21-2336-13W14 12 Schd. 100 Elbow-reducingtee N21-2336-Feedwater Loop A Circ Weld .11 3537-5(4) 11 3-316C 12 Schd. 100 Nozzle-safe-end N21-2336-Feedwater Loop A Circ 

Weld 

.11 3537-5(5) 10 3-316B 12 Schd. 100 Nozzle-safe-end N21-2336-Feedwater Loop A Circ 

Weld 

.11 3537-5(6) 10 3-316A 12 Schd. 100 Nozzle-safe-end N21-2336-Feedwater Loop A Circ 

Weld 

.11 3537-5(4) 5 FW-N21-2336-20WF2 12 Schd. 100 Pipe-pipe .11 3537-5(4) 6 FW-N21-2336-20WF3 12 Schd. 100 Pipe-pipe .11 3537-5(5) 5 FW-N21-2336-18WF1 12 Schd. 100 Pipe-pipe .11 3537-5(6) 3 FW-N21-2336-14WF1 12 Schd. 100 Pipe-pipe N21-2336-Feedwater Loop A Circ 

Weld 

.11 3537-5(6) 4 FW-N21-2336-14W15 12 Schd. 100 Pipe-pipe N21-2336-Feedwater Loop A Circ 

Weld 

.11 3537-5(6) 5 FW-N21-2336-15WF1 12 Schd. 100 Pipe-pipe .11 3537-5(4) 1 FW-N21-2336-16W19 12 Schd. 100 Pipe-reducer N21-2336-Feedwater Loop A Circ 

Weld 

.11 3537-5(4) 9 FW-N21-2336-20W0 12 Schd. 100 Pipe-safe-end 
Table D.B.2 BWR-2 – FW System Loop B Weld List 

ategory 

Item

Line 

Number

Weld 

Order

Component ID NPS 

(in)

Wall Thk (in) or 

Schedule

Configuration Description

.11 3537-5(6) 8 FW-N21-2336-15W0 12 Schd. 100 Pipe-safe-end N21-2336-Feedwater Loop A Circ 

Weld 

.11 3537-5(4) 10 N4-C 12 Schd. 100 Safe-end-Safe-end.11 3537-5(5) 9 N4-B 12 Schd. 100 Safe-end-Safe-end.11 3537-5(6) 9 N4A 12 Schd. 100 Safe-end-Safe-end N21-2336-Feedwater Loop A Circ 

Weld 

.11 3537-5(2) 3 SW-N21-2336-11WD 20 Schd. 100 Elbow-pipe .11 3537-5(3) 8 FW-N21-2336-2WB 20 Schd. 100 Elbow-pipe .11 3537-5(2) 4 SW-N21-2336-11WE 20 Schd. 100 Elbow-pipe .11 3537-5(3) 12 SW-N21-2336-13WC 20 Schd. 100 Elbow-reducingtee N21-2336-Feedwater Loop A Circ Weld .11 3537-5(3) 7 FW-N21-2336-0W02 20 Schd. 100 Elbow-valve .11 3537-5(3) 11 SW-N21-2336-13WB 20 Schd. 100 Elbow-valve .11 3537-5(3) 5 SW-X9A-W1 20 Schd. 100 Penetration Longitudinal weld .11 3537-5(3) 4 FW-N21-2336-11W01 20 Schd. 100 Pipe-penetration .11 3537-5(2) 2 FW-N21-2336-11WF1 20 Schd. 100 Pipe-pipe .11 3537-5(2) 5 FW-N21-2336-11WF2 20 Schd. 100 Pipe-pipe .11 3537-5(2) 6 FW-N21-2336-11WF3 20 Schd. 100 Pipe-pipe .11 3537-5(3) 14 FW-N21-2336-3AW13 20 Schd. 100 Pipe-pipe .11 3537-5(3) 15 FW-N21-2336-3W03A 20 Schd. 100 Pipe-pipe .11 3537-5(3) 16 FW-N21-2336-3W16 20 Schd. 100 Pipe-pipe .11 3537-5(2) 1 SW-N21-2336-11WB 20 Schd. 100 Pipe-reducer .11 3537-5(3) 13 SW-N21-2336-13WE 20 Schd. 100 Pipe-reducing-tee N21-2336-Feedwater Loop A Circ 

Weld 

.11 3537-5(3) 17 SW-N21-2336-16WC 20 Schd. 100 Pipe-reducing-tee.11 3537-5(2) 7 SW-N21-2336-11WN 20 Schd. 100 Pipe-tee .11 3537-5(3) 3 SW-N21-2336-11WH 20 Schd. 100 Pipe-tee .11 3537-5(3) 1 FW-N21-2336-1VW11 20 Schd. 100 Pipe-valve .11 3537-5(3) 6 FW-N21-2336-1VW12 20 Schd. 100 Pipe-valve .11 3537-5(3) 9 FW-N21-2336-2W0 20 Schd. 100 Pipe-valve N21-2336-Feedwater Loop A Circ 

Weld 

.11 3537-5(3) 10 FW-N21-2336-0W13 20 Schd. 100 Pipe-valve N21-2336-Feedwater Loop A Circ 

Weld 

.11 3537-5(3) 18 SW-N21-2336-16WE 20 Schd. 100 Reducer-reducingtee .11 3537-5(3) 2 FW-N21-2336-0W11 20 Schd. 100 Tee-valve 
Table D.B.3 PWR-1 – Reactor Coolant System Hot Leg

on

y

Category

Item

Component 

ID

Configuration NPS 

(in)

Wall Thk

(in)

Weld Material

B9.11 1-4100A- 8 ELBOW TO PIPE 31.00 2.600 304N/351CF B9.11 1-4100A- 9 ELBOW TO PIPE 31.00 2.625 SS B9.11 1-4100A- 10 ELBOW TO PIPE 31.00 2.625 SS B9.11 1-4100A- 11 PIPE TO ELBOW 31.00 2.625 SS B9.11 1-4100A- 12 ELBOW TO PUMP 31.00 2.625 SS B9.11 1-4100A- 13 PIPE TO PUMP 27.50 2.375 304NB9.11 1-4100A- 14 PIPE TO ELBOW 27.50 2.375 SS B9.11 1-4100A- 15 BIMETAL (INCONEL) WELD.ELBOW 

TO SAFE END 

27.50 2.375 SS/INCONEL 

B5.10 1-4100A- 16(DM) BIMETAL (INCONEL) WELD. R.V. 

LOOP A INLET NOZZLE TO SAFE END

27.50 2.375 CS/SS 
D-90

Table D.B.4 PWR-2 – Pressurizer Surge Line 

ASME XI 

Examination 

Category 

Component 

ID 

Description / Configuration NPS 

[inch] 

Wall Thk

[inch] 

Weld Material

B-J 1-4100A- 19BC 14" Branch (CGE-1-4500) Connection 

to 29-inch pipe 

14.00 2.350 304N/SA182 

B-F 1-4500A- 1(DM) Bimetal (INCONEL) weld; 

Pressurizer Surge Line Nozzle to Safe 

End 

14.00 1.406 SS 

B-J 1-4500A- 2 Bimetal (INCONEL) Weld: Safe End to 

Pipe 

14.00 1.406 SS 

B-J 1-4500A- 3 Pipe-to-Elbow 14.00 1.406 SS 

B-J 1-4500A- 4 Pipe-to-Elbow 14.00 1.406 SS 

B-J 1-4500A- 5 Pipe-to-Elbow 14.00 1.406 SS 

B-J 1-4500A- 6 Pipe-to-Elbow 14.00 1.406 SS 

B-J 1-4500A- 7 Pipe-to-Elbow 14.00 1.406 SS 

B-J 1-4500A- 8 Pipe-to-Elbow 14.00 1.406 SS 

B-J 1-4500A- 9 Pipe-to-Elbow 14.00 1.406 SS 

B-J 1-4500A- 10 Pipe-to-Elbow 14.00 1.406 SS 

B-J 1-4500A- 11 Pipe-to-Elbow 14.00 1.406 SS 

B-J 1-4500A- 12 Pipe-to-Elbow 14.00 1.406 SS 

B-J 1-4500A- 13 Pipe-to-Branch Connection 14.00 1.406 SS 

Table D.B.5 PWR-5 – HPI/NMU Line 

Weld ID Configuration Inside 

Diameter

17-MU-23-73AR Elbow-valve 2.50 

17-MU-23-21-055 Elbow-pipe 2.50 

17-MU-23-21-057 Elbow-pipe 2.50 

17-MU-23-21-058 Elbow-pipe 2.50 

17-MU-23-21-059 Elbow-valve 2.50 

17-MU-23-21-059B Elbow-valve 2.50 

17-MU-23-21-061 Pipe-pipe 2.50 

17-MU-23-21-062 Elbow-pipe 2.50 

17-MU-23-21-063 Elbow-nozzle 2.50 
Table D.B.6 BWR-1 Weld Failure Rate Calculation Sheet 

Mean 

(Prior)24

Range 

Factor 

Failures 

(Evidence) 

Exposure 

[Weld-Yr] 

Weld 

Count25 Mean 5th 50th 95th 

Range 

Factor Calc Date 

7.19E-04 10 0 4,725 5 1.28E-04 1.41E-05 8.78E-05 3.70E-04 5.1 5/13/2003 15:16 

7.19E-04 10 0 9,450 10 8.37E-05 1.10E-05 6.09E-05 2.29E-04 4.6 5/13/2003 15:16 

7.19E-04 10 0 14,175 15 6.42E-05 9.28E-06 4.81E-05 1.70E-04 4.3 5/13/2003 15:16 

2.25E-04 10 0 3,780 4 8.20E-05 6.29E-06 4.84E-05 2.66E-04 6.5 5/13/2003 15:16 

2.25E-04 10 0 7,560 8 5.86E-05 5.41E-06 3.75E-05 1.80E-04 5.8 5/13/2003 15:16 

2.25E-04 10 0 11,340 12 4.71E-05 4.84E-06 3.15E-05 1.40E-04 5.4 5/13/2003 15:16 

1.50E-04 10 0 1,890 2 8.34E-05 4.89E-06 4.27E-05 2.93E-04 7.7 5/13/2003 15:16 

1.50E-04 10 0 3,780 4 6.49E-05 4.49E-06 3.64E-05 2.18E-04 7.0 5/13/2003 15:16 

1.50E-04 10 0 5,670 6 5.46E-05 4.20E-06 3.23E-05 1.78E-04 6.5 5/13/2003 15:16 6.74E-04 10 0 1,890 2 2.03E-04 1.74E-05 1.26E-04 6.39E-04 6.1 5/13/2003 15:16 

6.74E-04 10 0 3,780 4 1.41E-04 1.45E-05 9.43E-05 4.19E-04 5.4 5/13/2003 15:16 

6.74E-04 10 0 5,670 6 1.12E-04 1.28E-05 7.76E-05 3.20E-04 5.0 5/13/2003 15:16 

4.99E-05 20 0 8,505 9 1.75E-05 3.99E-07 6.56E-06 7.03E-05 13.3 5/13/2003 15:16 

4.99E-05 20 0 17,010 18 1.27E-05 3.64E-07 5.45E-06 4.91E-05 11.6 5/13/2003 15:16 

4.99E-05 20 0 25,515 27 1.04E-05 3.39E-07 4.77E-06 3.88E-05 10.7 5/13/2003 15:16 2.25E-04 10 0 1,890 2 1.09E-04 7.01E-06 5.88E-05 3.73E-04 7.3 5/13/2003 15:16 

2.25E-04 10 0 3,780 4 8.20E-05 6.29E-06 4.84E-05 2.66E-04 6.5 5/13/2003 15:16 

2.25E-04 10 0 5,670 6 6.78E-05 5.79E-06 4.20E-05 2.13E-04 6.1 5/13/2003 15:16 

1.25E-04 10 0 1,890 2 7.36E-05 4.15E-06 3.68E-05 2.61E-04 7.9 5/13/2003 15:16 1.25E-04 10 0 3,780 4 5.81E-05 3.84E-06 3.18E-05 1.98E-04 7.2 5/13/2003 15:16 

1.25E-04 10 0 5,670 6 4.93E-05 3.61E-06 2.85E-05 1.63E-04 6.7 5/13/2003 15:16 

7.49E-05 20 0 945 1 4.63E-05 6.74E-07 1.27E-05 1.95E-04 17.0 5/13/2003 15:16 

7.49E-05 20 0 1,890 2 3.79E-05 6.51E-07 1.17E-05 1.58E-04 15.6 5/13/2003 15:16 

7.49E-05 20 0 2,835 3 3.29E-05 6.33E-07 1.10E-05 1.36E-04 14.6 5/13/2003 15:16 7.49E-05 20 0 945 1 4.63E-05 6.74E-07 1.27E-05 1.95E-04 17.0 5/13/2003 15:16 

7.49E-05 20 0 1,890 2 3.79E-05 6.51E-07 1.17E-05 1.58E-04 15.6 5/13/2003 15:16 

7.49E-05 20 0 2,835 3 3.29E-05 6.33E-07 1.10E-05 1.36E-04 14.6 5/13/2003 15:16 

ken from PlantBWelds.xls (see Section 1.3); Table B-1 is an excerpt from this Excel-file. 
Table D.B.7 BWR-2 Weld Failure Rate Calculation Sheet 

Mean 

(Prior)

Range 

Factor

Failures 

(Evidence)

Exposure 

[Weld-Yr] Welds Mean 5th 50th 95th

Range Factor Calc Date

1.26E-04 10 0 2,835 3 6.47E-05 4.00E-06 3.42E-05 2.24E-04 7.5 5/13/2003 15:05 

1.26E-04 10 0 5,670 6 4.95E-05 3.63E-06 2.86E-05 1.63E-04 6.7 5/13/2003 15:05 

1.26E-04 10 0 8,505 9 4.12E-05 3.36E-06 2.50E-05 1.32E-04 6.3 5/13/2003 15:05 

6.73E-06 20 0 13,230 14 3.92E-06 6.00E-08 1.11E-06 1.64E-05 16.6 5/13/2003 15:05 

6.73E-06 20 0 26,460 28 3.15E-06 5.77E-08 1.02E-06 1.31E-05 15.1 5/13/2003 15:05 

6.73E-06 20 0 39,690 42 2.71E-06 5.58E-08 9.51E-07 1.11E-05 14.1 5/13/2003 15:05 

1.05E-05 15 0 5,670 6 7.48E-06 1.71E-07 2.45E-06 3.01E-05 13.3 5/13/2003 15:05 

1.05E-05 15 0 11,340 12 6.33E-06 1.66E-07 2.29E-06 2.53E-05 12.3 5/13/2003 15:05 

1.05E-05 15 0 17,010 18 5.61E-06 1.62E-07 2.17E-06 2.21E-05 11.7 5/13/2003 15:05 

3.14E-05 15 0 945 1 2.54E-05 5.23E-07 7.64E-06 1.02E-04 13.9 5/13/2003 15:05 

3.14E-05 15 0 1,890 2 2.24E-05 5.14E-07 7.34E-06 9.04E-05 13.3 5/13/2003 15:05 

3.14E-05 15 0 2,835 3 2.05E-05 5.06E-07 7.09E-06 8.22E-05 12.7 5/13/2003 15:05 

8.80E-06 20 0 4,725 5 6.16E-06 8.05E-08 1.55E-06 2.57E-05 17.9 5/13/2003 15:05 

8.80E-06 20 0 9,450 10 5.22E-06 7.86E-08 1.47E-06 2.19E-05 16.7 5/13/2003 15:05 

8.80E-06 20 0 14,175 15 4.63E-06 7.71E-08 1.40E-06 1.94E-05 15.9 5/13/2003 15:05 9.43E-06 20 0 1,890 2 7.63E-06 8.76E-08 1.72E-06 3.12E-05 18.9 5/13/2003 15:05 

9.43E-06 20 0 3,780 4 6.80E-06 8.65E-08 1.67E-06 2.83E-05 18.1 5/13/2003 15:05 

9.43E-06 20 0 5,670 6 6.24E-06 8.56E-08 1.63E-06 2.62E-05 17.5 5/13/2003 15:05 
Table D.B.8 PWR-1 Weld Failure Rate Calculation Sheet 

Mean 

(Prior)26

Range 

Factor 

Failures 

(Evidence) 

Exposure 

[Weld-Yr] 

Weld 

Count Mean 5th 50th 95th 

Range 

Factor Calc Date 

3.65E-06 100 0 6,720 20 1.49E-06 7.05E-10 6.85E-08 5.61E-06 89.2 5/2/2003 9:25 

3.65E-06 100 0 13,440 40 1.19E-06 6.96E-10 6.65E-08 4.88E-06 83.7 5/2/2003 9:25 

3.65E-06 100 0 20,160 60 1.02E-06 6.89E-10 6.49E-08 4.40E-06 79.9 5/2/2003 9:25 4.06E-05 100 0 504 2 1.75E-05 7.86E-09 7.67E-07 6.42E-05 90.4 5/2/2003 9:25 

4.06E-05 100 0 1,008 3 1.40E-05 7.77E-09 7.47E-07 5.66E-05 85.3 5/2/2003 9:25 

4.06E-05 100 0 1,512 5 1.22E-05 7.71E-09 7.30E-07 5.14E-05 81.7 5/2/2003 9:25 8.12E-04 100 0 504 2 1.06E-04 1.42E-07 1.19E-05 5.07E-04 59.8 5/2/2003 9:25 

8.12E-04 100 0 1,008 3 7.35E-05 1.33E-07 1.04E-05 3.53E-04 51.4 5/2/2003 9:25 

8.12E-04 100 0 1,512 5 5.85E-05 1.28E-07 9.41E-06 2.80E-04 46.8 5/2/2003 9:25 8.12E-05 100 0 504 2 2.81E-05 1.55E-08 1.49E-06 1.13E-04 85.3 5/2/2003 9:25 

8.12E-05 100 0 1,008 3 2.18E-05 1.53E-08 1.43E-06 9.49E-05 78.8 5/2/2003 9:25 

8.12E-05 100 0 1,512 5 1.85E-05 1.51E-08 1.39E-06 8.34E-05 74.3 5/2/2003 9:25 

5.07E-05 100 0 504 2 2.04E-05 9.78E-09 9.51E-07 7.74E-05 89.0 5/2/2003 9:25 

5.07E-05 100 0 1,008 3 1.62E-05 9.66E-09 9.22E-07 6.72E-05 83.4 5/2/2003 9:25 

5.07E-05 100 0 1,512 5 1.40E-05 9.57E-09 8.99E-07 6.04E-05 79.4 5/2/2003 9:25 
Table D.B.9 PWR-2 Weld Failure Rate Calculation Sheet 

Mean 

(Prior) 

Range 

Factor 

Failures 

(Evidence) Exposure 

Welds 

Count Mean 5th 50th 95th 

Range 

Factor Calc Date 

6.70E-07 100 0 18,105 5 3.33E-07 1.30E-10 1.28E-08 1.13E-06 93.3 5/2/2003 10:14 

6.70E-07 100 0 36,210 10 2.74E-07 1.29E-10 1.26E-08 1.03E-06 89.3 5/2/2003 10:14 

6.70E-07 100 0 54,315 15 2.41E-07 1.29E-10 1.24E-08 9.57E-07 86.3 5/2/2003 10:14 

3.75E-05 100 0 1,811 1 1.08E-05 7.10E-09 6.70E-07 4.62E-05 80.7 5/2/2003 10:14 

3.75E-05 100 0 3,621 1 8.18E-06 6.94E-09 6.34E-07 3.72E-05 73.2 5/2/2003 10:14 

3.75E-05 100 0 5,432 2 6.84E-06 6.81E-09 6.06E-07 3.19E-05 68.4 5/2/2003 10:14 3.35E-06 100 0 1,811 1 1.96E-06 6.55E-10 6.48E-08 6.03E-06 96.0 5/2/2003 10:14 

3.35E-06 100 0 3,621 1 1.67E-06 6.52E-10 6.41E-08 5.67E-06 93.3 5/2/2003 10:14 

3.35E-06 100 0 5,432 2 1.49E-06 6.49E-10 6.35E-08 5.39E-06 91.1 5/2/2003 10:14 

4.69E-05 100 0 1,811 1 1.24E-05 8.82E-09 8.25E-07 5.42E-05 78.4 5/2/2003 10:14 

4.69E-05 100 0 3,621 1 9.28E-06 8.60E-09 7.74E-07 4.28E-05 70.6 5/2/2003 10:14 

4.69E-05 100 0 5,432 2 7.71E-06 8.42E-09 7.37E-07 3.63E-05 65.7 5/2/2003 10:14 3.35E-06 100 0 1,811 1 1.96E-06 6.55E-10 6.48E-08 6.03E-06 96.0 5/2/2003 10:14 

3.35E-06 100 0 3,621 1 1.67E-06 6.52E-10 6.41E-08 5.67E-06 93.3 5/2/2003 10:14 

3.35E-06 100 0 5,432 2 1.49E-06 6.49E-10 6.35E-08 5.39E-06 91.1 5/2/2003 10:14 
Table D.B.10 PWR-3 Weld Failure Rate Calculation Sheet 

Mean 

(Prior)

Range 

Factor

Failures 

(Evidence) Exposure

Weld 

Count Mean 5th 50th 95th

Range 

Factor Calc Date

1.90E-06 97 0 140 4 1.69E-06 4.05E-10 3.93E-08 3.79E-06 96.7 5/2/2003 14:14 

1.90E-06 97 0 280 8 1.61E-06 4.05E-10 3.92E-08 3.77E-06 96.5 5/2/2003 14:14 

1.90E-06 97 0 420 12 1.54E-06 4.05E-10 3.92E-08 3.75E-06 96.3 5/2/2003 14:14 

1.31E-06 98 0 105 3 1.21E-06 2.69E-10 2.64E-08 2.59E-06 98.2 5/2/2003 14:14 

1.31E-06 98 0 210 6 1.16E-06 2.69E-10 2.64E-08 2.59E-06 98.1 5/2/2003 14:14 

1.31E-06 98 0 315 9 1.13E-06 2.69E-10 2.64E-08 2.58E-06 98.0 5/2/2003 14:14 

6.48E-06 100 0 35 1 5.81E-06 1.28E-09 1.27E-07 1.27E-05 99.6 5/2/2003 14:14 

6.48E-06 100 0 70 2 5.52E-06 1.28E-09 1.27E-07 1.26E-05 99.4 5/2/2003 14:14 

6.48E-06 100 0 105 3 5.32E-06 1.28E-09 1.27E-07 1.26E-05 99.2 5/2/2003 14:14 9.86E-04 16 0 35 1 7.59E-04 1.37E-05 2.17E-04 3.09E-03 15.0 5/2/2003 14:15 

9.86E-04 16 0 70 2 6.61E-04 1.34E-05 2.07E-04 2.70E-03 14.2 5/2/2003 14:15 

9.86E-04 16 0 105 3 5.97E-04 1.32E-05 2.00E-04 2.43E-03 13.6 5/2/2003 14:15 
D-96

ATTACHMENT C TO APPENDIX D 

BASIS FOR LOCA FREQUENCY MODELS

Attached as Table D.C.1 is an Excel spreadsheet used to calculate time-dependent LOCA frequencies. Table 

D.C.1 includes the parameters input to the BWR-1 Cat1 LOCA frequency calculation. 
Table D.C.1 Application of Markov Model to BWR-1 Cat 1 LOCA Frequency 

 CASE INSPECTION INDEPENDENT PARAMETERS

INSPECTION DEPENDENT 

PARAMETERS 

Leak 

Inspection 

Interval PLI TSUBLI MU PHI RHO_F RHO_L Lambda TFI TSUBR PFI PFD OMEGA Haz{T} 

None 0.00 1.50 0.0000 3.34E-03 5.96E-07 2.00E-02 4.40E-04 10.00 0.023 0.000 0.000 0.000 7.62E-06 

RF 0.90 1.50 0.5910 3.34E-03 5.96E-07 2.00E-02 4.40E-04 10.00 0.023 0.000 0.000 0.000 1.12E-06 

Wk 0.90 1.92E-02 21.3971 3.34E-03 5.96E-07 2.00E-02 4.40E-04 10.00 0.023 0.000 0.000 0.000 8.01E-08 

Shift 0.90 9.13E-04 37.9048 3.34E-03 5.96E-07 2.00E-02 4.40E-04 10.00 0.023 0.000 0.000 0.000 6.59E-08 

None 0.00 1.50 0.0000 3.34E-03 5.96E-07 2.00E-02 4.40E-04 10.00 0.023 1.000 0.500 0.050 5.18E-06 

RF 0.90 1.50 0.5910 3.34E-03 5.96E-07 2.00E-02 4.40E-04 10.00 0.023 1.000 0.500 0.050 6.66E-07 

Wk 0.90 1.92E-02 21.3971 3.34E-03 5.96E-07 2.00E-02 4.40E-04 10.00 0.023 1.000 0.500 0.050 4.62E-08 

Shift 0.90 9.13E-04 37.9048 3.34E-03 5.96E-07 2.00E-02 4.40E-04 10.00 0.023 1.000 0.500 0.050 3.80E-08 

None 0.00 1.50 0.0000 3.34E-03 5.96E-07 2.00E-02 4.40E-04 10.00 0.023 1.000 0.900 0.090 4.02E-06 

RF 0.90 1.50 0.5910 3.34E-03 5.96E-07 2.00E-02 4.40E-04 10.00 0.023 1.000 0.900 0.090 4.74E-07 

Wk 0.90 1.92E-02 21.3971 3.34E-03 5.96E-07 2.00E-02 4.40E-04 10.00 0.023 1.000 0.900 0.090 3.24E-08 

Shift 0.90 9.13E-04 37.9048 3.34E-03 5.96E-07 2.00E-02 4.40E-04 10.00 0.023 1.000 0.900 0.090 2.67E-08 

None 0.00 1.50 0.0000 3.29E-03 5.88E-07 2.00E-02 4.34E-04 10.00 0.023 0.000 0.000 0.000 7.43E-06 

RF 0.90 1.50 0.5910 3.29E-03 5.88E-07 2.00E-02 4.34E-04 10.00 0.023 0.000 0.000 0.000 1.09E-06 

Wk 0.90 1.92E-02 21.3971 3.29E-03 5.88E-07 2.00E-02 4.34E-04 10.00 0.023 0.000 0.000 0.000 7.81E-08 

Shift 0.90 9.13E-04 37.9048 3.29E-03 5.88E-07 2.00E-02 4.34E-04 10.00 0.023 0.000 0.000 0.000 6.42E-08 

None 0.00 1.50 0.0000 3.29E-03 5.88E-07 2.00E-02 4.34E-04 10.00 0.023 1.000 0.500 0.050 5.05E-06 

RF 0.90 1.50 0.5910 3.29E-03 5.88E-07 2.00E-02 4.34E-04 10.00 0.023 1.000 0.500 0.050 6.49E-07 

Wk 0.90 1.92E-02 21.3971 3.29E-03 5.88E-07 2.00E-02 4.34E-04 10.00 0.023 1.000 0.500 0.050 4.50E-08 

Shift 0.90 9.13E-04 37.9048 3.29E-03 5.88E-07 2.00E-02 4.34E-04 10.00 0.023 1.000 0.500 0.050 3.71E-08 

None 0.00 1.50 0.0000 3.29E-03 5.88E-07 2.00E-02 4.34E-04 10.00 0.023 1.000 0.900 0.090 3.92E-06 

RF 0.90 1.50 0.5910 3.29E-03 5.88E-07 2.00E-02 4.34E-04 10.00 0.023 1.000 0.900 0.090 4.62E-07 

Wk 0.90 1.92E-02 21.3971 3.29E-03 5.88E-07 2.00E-02 4.34E-04 10.00 0.023 1.000 0.900 0.090 3.16E-08 

Shift 0.90 9.13E-04 37.9048 3.29E-03 5.88E-07 2.00E-02 4.34E-04 10.00 0.023 1.000 0.900 0.090 2.60E-08 

None 0.00 1.50 0.0000 7.36E-03 1.32E-06 2.00E-02 9.71E-04 10.00 0.023 0.000 0.000 0.000 3.57E-05 

RF 0.90 1.50 0.5910 7.36E-03 1.32E-06 2.00E-02 9.71E-04 10.00 0.023 0.000 0.000 0.000 5.19E-06 

Wk 0.90 1.92E-02 21.3971 7.36E-03 1.32E-06 2.00E-02 9.71E-04 10.00 0.023 0.000 0.000 0.000 3.69E-07 

Shift 0.90 9.13E-04 37.9048 7.36E-03 1.32E-06 2.00E-02 9.71E-04 10.00 0.023 0.000 0.000 0.000 3.04E-07 

None 0.00 1.50 0.0000 7.36E-03 1.32E-06 2.00E-02 9.71E-04 10.00 0.023 1.000 0.500 0.050 2.45E-05 

RF 0.90 1.50 0.5910 7.36E-03 1.32E-06 2.00E-02 9.71E-04 10.00 0.023 1.000 0.500 0.050 3.11E-06 

Wk 0.90 1.92E-02 21.3971 7.36E-03 1.32E-06 2.00E-02 9.71E-04 10.00 0.023 1.000 0.500 0.050 2.15E-07 

Shift 0.90 9.13E-04 37.9048 7.36E-03 1.32E-06 2.00E-02 9.71E-04 10.00 0.023 1.000 0.500 0.050 1.77E-07 

None 0.00 1.50 0.0000 7.36E-03 1.32E-06 2.00E-02 9.71E-04 10.00 0.023 1.000 0.900 0.090 1.91E-05 

RF 0.90 1.50 0.5910 7.36E-03 1.32E-06 2.00E-02 9.71E-04 10.00 0.023 1.000 0.900 0.090 2.23E-06 

Wk 0.90 1.92E-02 21.3971 7.36E-03 1.32E-06 2.00E-02 9.71E-04 10.00 0.023 1.000 0.900 0.090 1.52E-07 

Shift 0.90 9.13E-04 37.9048 7.36E-03 1.32E-06 2.00E-02 9.71E-04 10.00 0.023 1.000 0.900 0.090 1.25E-07 

None 0.00 1.50 0.0000 5.66E-03 1.01E-06 2.00E-02 7.47E-04 10.00 0.023 0.000 0.000 0.000 2.15E-05 
Table D.C.1 Application of Markov Model to BWR-1 Cat 1 LOCA Frequency 

 CASE INSPECTION INDEPENDENT PARAMETERS

INSPECTION DEPENDENT 

PARAMETERS 

Leak 

Inspection 

Interval PLI TSUBLI MU PHI RHO_F RHO_L Lambda TFI TSUBR PFI PFD OMEGA Haz{T} 

RF 0.90 1.50 0.5910 5.66E-03 1.01E-06 2.00E-02 7.47E-04 10.00 0.023 0.000 0.000 0.000 3.13E-06 

Wk 0.90 1.92E-02 21.3971 5.66E-03 1.01E-06 2.00E-02 7.47E-04 10.00 0.023 0.000 0.000 0.000 2.23E-07 

Shift 0.90 9.13E-04 37.9048 5.66E-03 1.01E-06 2.00E-02 7.47E-04 10.00 0.023 0.000 0.000 0.000 1.84E-07 

None 0.00 1.50 0.0000 5.66E-03 1.01E-06 2.00E-02 7.47E-04 10.00 0.023 1.000 0.500 0.050 1.47E-05 

RF 0.90 1.50 0.5910 5.66E-03 1.01E-06 2.00E-02 7.47E-04 10.00 0.023 1.000 0.500 0.050 1.87E-06 

Wk 0.90 1.92E-02 21.3971 5.66E-03 1.01E-06 2.00E-02 7.47E-04 10.00 0.023 1.000 0.500 0.050 1.30E-07 

Shift 0.90 9.13E-04 37.9048 5.66E-03 1.01E-06 2.00E-02 7.47E-04 10.00 0.023 1.000 0.500 0.050 1.07E-07 

None 0.00 1.50 0.0000 5.66E-03 1.01E-06 2.00E-02 7.47E-04 10.00 0.023 1.000 0.900 0.090 1.14E-05 

RF 0.90 1.50 0.5910 5.66E-03 1.01E-06 2.00E-02 7.47E-04 10.00 0.023 1.000 0.900 0.090 1.34E-06 

Wk 0.90 1.92E-02 21.3971 5.66E-03 1.01E-06 2.00E-02 7.47E-04 10.00 0.023 1.000 0.900 0.090 9.15E-08 

Shift 0.90 9.13E-04 37.9048 5.66E-03 1.01E-06 2.00E-02 7.47E-04 10.00 0.023 1.000 0.900 0.090 7.52E-08 

None 0.00 1.50 0.0000 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 0.000 0.000 0.000 8.70E-07 

RF 0.90 1.50 0.5910 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 0.000 0.000 0.000 1.29E-07 

Wk 0.90 1.92E-02 21.3971 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 0.000 0.000 0.000 9.23E-09 

Shift 0.90 9.13E-04 37.9048 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 0.000 0.000 0.000 7.59E-09 

None 0.00 1.50 0.0000 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.500 0.050 5.89E-07 

RF 0.90 1.50 0.5910 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.500 0.050 7.62E-08 

Wk 0.90 1.92E-02 21.3971 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.500 0.050 5.29E-09 

Shift 0.90 9.13E-04 37.9048 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.500 0.050 4.35E-09 

None 0.00 1.50 0.0000 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.900 0.090 4.55E-07 

RF 0.90 1.50 0.5910 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.900 0.090 5.40E-08 

Wk 0.90 1.92E-02 21.3971 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.900 0.090 3.70E-09 

Shift 0.90 9.13E-04 37.9048 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.900 0.090 3.04E-09 

None 0.00 1.50 0.0000 7.23E-03 1.29E-06 2.00E-02 9.53E-04 10.00 0.023 0.000 0.000 0.000 3.44E-05 

RF 0.90 1.50 0.5910 7.23E-03 1.29E-06 2.00E-02 9.53E-04 10.00 0.023 0.000 0.000 0.000 5.00E-06 

Wk 0.90 1.92E-02 21.3971 7.23E-03 1.29E-06 2.00E-02 9.53E-04 10.00 0.023 0.000 0.000 0.000 3.56E-07 

Shift 0.90 9.13E-04 37.9048 7.23E-03 1.29E-06 2.00E-02 9.53E-04 10.00 0.023 0.000 0.000 0.000 2.93E-07 

None 0.00 1.50 0.0000 7.23E-03 1.29E-06 2.00E-02 9.53E-04 10.00 0.023 1.000 0.500 0.050 2.36E-05 

RF 0.90 1.50 0.5910 7.23E-03 1.29E-06 2.00E-02 9.53E-04 10.00 0.023 1.000 0.500 0.050 3.00E-06 

Wk 0.90 1.92E-02 21.3971 7.23E-03 1.29E-06 2.00E-02 9.53E-04 10.00 0.023 1.000 0.500 0.050 2.08E-07 

Shift 0.90 9.13E-04 37.9048 7.23E-03 1.29E-06 2.00E-02 9.53E-04 10.00 0.023 1.000 0.500 0.050 1.71E-07 

None 0.00 1.50 0.0000 7.23E-03 1.29E-06 2.00E-02 9.53E-04 10.00 0.023 1.000 0.900 0.090 1.84E-05 

RF 0.90 1.50 0.5910 7.23E-03 1.29E-06 2.00E-02 9.53E-04 10.00 0.023 1.000 0.900 0.090 2.15E-06 

Wk 0.90 1.92E-02 21.3971 7.23E-03 1.29E-06 2.00E-02 9.53E-04 10.00 0.023 1.000 0.900 0.090 1.47E-07 

Shift 0.90 9.13E-04 37.9048 7.23E-03 1.29E-06 2.00E-02 9.53E-04 10.00 0.023 1.000 0.900 0.090 1.21E-07 

None 0.00 1.50 0.0000 5.15E-03 9.20E-07 2.00E-02 6.80E-04 10.00 0.023 0.000 0.000 0.000 1.78E-05 

RF 0.90 1.50 0.5910 5.15E-03 9.20E-07 2.00E-02 6.80E-04 10.00 0.023 0.000 0.000 0.000 2.61E-06 

Wk 0.90 1.92E-02 21.3971 5.15E-03 9.20E-07 2.00E-02 6.80E-04 10.00 0.023 0.000 0.000 0.000 1.86E-07 
Table D.C.1 Application of Markov Model to BWR-1 Cat 1 LOCA Frequency 

 CASE INSPECTION INDEPENDENT PARAMETERS

INSPECTION DEPENDENT 

PARAMETERS 

Leak 

Inspection 

Interval PLI TSUBLI MU PHI RHO_F RHO_L Lambda TFI TSUBR PFI PFD OMEGA Haz{T} 

Shift 0.90 9.13E-04 37.9048 5.15E-03 9.20E-07 2.00E-02 6.80E-04 10.00 0.023 0.000 0.000 0.000 1.53E-07 

None 0.00 1.50 0.0000 5.15E-03 9.20E-07 2.00E-02 6.80E-04 10.00 0.023 1.000 0.500 0.050 1.22E-05 

RF 0.90 1.50 0.5910 5.15E-03 9.20E-07 2.00E-02 6.80E-04 10.00 0.023 1.000 0.500 0.050 1.56E-06 

Wk 0.90 1.92E-02 21.3971 5.15E-03 9.20E-07 2.00E-02 6.80E-04 10.00 0.023 1.000 0.500 0.050 1.08E-07 

Shift 0.90 9.13E-04 37.9048 5.15E-03 9.20E-07 2.00E-02 6.80E-04 10.00 0.023 1.000 0.500 0.050 8.88E-08 

None 0.00 1.50 0.0000 5.15E-03 9.20E-07 2.00E-02 6.80E-04 10.00 0.023 1.000 0.900 0.090 9.46E-06 

RF 0.90 1.50 0.5910 5.15E-03 9.20E-07 2.00E-02 6.80E-04 10.00 0.023 1.000 0.900 0.090 1.11E-06 

Wk 0.90 1.92E-02 21.3971 5.15E-03 9.20E-07 2.00E-02 6.80E-04 10.00 0.023 1.000 0.900 0.090 7.61E-08 

Shift 0.90 9.13E-04 37.9048 5.15E-03 9.20E-07 2.00E-02 6.80E-04 10.00 0.023 1.000 0.900 0.090 6.26E-08 

None 0.00 1.50 0.0000 1.24E-02 2.22E-06 2.00E-02 1.64E-03 10.00 0.023 0.000 0.000 0.000 9.72E-05 

RF 0.90 1.50 0.5910 1.24E-02 2.22E-06 2.00E-02 1.64E-03 10.00 0.023 0.000 0.000 0.000 1.39E-05 

Wk 0.90 1.92E-02 21.3971 1.24E-02 2.22E-06 2.00E-02 1.64E-03 10.00 0.023 0.000 0.000 0.000 9.84E-07 

Shift 0.90 9.13E-04 37.9048 1.24E-02 2.22E-06 2.00E-02 1.64E-03 10.00 0.023 0.000 0.000 0.000 8.10E-07 

None 0.00 1.50 0.0000 1.24E-02 2.22E-06 2.00E-02 1.64E-03 10.00 0.023 1.000 0.500 0.050 6.72E-05 

RF 0.90 1.50 0.5910 1.24E-02 2.22E-06 2.00E-02 1.64E-03 10.00 0.023 1.000 0.500 0.050 8.44E-06 

Wk 0.90 1.92E-02 21.3971 1.24E-02 2.22E-06 2.00E-02 1.64E-03 10.00 0.023 1.000 0.500 0.050 5.83E-07 

Shift 0.90 9.13E-04 37.9048 1.24E-02 2.22E-06 2.00E-02 1.64E-03 10.00 0.023 1.000 0.500 0.050 4.79E-07 

None 0.00 1.50 0.0000 1.24E-02 2.22E-06 2.00E-02 1.64E-03 10.00 0.023 1.000 0.900 0.090 5.27E-05 

RF 0.90 1.50 0.5910 1.24E-02 2.22E-06 2.00E-02 1.64E-03 10.00 0.023 1.000 0.900 0.090 6.11E-06 

Wk 0.90 1.92E-02 21.3971 1.24E-02 2.22E-06 2.00E-02 1.64E-03 10.00 0.023 1.000 0.900 0.090 4.16E-07 

Shift 0.90 9.13E-04 37.9048 1.24E-02 2.22E-06 2.00E-02 1.64E-03 10.00 0.023 1.000 0.900 0.090 3.42E-07 

None 0.00 1.50 0.0000 5.00E-03 8.92E-07 2.00E-02 6.59E-04 10.00 0.023 0.000 0.000 0.000 1.68E-05 

RF 0.90 1.50 0.5910 5.00E-03 8.92E-07 2.00E-02 6.59E-04 10.00 0.023 0.000 0.000 0.000 2.46E-06 

Wk 0.90 1.92E-02 21.3971 5.00E-03 8.92E-07 2.00E-02 6.59E-04 10.00 0.023 0.000 0.000 0.000 1.75E-07 

Shift 0.90 9.13E-04 37.9048 5.00E-03 8.92E-07 2.00E-02 6.59E-04 10.00 0.023 0.000 0.000 0.000 1.44E-07 

None 0.00 1.50 0.0000 5.00E-03 8.92E-07 2.00E-02 6.59E-04 10.00 0.023 1.000 0.500 0.050 1.15E-05 

RF 0.90 1.50 0.5910 5.00E-03 8.92E-07 2.00E-02 6.59E-04 10.00 0.023 1.000 0.500 0.050 1.47E-06 

Wk 0.90 1.92E-02 21.3971 5.00E-03 8.92E-07 2.00E-02 6.59E-04 10.00 0.023 1.000 0.500 0.050 1.02E-07 

Shift 0.90 9.13E-04 37.9048 5.00E-03 8.92E-07 2.00E-02 6.59E-04 10.00 0.023 1.000 0.500 0.050 8.37E-08 

None 0.00 1.50 0.0000 5.00E-03 8.92E-07 2.00E-02 6.59E-04 10.00 0.023 1.000 0.900 0.090 8.91E-06 

RF 0.90 1.50 0.5910 5.00E-03 8.92E-07 2.00E-02 6.59E-04 10.00 0.023 1.000 0.900 0.090 1.05E-06 

Wk 0.90 1.92E-02 21.3971 5.00E-03 8.92E-07 2.00E-02 6.59E-04 10.00 0.023 1.000 0.900 0.090 7.16E-08 

Shift 0.90 9.13E-04 37.9048 5.00E-03 8.92E-07 2.00E-02 6.59E-04 10.00 0.023 1.000 0.900 0.090 5.89E-08 

None 0.00 1.50 0.0000 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 0.000 0.000 0.000 8.70E-07 

RF 0.90 1.50 0.5910 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 0.000 0.000 0.000 1.29E-07 

Wk 0.90 1.92E-02 21.3971 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 0.000 0.000 0.000 9.23E-09 

Shift 0.90 9.13E-04 37.9048 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 0.000 0.000 0.000 7.59E-09 

None 0.00 1.50 0.0000 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.500 0.050 5.89E-07 
Table D.C.1 Application of Markov Model to BWR-1 Cat 1 LOCA Frequency 

 CASE INSPECTION INDEPENDENT PARAMETERS

INSPECTION DEPENDENT 

PARAMETERS 

Leak 

Inspection 

Interval PLI TSUBLI MU PHI RHO_F RHO_L Lambda TFI TSUBR PFI PFD OMEGA Haz{T} 

RF 0.90 1.50 0.5910 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.500 0.050 7.62E-08 

Wk 0.90 1.92E-02 21.3971 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.500 0.050 5.29E-09 

Shift 0.90 9.13E-04 37.9048 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.500 0.050 4.35E-09 

None 0.00 1.50 0.0000 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.900 0.090 4.55E-07 

RF 0.90 1.50 0.5910 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.900 0.090 5.40E-08 

Wk 0.90 1.92E-02 21.3971 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.900 0.090 3.70E-09 

Shift 0.90 9.13E-04 37.9048 1.11E-03 1.99E-07 2.00E-02 1.47E-04 10.00 0.023 1.000 0.900 0.090 3.04E-09 
D-101

ATTACHMENT D TO APPENDIX D 

SIGNIFICANT FAILURES OF SAFETY RELATED PIPING 

Table D.D.1 is a list of selected significant pipe failures during the period 1970 – 2003. The list includes 

failures of Code Class 1 and 2 piping systems inside the containment/drywell and auxiliary/reactor building 

structures of commercial nuclear power plants. The technical information has been extracted from the OPDE 

database (Attachment A). 

Table D.D.1 Selected Historical Pipe Failure Information 

Event 

Date

Plant Country Estimated Peak

Leak/Flow 

Rate [gpm]

Description

12/14/02 Brunsbüttel 

(BWR) 

Germany -- 

(see Description) 

Rupture of reactor head cooling pipe (the 

rupture occurred in section of pipe that was 

separated from the RPV through an 

isolation valve – no RPV steam leakage 

observed 

11/7/02 Hamaoka-1 

(BWR) 

Japan >> 50 Rupture of pipe in High Pressure Coolant 

Injection system; the rupture occurred 

during a functional system test 

7/12/99 Tsuruga-2 

(PWR) 

Japan 16 Thermal fatigue induced fracture of elbow 

connected to regenerative heat exchanger 

5/12/98 Civaux-1 

(PWR) 

France 131 Thermal fatigue induced fracture of seam

welded elbow in the Residual Heat 

Removal System 

5/27/97 Calvert Cliffs-1 

(PWR) 

USA 8.0 Fractured pressurizer instrument sensing 

line; attributed to vibration fatigue 

4/21/97 Oconee-2 

(PWR) 

USA 12.0 Thermal fatigue induced fracture of weld 

connecting HPI/NMU pipe to RCS (see 

Base Case PWR-3) 

12/21/96 Dampierre-1 

(PWR) 

France 0.6 Thermal-fatigue induced weld crack in 

straight section of Safety Injection line to 

RCS hot leg. 

3/8/95 Borssele 

(PWR) 

The 

Netherlands 

65.8 While in hot standby prior to startup a weld 

fractured on the High Head Safety Injection 

common discharge header; attributed to 

vibration-fatigue 

2/23/95 Biblis-B 

(PWR) 

Germany 15.8 Thermal fatigue induced fracture of base 

metal of pipe in Chemical and Volume 

Control system 

3/3/94 Kola-2 

(PWR) 

Russia S-LOCA Soviet-designed PWR of type WWER440/230; full circumferential fracture of 

NPS2 makeup pipe while shutting down 

for maintenance outage. Event resulted in 

High Pressure Safety Injection system 

actuation; a beyond-design basis accident. 

9/20/92 Dampierre-2 

(PWR) 

France 3.2 Non-isolable, thermal fatigue induced weld 

fracture in Safety Injection System. 

6/18/88 Tihange-1 Belgium 6.3 Thermal fatigue induced fracture of base 
D-102

Table D.D.1 Selected Historical Pipe Failure Information 

Event 

Date

Plant Country Estimated Peak

Leak/Flow 

Rate [gpm]

Description

(PWR) metal of Safety Injection line to RCS hot 

leg 

12/9/87 Farley-2 

(PWR) 

USA 0.7 Thermal fatigue induced weld fracture in 

Safety Injection line to RCS cold leg 

8/16/87 McGuire-1 

(PWR) 

USA 39.5 Fracture (80% of circumference) of 1-inch 

socket weld in drain line off of letdown line 

inside containment. The weld fracture 

occurred during startup operations (8% 

reactor power) 

5/31/86 Obrigheim 

(PWR) 

Germany 0.32 Thermal fatigue induced weld fracture in 

makeup line to RCS. 

7/29/85 Sequoyah-2 

(PWR) 

USA 60.0 Fractured sample line in Chemical and 

Volume Control system; attributed to 

vibration-fatigue 

8/6/84 McGuire-2 

(PWR) 

USA 8.0 Water hammer induced fracture of socket 

weld in letdown line 

1/25/83 Maine Yankee 

(PWR) 

USA 100 Fractured main feedwater pipe adjacent to 

weld joining pipe and steam generator safe 

end; attributed to severe water hammer. 

1/21/82 Crystal River-3 

(PWR) 

USA 1 140-degree circumferential crack in 

makeup line near valve-to-safe end weld; 

attributed to thermal fatigue 

2/12/80 Santa Maria de 

Garona (BWR) 

Spain 0.8 IGSCC induced through-wall flaw in 

Reactor Recirculation nozzle-to-safe end 

weld 

8/29/80 TVO-1 

(BWR) 

Finland 315 Thermal fatigue induced fracture of tee in 

Reactor Water Cleanup system. The 

fracture occurred during the 

commissioning of this reactor unit. 

6/14/78 Duane Arnold 

(BWR) 

USA 3.0 IGSCC induced through-wall flaw in 

Reactor Recirculation nozzle-to-safe end 

weld 

11/13/73 Indian Point-2 

(PWR) 

USA 15.8 180-degree circumferential crack of 18-

inch feedwater line weld inside 

containment 

4/28/70 H.B. Robinson-2 

(PWR) 

USA >> 5027 360-degree break in 6-inch branch line 

between No. 3 steam generator main steam 

line and safety valve. The failure occurred 

during the final stages of hot functional 

testing 

 

27 At the time of the pipe break the primary system was at 278 C (533 F) and 15.3 MPa (2,225 psi) primary system pressure with a secondary system 

pressure of 6.2 MPa (900 psi) 
APPENDIX E 

PIPING BASE CASE RESULTS OF 

WILLIAM GALYEAN 

E-1 

APPENDIX E 

PIPING BASE CASE RESULTS OF WILLIAM GALYEAN

E.1 Summary

In this base case study, LOCA frequencies are calculated using a “top-down” approach. Specifically, a

total LOCA frequency is calculated using U.S. commercial nuclear power plant (NPP) operating 

experience. This total frequency is then allocated to the LOCA size categories, RCS subsystems and 

components, and degradation mechanisms. This allocation is performed using data on primary system 

leaks and cracks from both U.S. and foreign PWR and BWR reactors. 

E.2 Assumptions and Observations

As with all analyses, there are a number of implicit assumptions associated with this approach. First is 

that past performance is representative of future performance. The common scenario for the occurrence

of a LOCA starts with postulating the existence of a flaw or defect in the primary reactor coolant 

boundary. This flaw is then subjected to a stress that results in the catastrophic failure of the primary 

pressure boundary, producing a LOCA. The U.S. LWR operating experience to date consists of 

approximately 100 reactors with an average age of about 23 years. During this time the RCS of these 

plants have experience numerous transients and loads, which have produce a wide range of stresses. 

Whether these plants operate for 40 years (or 60 years with license extensions) this available operating 

experience represents a significant portion of the average plants lifetime. It is therefore reasonable to 

assume that the stresses that have already occurred are representative of those that will occur in the future. 

Similarly, various degradation mechanisms have affected RCS pipe, welds and components. However, 

when these degradation mechanisms have been detected, mitigation programs have subsequently been 

implemented (e.g., IGSCC in BWRs). Therefore, the number of flaws and defects in the RCS is likely to

be cyclic over time. As the degradation mechanism manifests itself, the number of defects grows, as the 

degradation mechanism is addressed and mitigated, the number of defects is reduced. Again, the 

assumption here is that current 23 years of operating (on average, per reactor) are representative of the 

remaining operating life. 

Another observation is the occurrence of zero LOCAs for both PWRs and BWRs. Although this does not 

prove that the LOCA frequencies are the same for both designs, it likewise does not support different 

LOCA frequencies. Therefore, for this analysis, the operating experience data (i.e., zero failures) will be 

pooled to generate a single LOCA frequency. 

Furthermore, this analysis, just as every LOCA frequency estimate performed to date, assumes that the 

frequency of a LOCA decreases as pipe size increases. This might be attributable to a couple of issues. 

First, for small diameter pipe, some failure mechanisms exist that don’t apply to larger diameter pipe

(e.g., compression fitting failures and socket welds). Second, the same flaw in both a small diameter pipe 

and a large diameter pipe represents a large percentage of the pipe diameter in the small diameter pipe. 

Third, inspection is probably more thorough in larger diameter pipe so that the chance of a defect going 

undetected is less in the larger diameter pipe. For all of these reasons (and probably others), the total 

LOCA frequency is reduced as LOCA size category increases. The scaling factor of 1⁄2 order of 

magnitude (assuming a lognormal probability distribution on LOCA frequency) appears to be reasonably 

consistent with historical LOCA frequency estimates. 
E-2 

This assumption of a half-order of magnitude (i.e., approximately a factor of 3) decrease in frequency for 

each increase in LOCA size is an assumption based on the general practice employed in estimating LOCA 

frequencies over the past 30 years starting with the Reactor Safety Study (Ref. E.1). This assumption is 

further supported by work done by Beliczey and Schulz (Ref. E.2). In this study, a combination of 

operating experience and fracture mechanics is used to demonstrate that the conditional probability of a 

rupture, given a leak, decreases as pipe diameter increases. This conclusion is reached because the size of 

detectable cracks and leaks remains relatively constant as a function of pipe size. Therefore, the relative 

crack or leak size as a function of the pipe circumference decreases, and the safety margin increases, as 

the pipe diameter increases. 

Additionally, Beliczey and Schulz developed a quantitative conditional failure probability --- which 

decreases by approximately 1⁄2 order of magnitude for each successively larger LOCA size --- that was 

based on the propensity of through-wall fatigue flaws to lead to successively larger LOCA sizes. 

Although the quantitative conditional failure probability is not applicable to all failure mechanisms and 

systems, this simple relationship has been extensively employed. This assumption was also employed in

these analyses. 

The final premise of this base case analysis is that the relative frequency of precursor data (i.e., leaks and 

cracks) is an indicator of the relative frequency of LOCA events. In the calculations that follow, the total 

LOCA frequency is allocated to the different RCS subsystems and components, and the different 

degradation mechanisms according to the relative frequency of observed leaks and cracks attributable to 

these subsystems and mechanism. Note that in order to determine the relative frequencies, complete 

crack and leak data are not needed, only consistent data that has not been biased by the over reporting of 

one attribute relative to another. Completeness in the data is neither required nor important, only 

consistency. 

E.3 Total LOCA Frequency Estimates

The total LOCA frequency is calculated using U.S. NPP experience of zero Category-1 LOCAs (i.e., 

greater than 380 lpm [100 gpm]) in 2,647 LWR-years of operation (as of 4/24/2003). A Bayesian update 

of a non-informative prior-distribution was performed to produce a total LOCA frequency of 1.9E-4 per 

LWR-year. 

Table E.1 Total LOCA Frequency (per LWR-Year) Including Uncertainty, Using a NonInformative Prior and U.S. LWR Operating Experience

5% 50% mean 95%

7.4E-07 8.6E-05 1.9E-04 7.3E-04

E.4 LOCA Frequency Allocation by RCS Pipe and Non-Pipe

The total LOCA frequency calculated above is first allocated between pipe and non-pipe passive 

components using data on primary system leaks and cracks collected from licensee event reports (LERs). 

These data records were collected, reviewed and categorized specifically for this effort. Since these data 

will only be used to ascribe a relative frequency between pipe and passive non-pipe components, 

complete data are not necessary, only data that have been reported consistently. These data and the 

resultant allocation are summarized in the table below. Steam generator tube ruptures are being assessed 

separately, and are therefore removed from this allocation. 
E-3 

Table E.2 Allocation of LOCA Frequency Between Pipe and Passive Non-Pipe Components 

Reactor Coolant Pressure Boundary 

failure (cracks or leaks) events 

1990-2002 LERs 

LWR PWR BWR 

Total number of failure events 448 388 60 

Number of SG tube failure events 112 112 0 

Total minus SG events 336 276 60 

Number of pipe failure events 54 24 30 

Exclude SG tube events since these can be estimated directly 

Therefore 

Number of non-pipe failure events 282 252 30

fraction of LOCA frequency attributed to 

pipes 0.16 0.09 0.50

non-pipes 0.84 0.91 0.50

total LOCA frequency = 1.9E-04

LOCA frequency attributable to 

pipes 3.0E-05 1.6E-05 9.4E-05

non-pipes 1.6E-04 1.7E-04 9.4E-05

E.5 LOCA Frequency by Size Category

The total LOCA frequencies calculated above are for Category-1 LOCAs. The simple approach taken 

here is that the LOCA frequency is reduced by 1⁄2 order of magnitude (assuming a lognormal distribution), 

for each step up in size category. There are a number of reasons for this approach. Between the smallest 

pipe size categories (i.e., < 2 inches, and > 2 inches) there is a significant difference in the failure 

mechanisms. For the smallest pipes, the operating experience includes failures of compression fittings 

and socket welds, which are not used in larger size pipe. Also, a number of studies on crack and leak

events indicate a decrease in these precursor frequencies, as pipe diameter increases. Lastly, virtually 

every estimate of LOCA frequencies ever made has resulted in a reduced frequency for the large LOCA 

sizes. 

E.6 LOCA Frequency by Degradation Mechanism and Subsystem

In addition to the LER data used to allocate the LOCA frequency between pipe and passive non-pipe 

components, the LOCA frequency was further allocated among the different degradation mechanisms 

observed and among the different RCS subsystems and components defined for this project. These 

allocations were based on data collected from both U.S. reactor operating experience (primarily LERs),

and from foreign LWR operating experience (SLAP database). One complication to this approach is the 

IGSCC-related experience in U.S. BWR plants. IGSCC was an issue for BWRs in the early 1980’s. 

Many U.S. BWRs implemented IGSCC mitigation programs in the mid-1980’s, which have greatly 

reduced the occurrence of IGSCC. To avoid unrealistically over weighting the IGSCC mechanism, the 
E-4 

BWR experience was segregated and only the post 1985 experience was used for allocating the relative 

contribution to LOCA frequency by degradation mechanism. 

Lastly, although the guidance for calculating base-case frequencies for this project included estimates for 

25 years, 40 years and 60 years, this particular base-case calculation assumed that the frequencies were 

generally independent of plant life. This is based on the IGSCC experience that demonstrated that 

although degradation mechanisms are at work that can result in an increase in the LOCA frequency over 

time, so to are mitigation programs and general performance improvement programs (e.g., more effective

inspections), that can result in a decrease in the LOCA frequency. Therefore, overall these competing

effects are assumed to cancel each other out for a net zero effect on LOCA frequency. That is, the current 

LOCA frequency (approximately 25 year life) is assumed to be application for 40 and 60 years as well. 

E.7 LOCA Frequency Tables

The following tables display the detailed results of Base Case #1 calculations. The legend of degradation 

categories is shown in Table E.3. LOCA data are presented on the tables listed below: 

 PWR pipe Table E.4 

 PWR passive non-pipe Table E.5 

 BWR pipe Table E.6 

 BWR passive non-pipe Table E.7 

Table E.3 Degradation Categories 

Deg Mech DM Description 

MA Material Aging 

FDR Fabrication Defect and Repair 

SCC Stress Corrosion Cracking 

LC Local Corrosion 

MF Mechanical Fatigue 

TF Thermal Fatigue 

FS Flow Sensitive (includes FAC and E/C) 

UNK Unknown 

E.8 References

E.1 “Reactor Safety Study: An Assessment of Accident Risks in U.S. Commercial Nuclear Power 

Plants,” WASH-1400, U.S. Nuclear Regulatory Commission, October 1975. 

E.2 Beliczey, S., and Schulz, H., “Comments on Probabilities of Leaks and Breaks of Safety-Related 

Piping in PWR Plants,” International Journal of Pressure Vessel and Piping, Vol. 43, pp. 219 – 

227, 1990. 
PWR LOCA Frequency (for Pipes) Allocated by System, Degradation Mechanism, and Size Category 

em (for PWRs) Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

e) Est. # of greater 

than

100 1500 5000 25000 100000 500000 gpm 

Deg. Mech. LOCA Min Max Welds Total 0.4 1.7 3.0 6.8 14.0937 31.5146 dia. (in.) 

t Fractional Fraction (mm) (mm) Dom. LOCA Freq 1 0.3 0.1 0.03 0.01 0.003 probabilty

s Contribution by System (in) (in) Total 1.6E-05 

6 0.01 70 100 1.4E-07 1.4E-07 4.1E-08 1.4E-08 0.0 0.0 0.0 

4 1 2.8 3.9 1 1.4E-07 1.4E-07 4.1E-08 1.4E-08 -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

4 1.00 1.4E-07 1.4E-07 4.1E-08 1.4E-08 -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

0 0.01 60 100 1.5E-07 1.5E-07 4.5E-08 1.5E-08 0.0 0.0 0.0 

5 1 2.4 3.9 1 1.5E-07 1.5E-07 4.5E-08 1.5E-08 -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

1 0.20 3.0E-08 3.0E-08 9.0E-09 3.0E-09 -- -- -- 2 0.40 6.0E-08 6.0E-08 1.8E-08 6.0E-09 -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

2 0.40 6.0E-08 6.0E-08 1.8E-08 6.0E-09 -- -- -- 

0 0.26 13 250 4.2E-06 4.2E-06 1.3E-06 4.2E-07 1.3E-07 0.0 0.0 

0 1 0.5 9.8 1 4.2E-06 4.2E-06 1.3E-06 4.2E-07 1.3E-07 -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 

2 0.09 3.6E-07 3.6E-07 1.1E-07 3.6E-08 1.1E-08 -- -- 9 0.14 5.7E-07 5.7E-07 1.7E-07 5.7E-08 1.7E-08 -- -- 4 0.67 2.8E-06 2.8E-06 8.5E-07 2.8E-07 8.5E-08 -- -- 

7 0.05 2.1E-07 2.1E-07 6.3E-08 2.1E-08 6.3E-09 -- -- 6 0.04 1.8E-07 1.8E-07 5.4E-08 1.8E-08 5.4E-09 -- -- 
em (for PWRs) Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

e) Est. # of greater 

than

100 1500 5000 25000 100000 500000 gpm 

Deg. Mech. LOCA Min Max Welds Total 0.4 1.7 3.0 6.8 14.0937 31.5146 dia. (in.) 

t Fractional Fraction (mm) (mm) Dom. LOCA Freq 1 0.3 0.1 0.03 0.01 0.003 probabilty

s Contribution by System (in) (in) Total 1.6E-05 2 0.01 6.0E-08 6.0E-08 1.8E-08 6.0E-09 1.8E-09 -- -- 

5 0.11 10 80 1.8E-06 1.8E-06 5.3E-07 1.8E-07 0.0 0.0 0.0 

6 1 0.4 3.1 1 1.8E-06 1.8E-06 5.3E-07 1.8E-07 -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

9 0.20 3.4E-07 3.4E-07 1.0E-07 3.4E-08 -- -- -- 5 0.11 1.9E-07 1.9E-07 5.7E-08 1.9E-08 -- -- -- 

7 0.59 1.0E-06 1.0E-06 3.1E-07 1.0E-07 -- -- -- 

2 0.04 7.7E-08 7.7E-08 2.3E-08 7.7E-09 -- -- -- 2 0.04 7.7E-08 7.7E-08 2.3E-08 7.7E-09 -- -- -- 1 0.02 3.8E-08 3.8E-08 1.1E-08 3.8E-09 -- -- -- 

6 0.03 10 25 5.0E-07 5.0E-07 0.0 0.0 0.0 0.0 0.0 

3 1 0.4 1.0 1 5.0E-07 5.0E-07 -- -- -- -- -- 

0.00 0.0E+00 0.0E+00 -- -- -- -- -- 0.00 0.0E+00 0.0E+00 -- -- -- -- -- 

4 0.31 1.5E-07 1.5E-07 -- -- -- -- -- 2 0.15 7.7E-08 7.7E-08 -- -- -- -- -- 2 0.15 7.7E-08 7.7E-08 -- -- -- -- -- 2 0.15 7.7E-08 7.7E-08 -- -- -- -- -- 

0.00 0.0E+00 0.0E+00 -- -- -- -- -- 0.00 0.0E+00 0.0E+00 -- -- -- -- -- 

3 0.23 1.2E-07 1.2E-07 -- -- -- -- -- 

8 0.28 9 200 4.6E-06 4.6E-06 1.4E-06 4.6E-07 1.4E-07 0.0 0.0 

9 1 0.4 7.9 1 4.6E-06 4.6E-06 1.4E-06 4.6E-07 1.4E-07 -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 

1 0.01 3.8E-08 3.8E-08 1.2E-08 3.8E-09 1.2E-09 -- -- 1 0.01 3.8E-08 3.8E-08 1.2E-08 3.8E-09 1.2E-09 -- -- 

1 0.09 4.2E-07 4.2E-07 1.3E-07 4.2E-08 1.3E-08 -- -- 9 0.16 7.3E-07 7.3E-07 2.2E-07 7.3E-08 2.2E-08 -- -- 4 0.62 2.8E-06 2.8E-06 8.5E-07 2.8E-07 8.5E-08 -- -- 

2 0.02 7.7E-08 7.7E-08 2.3E-08 7.7E-09 2.3E-09 -- -- 1 0.01 3.8E-08 3.8E-08 1.2E-08 3.8E-09 1.2E-09 -- -- 

0 0.08 3.8E-07 3.8E-07 1.2E-07 3.8E-08 1.2E-08 -- -- 
em (for PWRs) Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

e) Est. # of greater 

than

100 1500 5000 25000 100000 500000 gpm 

Deg. Mech. LOCA Min Max Welds Total 0.4 1.7 3.0 6.8 14.0937 31.5146 dia. (in.) 

t Fractional Fraction (mm) (mm) Dom. LOCA Freq 1 0.3 0.1 0.03 0.01 0.003 probabilty

s Contribution by System (in) (in) Total 1.6E-05 

4 0.01 25 25 1.9E-07 1.9E-07 5.8E-08 0.0 0.0 0.0 0.0 

5 1 1.0 1.0 1 1.9E-07 1.9E-07 5.8E-08 -- -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 

2 0.40 7.7E-08 7.7E-08 2.3E-08 -- -- -- -- 2 0.40 7.7E-08 7.7E-08 2.3E-08 -- -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 

1 0.20 3.8E-08 3.8E-08 1.2E-08 -- -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 

4 0.01 20 75 1.9E-07 1.9E-07 5.8E-08 1.9E-08 0.0 0.0 0.0 

5 1 0.8 3.0 1 1.9E-07 1.9E-07 5.8E-08 1.9E-08 -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

1 0.20 3.8E-08 3.8E-08 1.2E-08 3.8E-09 -- -- -- 3 0.60 1.2E-07 1.2E-07 3.5E-08 1.2E-08 -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

1 0.20 3.8E-08 3.8E-08 1.2E-08 3.8E-09 -- -- -- 

3 0.002 3.8E-08 3.8E-08 1.2E-08 3.8E-09 1.2E-09 3.8E-10 1.2E-101 1 1 3.8E-08 3.8E-08 1.2E-08 3.8E-09 1.2E-09 3.8E-10 1.2E-10

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00

1 1.00 3.8E-08 3.8E-08 1.2E-08 3.8E-09 1.2E-09 3.8E-10 1.2E-10

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00

1 0.009 25 650 1.5E-07 1.5E-07 4.6E-08 1.5E-08 4.6E-09 1.5E-09 0.0E+00 5 1 1.0 25.6 1 1.5E-07 1.5E-07 4.6E-08 1.5E-08 4.6E-09 1.5E-09 -- 
em (for PWRs) Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

e) Est. # of greater 

than

100 1500 5000 25000 100000 500000 gpm 

Deg. Mech. LOCA Min Max Welds Total 0.4 1.7 3.0 6.8 14.0937 31.5146 dia. (in.) 

t Fractional Fraction (mm) (mm) Dom. LOCA Freq 1 0.3 0.1 0.03 0.01 0.003 probabilty

s Contribution by System (in) (in) Total 1.6E-05 

0.00 40.0 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 

2 0.44 6.8E-08 6.8E-08 2.1E-08 6.8E-09 2.1E-09 6.8E-10 -- 5 0.11 1.7E-08 1.7E-08 5.1E-09 1.7E-09 5.1E-10 1.7E-10 -- 1 0.22 3.4E-08 3.4E-08 1.0E-08 3.4E-09 1.0E-09 3.4E-10 -- 1 0.22 3.4E-08 3.4E-08 1.0E-08 3.4E-09 1.0E-09 3.4E-10 -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 

3 0.002 3.8E-08 3.8E-08 1.2E-08 3.8E-09 1.2E-09 3.8E-10 1.2E-10

1 1 1 3.8E-08 3.8E-08 1.2E-08 3.8E-09 1.2E-09 3.8E-10 1.2E-10

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+000.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00

1 1.00 3.8E-08 3.8E-08 1.2E-08 3.8E-09 1.2E-09 3.8E-10 1.2E-100 0.09 13 500 1.5E-06 1.5E-06 4.4E-07 1.5E-07 4.4E-08 1.5E-08 0.0 

9 1 0.5 19.7 1 1.5E-06 1.5E-06 4.4E-07 1.5E-07 4.4E-08 1.5E-08 -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 

9 0.18 2.7E-07 2.7E-07 8.1E-08 2.7E-08 8.1E-09 2.7E-09 -- 9 0.18 2.7E-07 2.7E-07 8.1E-08 2.7E-08 8.1E-09 2.7E-09 -- 

3 0.47 6.9E-07 6.9E-07 2.1E-07 6.9E-08 2.1E-08 6.9E-09 -- 

1 0.02 3.0E-08 3.0E-08 9.0E-09 3.0E-09 9.0E-10 3.0E-10 -- 4 0.08 1.2E-07 1.2E-07 3.6E-08 1.2E-08 3.6E-09 1.2E-09 -- 3 0.06 9.0E-08 9.0E-08 2.7E-08 9.0E-09 2.7E-09 9.0E-10 -- 

3 0.03 15 100 4.3E-07 4.3E-07 1.3E-07 4.3E-08 0.0 0.0 0.0 

4 1 0.6 3.9 1 4.3E-07 4.3E-07 1.3E-07 4.3E-08 -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 
em (for PWRs) Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

e) Est. # of greater 

than

100 1500 5000 25000 100000 500000 gpm 

Deg. Mech. LOCA Min Max Welds Total 0.4 1.7 3.0 6.8 14.0937 31.5146 dia. (in.) 

t Fractional Fraction (mm) (mm) Dom. LOCA Freq 1 0.3 0.1 0.03 0.01 0.003 probabilty

s Contribution by System (in) (in) Total 1.6E-05 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

2 0.14 6.1E-08 6.1E-08 1.8E-08 6.1E-09 -- -- -- 5 0.36 1.5E-07 1.5E-07 4.6E-08 1.5E-08 -- -- -- 6 0.43 1.8E-07 1.8E-07 5.5E-08 1.8E-08 -- -- -- 1 0.07 3.1E-08 3.1E-08 9.2E-09 3.1E-09 -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

0 0.14 9 600 2.3E-06 2.3E-06 7.0E-07 2.3E-07 7.0E-08 2.3E-08 0.0 8 1 0.4 23.6 1 2.3E-06 2.3E-06 7.0E-07 2.3E-07 7.0E-08 2.3E-08 -- 

0.00 114 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 

5 0.06 1.5E-07 1.5E-07 4.5E-08 1.5E-08 4.5E-09 1.5E-09 -- 

8 0.36 8.4E-07 8.4E-07 2.5E-07 8.4E-08 2.5E-08 8.4E-09 -- 6 0.33 7.8E-07 7.8E-07 2.3E-07 7.8E-08 2.3E-08 7.8E-09 -- 

9 0.12 2.7E-07 2.7E-07 8.1E-08 2.7E-08 8.1E-09 2.7E-09 -- 4 0.05 1.2E-07 1.2E-07 3.6E-08 1.2E-08 3.6E-09 1.2E-09 -- 6 0.08 1.8E-07 1.8E-07 5.4E-08 1.8E-08 5.4E-09 1.8E-09 -- 

4 0.01 50 75 1.9E-07 1.9E-07 5.8E-08 1.9E-08 0.0 0.0 0.0 6 1 2.0 3.0 1 1.9E-07 1.9E-07 5.8E-08 1.9E-08 -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

1 0.17 3.2E-08 3.2E-08 9.6E-09 3.2E-09 -- -- -- 3 0.50 9.6E-08 9.6E-08 2.9E-08 9.6E-09 -- -- -- 1 0.17 3.2E-08 3.2E-08 9.6E-09 3.2E-09 -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

1 0.17 3.2E-08 3.2E-08 9.6E-09 3.2E-09 -- -- -- 

5 0.001 1.5E-08 1.5E-08 4.5E-09 1.5E-09 4.5E-10 0.0E+00 0.0E+00

5 1 10 1 1.5E-08 1.5E-08 4.5E-09 1.5E-09 4.5E-10 -- -- 0 0.00 13 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 
em (for PWRs) Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

e) Est. # of greater 

than

100 1500 5000 25000 100000 500000 gpm 

Deg. Mech. LOCA Min Max Welds Total 0.4 1.7 3.0 6.8 14.0937 31.5146 dia. (in.) 

t Fractional Fraction (mm) (mm) Dom. LOCA Freq 1 0.3 0.1 0.03 0.01 0.003 probabilty

s Contribution by System (in) (in) Total 1.6E-05 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 

3 0.50 7.5E-09 7.5E-09 2.3E-09 7.5E-10 2.3E-10 -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 

3 0.50 7.5E-09 7.5E-09 2.3E-09 7.5E-10 2.3E-10 -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 

5 1 1.64E-05 1.64E-05 4.78E-06 1.57E-06 3.86E-07 4.06E-08 2.31E-10
Table E.5 PWR Passive, Non-Pipe Component LOCA Frequency

ontributors 

unt = 252 PWR 

52 Non-Pipe Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

LOCA 100 1500 5000 25000 100000 500000 gpm 

ation Mechanism Non-Pipe Freq 0.5 1.8 3.3 7.3 18.4 41.2 dia. (in.) 

Fract Contr LOCA Fract 1.7E-04 1 0.3 0.1 0.03 0.01 0.003 probabilty 

3 1.0 0.01 2.05E-06 2.05E-06 6.16E-07 2.05E-07 6.16E-08 2.05E-08 6.16E-09

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

2 0.67 1.37E-06 1.37E-06 4.10E-07 1.37E-07 4.10E-08 1.37E-08 4.10E-09

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

1 0.33 6.84E-07 6.84E-07 2.05E-07 6.84E-08 2.05E-08 6.84E-09 2.05E-09

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

1 1.0 0.00 6.84E-07 6.84E-07 2.05E-07 6.84E-08 2.05E-08 6.84E-09 2.05E-09

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

1 1.00 2.05E-06 2.05E-06 6.16E-07 2.05E-07 6.16E-08 2.05E-08 6.16E-09

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

65 1.0 0.26 4.45E-05 4.45E-05 1.33E-05 4.45E-06 1.33E-06 4.45E-07 1.33E-07

4 0.06 1.26E-07 1.26E-07 3.79E-08 1.26E-08 3.79E-09 1.26E-09 3.79E-10

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

1 0.02 3.16E-08 3.16E-08 9.47E-09 3.16E-09 9.47E-10 3.16E-10 9.47E-111 0.02 3.16E-08 3.16E-08 9.47E-09 3.16E-09 9.47E-10 3.16E-10 9.47E-11

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

59 0.91 1.86E-06 1.86E-06 5.59E-07 1.86E-07 5.59E-08 1.86E-08 5.59E-09

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00
ontributors 

unt = 252 PWR 

52 Non-Pipe Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

LOCA 100 1500 5000 25000 100000 500000 gpm 

ation Mechanism Non-Pipe Freq 0.5 1.8 3.3 7.3 18.4 41.2 dia. (in.) 

Fract Contr LOCA Fract 1.7E-04 1 0.3 0.1 0.03 0.01 0.003 probabilty 

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

4 1.0 0.02 2.74E-06 2.74E-06 8.21E-07 2.74E-07 8.21E-08 2.74E-08 8.21E-093 0.75 1.54E-06 1.54E-06 4.62E-07 1.54E-07 4.62E-08 1.54E-08 4.62E-09

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

1 0.25 5.13E-07 5.13E-07 1.54E-07 5.13E-08 1.54E-08 5.13E-09 1.54E-09

73 1.00 0.69 1.18E-04 1.18E-04 3.55E-05 1.18E-05 3.55E-06 1.18E-06 3.55E-07

6 0.03 7.12E-08 7.12E-08 2.14E-08 7.12E-09 2.14E-09 7.12E-10 2.14E-10

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

7 0.04 8.30E-08 8.30E-08 2.49E-08 8.30E-09 2.49E-09 8.30E-10 2.49E-10

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

53 0.88 1.82E-06 1.82E-06 5.45E-07 1.82E-07 5.45E-08 1.82E-08 5.45E-09

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

7 0.04 8.30E-08 8.30E-08 2.49E-08 8.30E-09 2.49E-09 8.30E-10 2.49E-106 1.00 0.02 4.10E-06 4.10E-06 1.23E-06 4.10E-07 1.23E-07 4.10E-08 1.23E-08

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

1 0.17 3.42E-07 3.42E-07 1.03E-07 3.42E-08 1.03E-08 3.42E-09 1.03E-09

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

1 0.17 3.42E-07 3.42E-07 1.03E-07 3.42E-08 1.03E-08 3.42E-09 1.03E-092 0.33 6.84E-07 6.84E-07 2.05E-07 6.84E-08 2.05E-08 6.84E-09 2.05E-09

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

2 0.33 6.84E-07 6.84E-07 2.05E-07 6.84E-08 2.05E-08 6.84E-09 2.05E-09
 BWR LOCA Frequency (for Pipes) Allocated by System, Degradation Mechanism, and Size Category

em (for BWRs) Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

e) Est. Pipe Sizes # of greater 

than

100 1500 5000 25000 100000 500000 gpm 

g. 

ch. 

LOCA Min Max Welds Total 0.5 1.8 3.3 7.3 18.4 41.2 dia. (in.) 

tionl 

Fraction (mm) (mm) Dom. LOCA 

Freq 

1 0.3 0.1 0.03 0.01 0.003 probabilty

ribu

on 

by 

System 

(in) (in) Total 9.4E-05 

0.09 10 550 8.0E-06 8.0E-06 2.4E-06 8.0E-07 2.4E-07 8.0E-08 0.0 

1 0.4 21.7 1 8.0E-06 8.0E-06 2.4E-06 8.0E-07 2.4E-07 8.0E-08 -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.19 1.5E-06 1.5E-06 4.5E-07 1.5E-07 4.5E-08 1.5E-08 -- 0.75 6.0E-06 6.0E-06 1.8E-06 6.0E-07 1.8E-07 6.0E-08 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.06 5.0E-07 5.0E-07 1.5E-07 5.0E-08 1.5E-08 5.0E-09 -- 

0.06 19 250 5.5E-06 5.5E-06 1.7E-06 5.5E-07 1.7E-07 0.0 0.0 

1 0.7 9.8 1 5.5E-06 5.5E-06 1.7E-06 5.5E-07 1.7E-07 -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.18 1.0E-06 1.0E-06 3.0E-07 1.0E-07 3.0E-08 -- -- 0.73 4.0E-06 4.0E-06 1.2E-06 4.0E-07 1.2E-07 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.09 5.0E-07 5.0E-07 1.5E-07 5.0E-08 1.5E-08 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 

0.04 8 300 4.0E-06 4.0E-06 1.2E-06 4.0E-07 1.2E-07 0.0 0.0 

1 0.3 11.8 1 4.0E-06 4.0E-06 1.2E-06 4.0E-07 1.2E-07 -- -- 
em (for BWRs) Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

e) Est. Pipe Sizes # of greater 

than

100 1500 5000 25000 100000 500000 gpm 

g. 

ch. 

LOCA Min Max Welds Total 0.5 1.8 3.3 7.3 18.4 41.2 dia. (in.) 

tionl 

Fraction (mm) (mm) Dom. LOCA 

Freq 

1 0.3 0.1 0.03 0.01 0.003 probabilty

ribu

on 

by 

System 

(in) (in) Total 9.4E-05 

0.00 123 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.88 3.5E-06 3.5E-06 1.1E-06 3.5E-07 1.1E-07 -- -- 0.13 5.0E-07 5.0E-07 1.5E-07 5.0E-08 1.5E-08 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 

0.11 10 500 1.0E-05 1.0E-05 3.0E-06 1.0E-06 3.0E-07 1.0E-07 0.0 

1 0.4 19.7 1 1.0E-05 1.0E-05 3.0E-06 1.0E-06 3.0E-07 1.0E-07 -- 

0.00 60 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.05 5.0E-07 5.0E-07 1.5E-07 5.0E-08 1.5E-08 5.0E-09 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.45 4.5E-06 4.5E-06 1.4E-06 4.5E-07 1.4E-07 4.5E-08 -- 0.25 2.5E-06 2.5E-06 7.5E-07 2.5E-07 7.5E-08 2.5E-08 -- 0.10 1.0E-06 1.0E-06 3.0E-07 1.0E-07 3.0E-08 1.0E-08 -- 0.15 1.5E-06 1.5E-06 4.5E-07 1.5E-07 4.5E-08 1.5E-08 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 

0.11 10 150 1.0E-05 1.0E-05 3.0E-06 1.0E-06 0.0 0.0 0.0 

1 0.4 5.9 1 1.0E-05 1.0E-05 3.0E-06 1.0E-06 -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.25 2.5E-06 2.5E-06 7.5E-07 2.5E-07 -- -- -- 0.20 2.0E-06 2.0E-06 6.0E-07 2.0E-07 -- -- -- 0.55 5.5E-06 5.5E-06 1.7E-06 5.5E-07 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

0.01 20 150 1.0E-06 1.0E-06 3.0E-07 1.0E-07 0.0 0.0 0.0 

1 0.8 5.9 1 1.0E-06 1.0E-06 3.0E-07 1.0E-07 -- -- -- 
em (for BWRs) Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

e) Est. Pipe Sizes # of greater 

than

100 1500 5000 25000 100000 500000 gpm 

g. 

ch. 

LOCA Min Max Welds Total 0.5 1.8 3.3 7.3 18.4 41.2 dia. (in.) 

tionl 

Fraction (mm) (mm) Dom. LOCA 

Freq 

1 0.3 0.1 0.03 0.01 0.003 probabilty

ribu

on 

by 

System 

(in) (in) Total 9.4E-05 

0.00 16 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 1.00 1.0E-06 1.0E-06 3.0E-07 1.0E-07 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- -- 

0.26 20 700 2.4E-05 2.4E-05 7.2E-06 2.4E-06 7.2E-07 2.4E-07 0.0E+00

0.8 27.6 1 2.4E-05 2.4E-05 7.2E-06 2.4E-06 7.2E-07 2.4E-07 0.0 

1 121 5.3E-05 5.3E-05 1.6E-05 5.3E-06 1.6E-06 5.3E-07 -- 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 

0.022 1.2E-06 1.2E-06 3.5E-07 1.2E-07 3.5E-08 1.2E-08 -- 

0.96 5.0E-05 5.0E-05 1.5E-05 5.0E-06 1.5E-06 5.0E-07 -- 0.02 1.2E-06 1.2E-06 3.5E-07 1.2E-07 3.5E-08 1.2E-08 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 

1 2.4E-06 2.4E-06 7.2E-07 2.4E-07 7.2E-08 2.4E-08 0.0 

0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.67 1.6E-06 1.6E-06 4.8E-07 1.6E-07 4.8E-08 1.6E-08 -- 0.33 8.0E-07 8.0E-07 2.4E-07 8.0E-08 2.4E-08 8.0E-09 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 
em (for BWRs) Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

e) Est. Pipe Sizes # of greater 

than

100 1500 5000 25000 100000 500000 gpm 

g. 

ch. 

LOCA Min Max Welds Total 0.5 1.8 3.3 7.3 18.4 41.2 dia. (in.) 

tionl 

Fraction (mm) (mm) Dom. LOCA 

Freq 

1 0.3 0.1 0.03 0.01 0.003 probabilty

ribu

on 

by 

System 

(in) (in) Total 9.4E-05 

0.21 19 600 2.0E-05 2.0E-05 5.9E-06 2.0E-06 5.9E-07 2.0E-07 0.0 

1 0.7 23.6 1 2.0E-05 2.0E-05 5.9E-06 2.0E-06 5.9E-07 2.0E-07 -- 

0.00 74 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.05 1.0E-06 1.0E-06 3.0E-07 1.0E-07 3.0E-08 1.0E-08 -- 0.85 1.7E-05 1.7E-05 5.0E-06 1.7E-06 5.0E-07 1.7E-07 -- 0.05 1.0E-06 1.0E-06 3.0E-07 1.0E-07 3.0E-08 1.0E-08 -- 0.05 1.0E-06 1.0E-06 3.0E-07 1.0E-07 3.0E-08 1.0E-08 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 

0.09 10 400 8.5E-06 8.5E-06 2.6E-06 8.5E-07 2.6E-07 0.0 0.0 

1 0.4 15.7 1 8.5E-06 8.5E-06 2.6E-06 8.5E-07 2.6E-07 -- -- 

0.00 72 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.06 5.0E-07 5.0E-07 1.5E-07 5.0E-08 1.5E-08 -- -- 0.35 3.0E-06 3.0E-06 9.0E-07 3.0E-07 9.0E-08 -- -- 0.18 1.5E-06 1.5E-06 4.5E-07 1.5E-07 4.5E-08 -- -- 0.35 3.0E-06 3.0E-06 9.0E-07 3.0E-07 9.0E-08 -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- -- 0.06 5.0E-07 5.0E-07 1.5E-07 5.0E-08 1.5E-08 -- -- 

0.02 20 75 1.5E-06 1.5E-06 4.5E-07 0.0 0.0 0.0 0.0 

1 0.8 3.0 1 1.5E-06 1.5E-06 4.5E-07 -- -- -- -- 

0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 1.00 1.5E-06 1.5E-06 4.5E-07 -- -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 0.00 0.0E+00 0.0E+00 0.0E+00 -- -- -- -- 
em (for BWRs) Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 

e) Est. Pipe Sizes # of greater 

than

100 1500 5000 25000 100000 500000 gpm 

g. 

ch. 

LOCA Min Max Welds Total 0.5 1.8 3.3 7.3 18.4 41.2 dia. (in.) 

tionl 

Fraction (mm) (mm) Dom. LOCA 

Freq 

1 0.3 0.1 0.03 0.01 0.003 probabilty

ribu

on 

by 

System 

(in) (in) Total 9.4E-05 

0.02 15 25 1.5E-06 1.5E-06 0.0 0.0 0.0 0.0 0.0 

1 0.6 1.0 1 1.5E-06 1.5E-06 -- -- -- -- -- 

0.00 0.0E+00 0.0E+00 -- -- -- -- -- 0.00 0.0E+00 0.0E+00 -- -- -- -- -- 0.00 0.0E+00 0.0E+00 -- -- -- -- -- 0.00 0.0E+00 0.0E+00 -- -- -- -- -- 0.67 1.0E-06 1.0E-06 -- -- -- -- -- 0.33 5.0E-07 5.0E-07 -- -- -- -- -- 0.00 0.0E+00 0.0E+00 -- -- -- -- -- 0.00 0.0E+00 0.0E+00 -- -- -- -- -- 0.00 0.0E+00 0.0E+00 -- -- -- -- -- 

0.01 20 100 5.0E-07 5.0E-07 1.5E-07 5.0E-08 0.0 0.0 0.0 

1 0.8 28.0 1 5.0E-07 5.0E-07 1.5E-07 5.0E-08 1.5E-08 5.0E-09 -- 

0.00 113.0 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 1.00 5.0E-07 5.0E-07 1.5E-07 5.0E-08 1.5E-08 5.0E-09 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 0.00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 0.0E+00 -- 

1.00 9.4E-05 9.4E-05 2.8E-05 9.1E-06 2.4E-06 6.2E-07 0.0E+00
Table E.7 BWR Passive Non-Pipe LOCA Frequency 

ontributors 

unt = 30 

Non-Pipe Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 LOCA 100 1500 5000 25000 100000 500000 gpm 

Mechanism Non-Pipe Freq 0.5 1.8 3.3 7.3 18.4 41.2 dia. (in.) 

ract Contr LOCA Fract 9.4E-05 1 0.3 0.1 0.03 0.01 0.003 probabilty 

1 0.03 3.15E-06 3.15E-06 9.44E-07 3.15E-07 9.44E-08 3.15E-08 0.00E+00

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+001.00 3.15E-06 3.15E-06 9.44E-07 3.15E-07 9.44E-08 3.15E-080.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

1 0.17 1.57E-05 1.57E-05 4.72E-06 1.57E-06 4.72E-07 1.57E-07 0.00E+00

0.60 1.89E-06 1.89E-06 5.66E-07 1.89E-07 5.66E-08 1.89E-080.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.40 1.26E-06 1.26E-06 3.78E-07 1.26E-07 3.78E-08 1.26E-080.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

1 0.80 7.55E-05 7.55E-05 2.27E-05 7.55E-06 2.27E-06 7.55E-07 2.27E-07

0.04 1.31E-07 1.31E-07 3.93E-08 1.31E-08 3.93E-09 1.31E-09 3.93E-100.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.96 3.02E-06 3.02E-06 9.05E-07 3.02E-07 9.05E-08 3.02E-08 9.05E-09
ontributors 

unt = 30 

Non-Pipe Cat-1 Cat-2 Cat-3 Cat-4 Cat-5 Cat-6 LOCA 100 1500 5000 25000 100000 500000 gpm 

Mechanism Non-Pipe Freq 0.5 1.8 3.3 7.3 18.4 41.2 dia. (in.) 

ract Contr LOCA Fract 9.4E-05 1 0.3 0.1 0.03 0.01 0.003 probabilty 

0.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+000.00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

APPENDIX F 

PIPING BASE CASE RESULTS OF 

DAVID HARRIS 

F-1

APPENDIX F 

PIPING BASE CASE RESULTS OF DAVID HARRIS 

Probabilistic Fracture Mechanics Analyses 

Performed in Support of LOCA Frequency 

Re-evaluation Effort 

D.O. Harris 

Engineering Mechanics Technology, Inc. 

San Jose, California

F.1 Introduction 

The purpose of this document is to report the procedures used and the results obtained in probabilistic 

fracture mechanics analyses of the base case systems considered in the LOCA Re-evaluation effort 

performed by use of expert elicitation by the Nuclear Regulatory Commission in the period February 

2003 – March 2004. The base case systems, which were defined in the kick-off meeting of the expert 

panel that was held in Rockville, Maryland in February 2003, consisted of the following: 

Pressurized Water Reactor

• hot leg (cast austenitic stainless steel) 

• surge line (austenitic stainless steel) 

• HPI makeup nozzle (austenitic stainless steel) 

Boiling Water Reactor

• recirculation line (austenitic stainless steel) 

• feedwater (carbon steel) 

These were identified as key systems that could serve as benchmarks for use by members of the expert 

panel in their estimation of LOCA frequencies. 

Piping isometrics of the base case systems and other systems identified in the kick-off meeting as 

important to estimations of flow rate probabilities were included in the FTP site that was set up for the use 

of panel members. Times in this appendix are in reactor-years (1 calendar year ~ 0.8 reactor years). 

F.2 Software 

The following discussion provides only a brief review of the PRAISE software. The references cited give 

the details. The results reported here were generated by use of the PRAISE software, which was 

developed with NRC support over a period of some 20 years. PRAISE is based on deterministic fracture 

mechanics, with some of the inputs considered as random variables. This allows the statistical 

distribution of lifetime to be computed, rather than a single deterministic failure time. The probability of 

failure (leaks of various sizes) is obtained from the computed lifetime distribution. 

Several versions of PRAISE were employed, depending on the nature of the problem. The original 

version of PRAISE [F.1] considers fatigue crack growth from crack-like weld defects introduced during 
F-2

fabrication. Semi-elliptical interior surface cracks are considered, usually circumferentially oriented. 

The initial crack size and fatigue crack growth properties are the major random variables, and Monte 

Carlo simulation is used to generate numerical results. Stratified sampling of crack depth and aspect ratio 

is employed to allow very small probabilities to be obtained without excessive computer time. Figure F.1 

is a schematic representation of the probabilistic fracture mechanics procedures used in the original 

version of PRAISE. The cumulative probability of a flow (leak) rate exceeding a specified size is 

generated by PRAISE as a function of time. If the stress history is specified in reactor-years, then the 

PRAISE results are also in reactor years. 

Figure F.1 Overview of PRAISE Methodology for Probabilistic Analysis of Fatigue Crack Growth 

A later version of PRAISE was used the analyze initiation and growth of stress corrosion cracks [F.2]. 

Both the fatigue crack growth and stress corrosion initiation and growth capabilities are included in 

pcPRAISE [F.3] and also in WinPRAISE [F.4], which is a Windows version that is much easier to use 

than the earlier DOS versions. WinPRAISE gives the same results as pcPRAISE for the same problem 

with the same inputs. 
F-3

In some cases, the cyclic stresses were such that fatigue crack initiation is the expected dominant 

degradation mechanism, and it was necessary to use a later version of PRAISE. This version was 

developed and used in Reference F.5, with additional capabilities described in Reference F.6 (such as the 

ability to consider detailed circumferential variations of the stresses). The fatigue crack initiation 

analyses employed probabilistic strain-life relations developed by Argonne National Laboratory [F.7, 

F.8]. Once a fatigue crack initiates, the original growth analysis capabilities in PRAISE are used for the 

crack growth portion of the lifetime. The depth of an initiated fatigue crack is taken to be 3.0 mm (0.12 

inches), in accordance with the ANL correlations, with a random surface length. In some instances, 

modifications to the source code of PRAISE were made to provide results of particular use for the 

problem at hand. These instances are discussed in the specific base case problem where they were 

employed. 

In cases where inspection was considered, the nondetection probability was represented by the expression 

1( )

2

1

P (a) = ε + + ε ND erfc 











*

ln

a

a

υ [F.1] 

In this expression, ε is the probability of not detecting a crack no matter how deep it is, a* is the crack 

depth having about a 50% chance of being detected, and ν controls the slope of the PND curve. “Good” 

and “outstanding” detection capabilities were considered [F.9], with the parameters given in Table F.1. 

Table F.1 Parameters in Non-Detection Relation 

Ferritic Austenitic 

 Fatigue Cracks Fatigue Cracks SCC Cracks 

outstanding 

a* 0.05h 0.05h 0.05h

ν 1.6 1.6 1.6 

ε 0.005 0.005 0.005 

good 

a* 0.15h 0.15h 0.4h

ν 1.6 1.6 1.6 

ε 0.02 0.02 0.10 

The operating and hydro pressures of Table F.2 were considered. 

Table F.2 Operating and Hydro Pressures, psi 

 Operating 

 Pressure 

Hydro 

Pressure 

PWR 2250 3125 

BWR 1250 1560 

A leak detection capability of 19 lpm (5 gpm) was considered, which means that any through-wall crack 

with a leak rate greater than 19 lpm (5 gpm) was immediately detected and removed from service. 
F-4

F.3 Post-Processing of PRAISE Results 

PRAISE analyses are performed on a piping location-by-location basis to provide the cumulative failure

probability as a function of time, whereas the desired end result is the system failure frequency within 

various time frames. The time increments of interest are now (0-25 years), the near future (25-40 years) 

and the more distant future (40-60 years). If the reactor transients analyzed are per reactor-year (rather 

than calendar year), then the times considered are also per reactor-year. As an average, one calendar year 

corresponds to 0.8 reactor years, but no adjustments for this are made in this appendix. The system 

failure frequency is obtained from the failure frequency for individual locations by analyzing the most 

highly stressed location and multiplying by the number of such highly stressed locations in the system. 

The failure frequency for a flow rate exceeding q& for a given time increment from t1 to t2 is obtained 

from the following relation 

2 1

2 1

2

( ) ( )

( )

t t

P t P t

p t

q q

q

−

−

=

& &

&

 [F.2] 

The values of P (t)

q&

 are the output from PRAISE for the dominant location(s) in the system. The failure 

frequency for the system is then obtained by multiplying by the number of locations in the system that

have the high stresses of the dominant location. 

F.3.1 Material Properties 

Numerous material properties enter into a PRAISE analysis, many of which are described in the 

references cited above. A Ramberg-Osgood stress-strain curve is used in the computation of the applied 

value of the J-integral, as represented by the following relation

n

E D













= +

σ σ

ε [F.3] 

Crack instability is governed by exceedance of a critical net section stress and/or a J-integral based 

tearing instability using a bilinear tearing resistance curve. 

A set of default material tensile and fracture properties are provided in WinPRAISE [F.4], which are 

summarized in Table F.3. Unless otherwise stated, these properties are used in the base case analyses. 
F-5

Table F.3 Summary of Default Material Properties 

Material Low Alloy Carbon Type 304 Type 316 

Flow Stress (normal) 

Mean, ksi 50. 52. 44.9 44.9 

Standard Deviation 2.1 2.2 1.9 1.9 

Tearing Instability Data 

D, ksi 146 154 106 106 

n 5 4 5 5 

JIc kips/in 0.6 1.5 5 5 

Tmat 60 45 300 300 

E, ksi 27300 27300 25800 25800 

ν 0.3 0.3 0.3 0.3 

Yield Strength, ksi 30.8 28.3 19.4 19.4 

Tensile Strength ksi 70.0 60.0 59.3 61.6 

Fatigue Crack Growth Properties 

(random) 

based on 

ASME Code 

see Reference 

F.1 or F.4 

The properties of Table F.3 are generally somewhat conservative and are representative of undegraded 

materials. In some instances, degraded material properties are considered, as discussed at the particular 

component involved. 

In the case of fatigue of initial cracks, the distributions of the initial crack depth and aspect ratio are 

probably the most important random variables. Unless otherwise stated, the depth distribution is taken 

from Reference F.10, which is the default in WinPRAISE and is also included in Reference F.9. 

F.3.2 Hot Leg 

The main coolant piping system is one of the base case systems. The failure probability of the large 

piping of this system is dominated by the hot leg to pressure vessel weld, because this location is at the 

highest temperature and sees the highest stress. 

F.3.2.1 Dimensions and Welds - From the piping isometrics made available to the panel members, the 

hot leg has a 29 inch inner diameter, and a thickness of 63.5 mm (2.5 inch) (OD=34 inches), fabricated

from SA-376 (which is an austenitic stainless steel). The example plant has two coolant loops. There are 

several welds in the hot leg, including shop and field welds and safe ends. There is a safe end and field 

weld at the pressure vessel. 

F.3.2.2 Stresses and Cycles - Table 1-2, page 9 of Reference F.1 summarizes the deadweight, 

pressure and restraint of thermal expansion stresses for the 14 field welds in one loop of the large primary 

piping in the plant considered in that report. Joint 1 is the hot leg to pressure vessel joint, and it has the 

highest stresses. The seismic stresses are included in Table 1-3, page 10 [F.1]. They are generally quite 

low. The postulated list of transients is provided in Table 4-1, page 152 [F.1], and is the transients

occurring over the 40 year design life, which corresponds to reactor years. The list contains 11 types of 

transients. There is sufficient information in Reference F.1 to consider all of these transients, but only the 

heat-up cool-down transient will be considered in this case, because it is the dominant transient 

contributing to fatigue crack growth [F.1]. The heat-up cool-down transient was postulated to occur 200 

times in 40 years (5/year). This is excessive, and 3/year is used herein. 
F-6

The seismic stresses are given in terms of the maximum load controlled stress (deadweight + pressure +

max seismic), and the summary of the stress history needed for fatigue crack growth analysis is also 

provided. This summary, denoted as S, is the sum of the cyclic stresses as follows 

= ∑ ∆

seismicstresshistory

S

2 2

max

4 σ σ [F.4] 

This is what controls the amount of fatigue crack growth during a seismic event for a fourth power crack 

growth law that includes R-ratio effects (See Reference F.1). Table F.4 summarizes the suggested stress 

history for the hot leg to pressure vessel joint. 

Table F.4 Summary of Stress History for Hot Leg to Pressure Vessel Joint 

deadweight stress = 2.08 ksi 

pressure stress = 6.49 ksi (axial) 

restraint of thermal expansion stress = 6.50 ksi 

3 times per year

max σLC 

ksi 

S

4

ksi4 

∆σ, 

ksi 

OBE 8.76 521.6 1.27 

SSE 9.06 2958.3 1.96 

3SSE 10.26 63430 4.22 

5SSE 10.62 162000 5.33 

The right-hand column in the above table is the cyclic stress if the seismic event contains 200 stress

cycles all of the same amplitude, with a low minimum load. This column is derived from the value of S

4

and Equation F.4, with σmax=∆σ, and is included just to provide an idea of the size of the seismic stresses. 

They are not large. 

Residual stresses, when considered, are taken to be the default values for large lines, as reported in

Reference F.2. 

F.3.2.3 Results - WinPRAISE runs were made for the hot leg to pressure vessel weld using the above 

stresses and default material properties. Table F.5 summarizes the results. “Good” inspections at 0, 20 

and 40 years were considered. These results are the cumulative leak probability. The left hand column 

gives the leak rate in gallons per minute. The next column in gives the time (25, 40 and 60 years), and the 

probabilities are directly from the PRAISE output for these times. For the Monte Carlo simulation, the 

crack size plane (a/h – a/b) was divided into 20 by 20 strata, with a maximum of 2000 trials drawn from 

each stratum. Sampling from a given stratum was stopped when 20 failures occurred in that stratum. 

Sampling began at the corner of the a/h-a/b plane corresponding to long deep cracks (1,0), and continued 

to shorter, then shallower cracks until no failures occurred in a stratum within 2000 trials. Sampling was 

then stopped. This procedure is referred to as automated stratification, and is a feature unique to 

WinPRAISE [F.4]. Earlier versions of the PRAISE software require the user to define each stratum and 

the sampling from each. 
F-7

Table F.5 Cumulative Probability PRAISE Results for Hot Leg-Pressure Vessel Weld for Fatigue 

Crack Growth from Pre-Existing Defects 

OD=34.0 inches, h=2.50 inches, cast austenitic, no σDL, times in reactor years 

 Base No hydro Aging 

Hydro yes no no 

Insp good good good 

tinsp 0,20,40 0,20,40 0,20,40 

Aging no no yes 

JIC 5 5 1.5 

dJ/da 23.44 23.44 15 

 no EQ SSE 5SSE no EQ SSE 5SSE no EQ SSE 5SSE 

25 1.20x10-18 2.34x10-16 2.45x10-16 6.61x10-15 7.04x10-15 7.08x10-15 1.43x10-14 1.47x10-14 1.47x10-14

40 1.29x10-18 2.35x10-16 2.45x10-16 6.61x10-15 7.04x10-15 7.08x10-15 1.43x10-14 1.47x10-14 1.47x10-14

60 1.29x10-18 6.01x10-18 6.19x10-18 6.61x10-15 6.62x10-15 6.62x10-15 1.43x10-14 1.44x10-14 1.44x10

>0 

-14

HLA0 HLB0 HLC0 

25 2.44x10-19 2.53x10-18 3.89x10-18 2.93x10-17 3.18x10-17 3.22x10-17 5.11x10-17 5.52x10-17 5.63x10-17

40 2.55x10-19 2.60x10-18 3.97x10-18 2.94x10-17 3.19x10-17 3.22x10-17 5.11x10-17 5.53x10-17 5.63x10-17

60 2.56x10-19 3.04x10-19 3.33x10-18 2.94x10-17 2.94x10-17 2.94710-17 2.12x10-17 5.12x10-17 5.13x10-17>100 

HLA1 HLB1 HLC1 

25 1.20x10-20 6.48x10-19 1.31x10-18 1.31x10-18 1.99x10-18 2.64x10-18 2.72x10-18 4.33x10-18 5.99x10-18

40 1.26x10-20 6.62x10-19 1.32x10-18 1.31x10-18 2.00x10-18 2.65x10-18 2.72x10-18 4.36x10-18 5.97x10-18

60 1.27x10-20 2.61x10-20 3.93x10-20 1.31x10-18 1.33x10-18 1.34x10-18 2.72x10-18 2.75x10-18 2.79x10-18

>1500 

HLA2 HLB2 HLC2 

25 1.19x10-20 6.48x10-19 1.31x10-18 1.31x10-18 1.99x10-18 2.64x10-18 2.72x10-18 4.33x10-18 5.99x10-18

40 1.26x10-20 6.62x10-19 1.32x10-18 1.31x10-18 2.00x10-18 2.65x10-18 2.72x10-18 4.36x10-18 5.97x10-18

60 1.27x10-20 2.61x10-20 3.93x10-20 1.31x10-18 1.33x10-18 1.34x10-18 2.72x10-18 2.75x10-18 2.79x10-18

>5000 

HLA3 HLB3 HLC3 

25 1.20x10-20 6.48x10-19 1.31x10-18 1.31x10-18 1.99x10-18 2.64x10-18 2.72x10-18 4.33x10-18 5.99x10-18

40 1.26x10-20 6.62x10-19 1.32x10-18 1.31x10-18 2.00x10-18 2.65x10-18 2.72x10-18 4.36x10-18 5.97x10-18

60 1.27x10-20 2.61x10-20 3.93x10-20 1.31x10-18 1.33x10-18 1.34x10-18 2.72x10-18 2.75x10-18 2.79x10-18

>500000 

HLA4 HLB4 HLC4 

noticeable effect of hydro 

noticeable effect of seismic when hydro test is performed, less effect when no hydro 

aging has about x2 effect 

>1500 gpm same as DEPB

Runs were made with and without a hydro test, and it is seen that hydro testing has a noticeable effect. 

Moderate material degradation is considered, with the values of JIc and (dJ/da)matl identified in the table. 

The failure probabilities are all very small, even the leak probabilities. The influence or seismic events is 

seen to be quite small. 

Table F.5 provides the base case results for the hot leg. Additional runs were made to study the following 

variables: 

• The effects of applying a load-controlled overload stress at a specified time were studied. This is 

called a design-limiting stress, and represents an overload event, such as water hammer or a seismic event 

even larger than the 5 SSE already considered for this component. 
F-8

• The effects of the fatigue crack growth relation employed were studied. The fatigue crack growth 

relation in PRAISE for austenitic stainless steel is based on information available during the original 

software development. More recent crack growth relations have been suggested [F.11]. For the simple 

stress history in this case, it is possible to run PRAISE with a crack growth relation that is equivalent to 

the more recent relation. 

• PWSCC crack initiation and growth has been identified in the control drive mechanisms (CRDM) 

in PWRs. This occurs in the Alloy 600 weldment. This alloy is also used in the safe end of the pressure 

vessel to main coolant piping welds, so is present in the hot leg to pressure vessel weld under 

consideration. In order to model the initiation and growth of PWSCC cracks, the initiation kinetics were 

assumed to be the same as for Type 316NG stainless steel as currently in PRAISE [F.2], but the crack 

growth kinetics were changed to be representative of Alloy 600. Based on information in Reference 

F.12, the crack growth kinetics is represented by the relation 

m CK

dt

da

= [F.5] 

where m equals 1.16, and C is lognormally distributed with a median value that depends on the 

temperature and material (weld, base metal, etc.). The median value of C for a weld at 315 C (600°F) is 

7.86x10-7, when crack growth rates are in inches/hour and K is in ksi-in1/2. Combining the within-heat 

and heat-to-heat variation in C, the second parameter of the lognormal distribution is 1.193 (standard 

deviation of lnC = 1.193). 

• The effects of more severe material degradation were studied, with the values of the degraded 

toughness given along with the results. Since PRAISE can not consider time-dependent material 

properties, the degraded material properties are present even in new pipe. The values of the degraded

properties are from Reference F.13. 

The results of these additional runs are summarized in Tables F.6 and F.7. 
F-9

Table F.6 Cumulative PRAISE Results Additional Runs for Hot Leg Pressure Vessel Weld 

 From 

Table F.5 

Ref F.11 

da/dN 

σDL @ 

 t-1 

PWSCC 

Growth 

no σres

PWSCC 

Growth 

σres

PWSCC 

Initiation 

σres 

Hydro yes yes yes yes yes -- 

Insp good good good good good good 

tinsp 0, 20,40 0,20,40 0,20,40 0,20,40 0,20,40 20,40 

Aging no no no no no no 

JIC 5 5 5 5 5 5 

dJ/da 23.44 23.44 23.44 23.44 23.44 23.44 

 no EQ no EQ no EQ no EQ no EQ no EQ 

25 1.20x10-18 2.20x10-18 2.38x10-16 0.923 0.916 0.001 

40 1.29x10-18 2.42x10-18 -- 0.926 0.918 0.020 

60 1.29x10-18 2.43x10-18 -- 0.926 0.919 0.068 

>0 

HLD0 HLE0 

25 2.44x10-19 2.61x10-19 2.38x10-16 7.97x10-7 2.16x10-7 1.0x10-5 

40 2.55x10-19 2.71x10-19 -- 7.97x10-7 2.16x10-7 2.69x10-4

60 2.56x10-19 2.72x10-19 -- 7.97x10-7 2.16x10-7 1.78x10-3>100 

HLD1 HLE1 

25 1.20x10-20 1.63x10-20 6.41x10-20 9.68x10-10 2.78x10-11 <10-4

40 1.26x10-20 1.65x10-20 -- 9.68x10-10 2.78x10-11 1.0x10-4

60 1.27x10-20 1.66x10-20 -- 9.68x10-10 2.78x10-11 4.85x10-4

>1500 

 HLD2 HLE2 

25 1.20x10-20 1.63x10-20 6.41x10-20 2.78x10-11 4.66x10-11 <10-5

40 1.26x10-20 1.65x10-20 -- 2.78x10-11 4.66x10-11 9.0x10-5

60 1.27x10-20 1.66x10-20 -- 2.78x10-11 4.66x10-11 3.77x10-4

>5000 

 HLD3 HLE3 

25 1.20x10-20 1.63x10-20 6.41x10-20 2.19x10-14 2.59x10-13 <10-5

40 1.26x10-20 1.65x10-20 -- 2.19x10-14 2.59x10-13 9.0x10-5

60 1.27x10-20 1.66x10-20 -- 2.19x10-14 2.59x10-13 3.77x10-4 

break 

 HLD4 HLE4 DEPB DEPB 

The design limiting stress was 40.0 MPa (4.49 ksi). In the case of PWSCC growth, initial fabrication 

defects were considered with the default depth distribution discussed above. Both initiation and growth 

were considered for the column identified as PWSCC initiation. The higher large leak rates for the 

initiation relative to the PWSCC growth are due to the possibility of multiple initiation sites, whereas the 

growth considers only one initial crack. 
F-10

Table F.7 Additional Hot Leg Pressure Vessel Runs Considering Material Aging 

OD=34 inches Good Inspection at 0, 20 ,40 Updated da/dN 

t=2.50 inches 3 HU-CD per year No Hydro Unless Specified 

σdw=2.08 ksi Type 304 Stainless 

σte=6.50 ksi Degraded Properties Used for All Times 

 Base no Hydro A B C D E 

JIc kips/in 5 1.11 0.67 1.72 0.75 0.20 

dJ/da ksi 23.44 13.4 8.0 22.6 6.5 0.05 

σys ksi 19.4 29.2 

σult ksi -- 76.7 

σflo ksi 44.9 53.0 

D ksi 106 104.5 

N 5 4.84 

25 1.20x10-18 6.61x10-15 1.34x10-14 1.96x10-14 9.73x10-15 2.07x10-14 4.02x10-13 

40 1.29x10-18 6.61x10-15 1.34x10-18 1.96x10-14 9.73x10-15 2.07x10-14 4.02x10-13

60 1.29x10-18 6.61x10-15 1.34x10-18 1.96x10-14 9.73x10-15 2.07x10-14 4.02x10

>0 

-13

 

25 2.44x10-19 2.93x10-17 5.26x10-18 6.76x10-17 -- -- 2.81x10-14 

40 2.55x10-19 2.94x10-17 5.27x10-18 6.77x10-17 -- -- 2.81x10-14

60 2.56x10-19 2.94x10-17 5.27x10-18 6.77x10-17 -- -- 2.81x10-14

>100 

 

25 1.20x10-20 1.31x10-18 2.67x10-18 4.30x10-18 1.48x10-18 5.31x10-18 2.81x10-14 

40 1.26x10-20 1.31x10-18 2.67x10-18 4.31x10-18 1.48x10-18 5.31x10-18 2.81x10-14

60 1.27x10-20 1.31x10-18 2.67x10-18 4.31x10-18 1.48x10-18 5.31x10-18 2.81x10-14

break 

earlier base 

case, default 

WinPRAISE 

properties, 

with hydro 

no hydro unaged weld 

metal J-T 

CF8M 

tensile 

mult J-T 

by 0.6 

all CF8M more 

sensitive 

aged 

extremely 

sensitive 

aged 

The biggest effect in Table F.7 is not having a hydro test. This assumption is necessary, because when 

degraded material properties are used, everything that fails does so during the hydro test. 

“Extremely sensitive aged” material properties are needed before degradation has a large effect. 

F.3.3 Surge Line 

The surge line is one of the base case systems. 

F.3.3.1 Dimensions and Welds - From the piping isometric available to the panel members, the surge 

line is a 14 inch line (14 inch outer diameter) with a thickness of 35.7 mm (1.406 inches). The material is 

SA376 Type 304, which is an austenitic stainless steel. There are some 13 welds in the line. 

F.3.3.2 Stresses and Cycles - The stresses at the surge line elbow are provided in Reference F.5, 

which is evidently the highest stressed location in the line. These stresses include seismic events and are 

given in Table F.8. The stress amplitude is contained in this table, which is one-half the stress range 

(peak-to-peak value). 
F-11

 Table F.8 Summary of Stress Cycles for Surge Line Elbow 

 (Stress Amplitudes with Seismic Stresses) 

Load Pair Amplitude 

(ksi) 

Number/ 

40 yr 

Load Pair Amplitude 

(ksi) 

Number/ 

40 yr 

HYDRO-EXTREME 190.17 6 9D-LEAK TEST 52.20 50 

8A-OBE 163.18 14 8G-LEAK TEST 52.20 65 

9B-OBE 162.06 14 8G-UPSET3 51.00 30 

8B-HYDRO 138.05 4 8G-12 50.96 90 

8B-OBE 127.94 10 8G-16 50.93 90 

9A-OBE 127.04 14 8E-8G 50.92 13 

8C-OBE 64.76 68 8E-OBE 43.38 77 

9F-OBE 64.17 68 9H-OBE 42.79 500 

8F-18 63.40 68 8H-13 39.82 90 

9C-11 63.38 68 8H-OBE 37.43 203 

8D-OBE 54.02 72 8H-UPSET4 35.42 40 

9G-OBE 53.42 400 8H-9E 33.94 90 

8G-18 52.38 22 2A-8H 33.94 77 

9D-11 52.35 22 3A-10A 33.10 4120 

8G-17 52.35 90 6-10A 33.10 200 

9D-LEAK TEST 52.20 50 3B-10A 33.10 4120 

8G-LEAK TEST 52.20 65 7-10A 33.10 4580 

8G-UPSET3 51.00 30 2B-SLUG1 32.87 100 

8G-12 50.96 90 2B-SLUG2 32.87 500 

 4B-10A 29.90 17040 

To estimate the influence of seismic events, it is necessary to also have the stress history without such 

events. It is not possible to remove seismic events knowing only the information in the above table. This 

information was provided in Reference F.14 and is summarized in Table F.9. 
F-12

 Table F.9 Summary of Stress Cycles for Surge Line Elbow 

 (Stress Amplitudes without Seismic Stresses) 

Load Pair Amplitude 

(ksi) 

Number/ 

40 yr 

Load Pair Amplitude 

(ksi) 

Number/

40 yr 

HYDRO-EXTREME 190.17 6 8G-16 50.93 90 

9B-HYDRO 149.86 4 8G-9H 50.92 128 

8A-UPSET 4 140.42 14 2A-8E 40.10 90 

9B-UPSET4 139.43 10 8H-9H 40.09 100 

8B-UPSET4 105.89 14 9H-10A 40.09 272 

9A-UPSET4 105.13 2 9E-13 39.82 90 

9A-LEAK 103.86 12 3A-10A 33.10 4120 

8F-18 63.40 68 6-10A 33.10 200 

9C-11 63.38 68 3B-10A 33.10 4120 

9F-LEAK 63.37 68 7-10A 33.10 4580 

8C-LEAK 63.37 35 2B-SLUG1 32.87 100 

2A-8C 62.30 33 2B-SLUG2 32.87 500 

8G-18 52.38 22 5-10A 29.90 9400 

8G-17 52.35 90 4A-10A 29.90 17040 

9D-11 52.35 22 4B-10A 29.90 17040 

2A-8D 51.20 72 2B-10A 20.60 14400 

8H-9G 51.18 400 2A-10A 20.60 14805 

8G-UPSET3 51.00 30 10A-UPSET1 20.59 70 

9D-12 50.96 50 10A-UPSET5 20.59 30 

8G-12 50.96 40 10A-UPSET6 20.59 5 

 10A-UPSET2 20.59 95 

 1B-10A 20.59 1533 

 1B-10B 20.00 87710 

The cyclic stress amplitudes of Tables F.8 and F.9 provide the information for the initiation analysis, but 

additional information is required for the growth portion of the analysis. The spatial gradient (primarily 

radial) is required. Also, when analyzing the stability of a through-wall crack, the steady normal 

operating stress is needed. This stress is considered to be the sum of the pressure, deadweight and 

restraint of thermal expansion stresses. The values of these latter two are given in Reference F.5 as

 σdw = 0 σte = 102.6 MP (14.88 ksi). 

Many of the high stress contributors in Tables F.8 and F.9 are from rapid excursions of the coolant 

temperature. The largest stress amplitude (half the peak-to-peak) is 1,310 MPa (190 ksi), so the stresses 

are large (but localized). These are the stresses at the peak stress location, which is not at weld. The 

spatial stress gradients (both along the surface and into the pipe wall) are required for a thorough analysis. 

The radial gradient (into the pipe wall) can be estimated by the procedure given in Section 5.3 of 

Reference F.5, i.e., 

The following specific rules were applied to assign stress to the uniform and gradient categories: 

• Cyclic stresses associated with seismic loads were treated as 100 percent uniform stress. 

• Cyclic stresses greater than 310 MPa (45 ksi) were treated as having a uniform component of 310 

MPa (45 ksi), and the remainder were assigned to the gradient category. 

• For those transients with more than 1000 cycles over a 40 year life, it was assumed that 50% of 

the stress was uniform stress and 50% a through-wall gradient stress. In addition, for these 

transients, the uniform stress component was not permitted to exceed 69 MPa (10 ksi). 
F-13

The gradient stress mentioned above is assumed to vary through the thickness as 













= − +

2

2

3

σ (ξ ) σ o 1 3ξ ξ [F.6] 

In this equation, σo is the stress at the inner wall of the pipe, ξ = x / h , x is the distance into the pipe wall 

from the inner surface, and h is the wall thickness. The stresses and cycles are high enough that fatigue 

crack initiation is important, which has been considered in Reference F.5, which shows a probability of 

0.981 of a leak in 40 years for this component. The LOCA probabilities will be less. The use of the 

gradient along the surface will reduce this. 

A refined stress analysis was available as part of the efforts reported in Reference F.6. These stresses 

included details of the variation of the stress in the circumferential direction, and are referred to as the 

“refined stresses”. The stresses used in the surge line evaluation were based on the actual stress analysis 

for a CE-designed plant in response to NRC Bulletin 88-11 dealing with surge line stratification. The

loadings were based on the methods approved by the NRC staff in the CE Owner Group Report CEN 

387-NP, "Pressurizer Surge Line Flow Stratification Evaluation," Rev. 1-NP, December 1991. Additional 

evaluations of the local stress distributions in the elbow were conducted to get the detailed stress 

distribution around the circumference of the elbow. The critically stressed location that produced the 

highest probability of cracking was the circumferential stresses in the side of the elbow due to 

stratification bending. Detailed stresses are not provided, because they belong to the plant that allowed us 

to use them. 

F.3.3.3 Results - PRAISE runs were made using the versions that can treat fatigue crack initiation. No 

inspections were considered. Since crack initiation is considered, there will be no effect of a pre-service 

inspection. The results are summarized in Table F.10.
F-14

Table F.10 Cumulative PRAISE Results for the Surge Line Elbow 

Condition Ref. F.5 Table F.8 

Stresses 

Table F.9 

Stresses 

Refined 

Stresses 

Seismic yes yes no yes 

σDL no no no no 

25 0.372 0.233 

40 0.982 0.772 0.587 8x10-7 

60 0.998 0.968 0.882 3.3x10

>0 -5 

 CENC4H1 

25 -- 1.6x10-5 7.5x10-6 <10-7 

40 -- 3.11x10-4 7.1x10-5 <10-7

60 1.33x10-3 2.51x10-4 <10-7

>100 

 -- CENC4D01 CENC4A3 20 hrs 

25 -- <10-7 <10-7 

40 -- <10-7 <10-7 

60 -- 2.0x10-7 1.0x10-7 

>1500 

 CENC4D15 CENC4A4 

 axisymmetric 

seismic 

axisymmetric 

nonseismic 

strain rates and 

bivariate stresses 

It is seen that the seismic stresses do not have a large effect, roughly a factor of 3. The use of the refined 

stresses greatly reduces the calculated failure probabilities. The computer run for 380 lpm (100 gpm)

took about 20 hours and resulted in no failures in 107

 trials. The runs for > 5,700 lpm (1,500 gpm) with

the stresses from Tables F.8 and F.9 had 2 and 1 failures in 107 trials, respectively, and these runs each 

took many hours. Hence, it is evident that the Monte Carlo simulation with multiple fatigue crack 

initiation sites does not allow definition of the small probabilities of large leaks in the surge line elbow, 

and an alternate procedure was developed. Stratified sampling is not used for fatigue crack initiation. 

F.3.3.4 Alternate Procedure - In cases where the dominant degradation mechanism is fatigue crack 

initiation with subsequent growth, PRAISE currently has no way of generating low probability results 

other than conventional Monte Carlo simulation. This is the dominant mechanism for three of the base 

line components; the surge line elbow, the HPI make up nozzle and the BWR feedwater line elbow. 

Excessive computer time is needed to generate probabilities of various size leaks for these components, 

with some runs taking 4 days on a 3 GHz pc, with no leaks of even 380 lpm (100 gpm). An alternate 

procedure is needed to estimate leak probabilities for the large leaks of interest, and such a procedure is 

described below. 

As part of a standard analysis, the PRAISE software computes the crack opening area and leak rate as 

functions of the length of through-wall cracks. Hence, this information is readily available, and can be 

used to determine the length of a through-wall crack needed to produce a given leak rate, such as 380 lpm 

(100 gpm), 5,700 lpm (1,500 gpm), etc. The probability of having a leak of a given magnitude is then the 

probability of having a through-wall crack exceeding that length. The half-crack length, b, is considered, 

which is a function of the desired leak rate, q& . Hence, b( q& ) can be considered as known. 

The probability of a double-ended-pipe-break (DEPB) is also of interest. In the cases of interest here, the 

critical net section stress failure criterion is used. For a through-wall crack, the value of b for a DEPB is 

given by the expression 
F-15

flo

LC

I

DEPB

R

b

σ

σ

π

=1− [F.7] 

where RI is the inside radius, σflo is the flow stress (average of yield and ultimate) and σLC is the load 

controlled stress, which is equal to the pressure plus deadweight stress. 

The version of PRAISE that performs Monte Carlo simulation of fatigue crack initiation and growth 

commonly provides information on the probability of having any leak and a leak exceeding a given 

magnitude. In order to have a nonzero number for the latter, a leak exceeding that magnitude must occur 

during the simulation. The problem is that this often does not occur within a number of trials that can be 

reasonably performed. In order to overcome this, PRAISE was modified to print out the length of any 

crack resulting in a leak and the time at which it first became through-wall. This was then used to 

estimate the size distribution of through-wall cracks as a function of time. The complementary 

cumulative distribution, denoted as Pb(>b), is concentrated upon. Then the probability of a leak greater 

than q& is given by 

P ( q) P [ b(q)] LK b > & = > & [F.8] 

Table F.11 summarizes the information from a PRAISE run using the stresses from Table F.9 for the 

crack opening area (A) and leak rate ( q& ) for a given half-length of a through-wall crack (b). 

Table F.11 Half Crack Lengths and Areas for a Given Leak Rate 

(Surge Line Elbow, Table F.9 Stresses) 

Q, 

gpm 

b, 

inches 

b

πRI 

A, 

in2 

A

Apipe 

100 5.981 0.445 0.936 0.010 

1500 10.379 0.591 10.028 0.143 

5000 11.791 0.671 46.762 0.476 

DEPB 15.95 0.907 -- -- 

A table of lengths of through-wall cracks was generated from the modified version of PRAISE using 104

trials using the stresses of Table F.9 (no seismic). In this run, there were 2,162 leaks within 25 years, 

5,932 within 40 years and 8,890 within 60 years. Dividing these numbers by 104

 provides leak 

probabilities that are nearly the same as obtained from the Monte Carlo simulation with 107 trials. Figure 

F.2 is the complementary cumulative distribution of leaking crack sizes for the three times of interest. 

The upper curve is for 60 years, because there is a higher probability of encountering a longer crack at 

this longer time. The lines in this figure are least squares curve fits, which are discussed later. 
F-16

1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 6.0

half length of through-wall crack, b, inches

10-5

10-4

10-3

10-2

10-1

100

probability of a through-wall crack longer than b

Figure F.2 Complementary Cumulative Distribution of Half-Crack Length of Through-Wall 

Cracks in Surge Line Elbow at 25, 40 and 60 Years 

Figure F.2 shows changes in slope, and the “lumpiness” of the distribution is readily apparent. This 

“lumpiness” is representative of a multi-modal probability density function of crack length, which is most 

likely due to the fatigue crack initiation sites being taken as 50 mm (2 inches) in length. That is, each 

two-inch segment around the circumference is taken as an independent initiation site. The surface length 

of an initiated crack is also a random variable. Once a crack initiates, it grows, and can link with 

neighboring cracks. This growth and linking can lead to sudden increases in crack length (by linking) and 

evidently is responsible for the multi-modal nature of the probability density function of crack length. 

The multi-modal nature of the probability density function is clearly shown in Figure F.3. 

0.00

0.02

0.04

0.06

0.08

0.10

p(b)

0 1 2 3 4 5 6

half-length of through-wall crack, b, inches

Figure F.3 Histogram of the Half-Length of Through-Wall Cracks at 60 Years for the Surge Line 

Elbow 

Pleasing curve fits to the lines in Figure F.2 are not possible. A good fit could perhaps be obtained by 

assuming that the histogram of Figure F.3 consists of a sum of lognormals with medians to match the 
F-17

location of the modes and relative weights adjusted to match the relative heights of their modes. This is 

believed to be unwarranted, since the multi-modal nature of the density function is an artifact of the

modeling assumption of a 50 mm (2 inch) long initiation site. It is better to just smooth out the 

cumulative distribution, and this was accomplished by a linear least squares curve fit to the cumulatives 

on log-linear scales. Since it is desired to represent the curve at long cracks, the fit was performed only 

for cracks corresponding to a probability less than 0.2. This eliminates the numerous cracks at higher 

probabilities that would skew the curve fit if included in the least squares calculations. The lines in 

Figure F.2 are the curve fits obtained in this manner. 

The assumed functional form was 

( ) 2.0

( ) 2.0

C b b

P b e

− −

> = [F.9] 

The values of C and b0.2 depend on the time. Once they are evaluated, the leak probabilities of a given 

size are obtained using the crack sizes in Table F.11. Table F.12 summarizes the results. 

Table F.12 Summary of Results for Surge Line Elbow (Table F.9 Stresses) 

Time, years 25 40 60 

Number of cracks 2162 5932 8890 

Cracks above 0.2 1730 4744 7109 

b0.2 2.072 2.089 2.108 

C 1.876 1.597 1.425 

 

P(>5.981) 1.31x10-4 4.00x10-4 8.02x10-4 

P(>10.379) 3.41x10-8 3.56x10-7 1.52x10-6

P(>11.791) 2.42x10-9 3.73x10-8 2.04x10-7

P(>15.95) 9.86x10-13 4.86x10-11 5.43x10-10 

Table F.13 summarizes the results along with corresponding ones obtained directly from the Monte Carlo

simulation. The conventional Monte Carlo simulation used 106 trials for 380 lpm (100 gpm) and 107

trials for 5,700 lpm (1,500 gpm). 
F-18

Table F.13 Cumulative PRAISE Results for the Surge Line Elbow as Obtained from the Alternate 

Procedure and Directly from Monte Carlo Simulation 

(Table F.9 Stresses) 

Table F.13 shows that the alternate procedure is able to greatly extend the leak rates whose probabilities 

can be estimated. In cases where direct comparisons are possible, the alternate procedure gives higher 

leak probabilities. The direct Monte Carlo for 5,700 lpm (1,500 gpm) employed 107

 trials and took 36 

hours of computer time. The alternate procedure used 104

 trials, so took about 2 minutes. Even in this era 

of fast cheap computer time, it would still be prohibitive to use direct Monte Carlo to generate the results 

obtained by the alternate procedure. It would take 1010 trials to produce the DEPB results in the above 

table. This translates to 36,000 hours of computer time, or about 4 years. 

F.3.4 HPI Makeup Nozzle 

An HPI/makeup nozzle safe end from a B&W plant type was selected as one of the base case systems. 

F.3.4.1 Dimensions and Welds - This type of component was considered in Reference F. 5, which 

identifies the component as 2 1⁄2 inch schedule 160 pipe fabricated from Type 304 austenitic stainless 

steel. The location considered in Reference F.5 is in the safe end at the nozzle, which has a thickness of 

11.1 mm (0.4375 inches) and a mean radius of 32.5 mm (1.28 inches) at the location of high stresses. 

F.3.4.2 Stresses and Cycles - As shown in Reference F.5, the cyclic stress history is dominated by 

two types of transients, with the amplitudes and frequencies shown in Table F.14. 

Table F.14 Stress History for HPI/Make Up Nozzle 

from NUREG/CR-6674 [F.5] 

Name Stress 

Amplitude 

ksi 

Number in 

40 years 

HPI actuation A/B 221.24 33 

Test Null 169.31 7 

 Direct 

Monte Carlo 

Alternate 

Procedure 

25 0.233 0.216 

>0 

40 0.587 0.593 

60 0.882 0.889 

25 7.5x10-6 1.31x10-4

40 7.1x10-5 4.00x10-4

>100 

60 2.51x10-4 8.02x10-4

25 <10-7 3.41x10-8

40 <10-7 3.56x10-7

>1500 

60 1.0X10-7 1.52x10-6

25 -- 2.42x10-9

40 -- 3.73x10-8

>5000 

60 -- 2.04x10-7

25 -- 9.86x10-13

40 -- 4.86x10-11

DEPB 

60 -- 5.43x10-10
F-19

The deadweight and restraint of thermal expansion stresses for this location under normal operation that 

were used in Reference F.5 are 

σdw=0 

σte= 63.1 MPa (9.16 ksi) 

As discussed above, these stresses were composed of 310 MPa (45 ksi) uniform and the remainder the 

generic gradient of Equation F.6. These stresses are believed to be very conservative and are for the

thermal sleeve being intact. 

F.3.4.3 Results - The version of PRAISE that considers fatigue crack initiation was run for the 

HPI/make up nozzle. The stresses of Table F.14 were taken to be axisymmetric. Due to the small line 

size, only 4 initiation sites around the circumference were considered. Table F.15 summarizes the results. 

Table F.15 Cumulative Probability PRAISE Results for HPI/Make Up Nozzle 

(Intact Thermal Sleeve) 

Condition From 

Reference F.5 

Here 

25 1.004x10-5 

40 0.00210 6.08x10-4 

60 0.0309 1.04x10-2 

>0 

 Inel4a2 

25 -- 4.5x10-8 

40 -- 4.9x10-7 

60 -- 1.79x10-5

>100 

 Inel4a1 

25 -- 2.0x10-8 

40 -- 2.10x10-7 

60 -- 4.56x10-6 

>1500 

 Inel4a2 

Table F.15 shows a cumulative leak probability of 10

-5 in 25 years, which is quite low. However, leaks in 

this component have been observed in service, in which case the thermal sleeve in the component was 

failed. The results of Table F.15 use the stresses for an intact sleeve, and the stresses will be altered if the 

sleeve fails. A failed thermal sleeve is now considered. 

F.3.4.4 Failed Thermal Sleeve - There is a thermal sleeve at the HPI nozzle, and the results in Table 

F.15 are for the case of the thermal sleeve not failing. The thermal sleeve has been observed to fail in 

service, which changes the stresses in the component. 

In order to model the failure of the thermal sleeve, the following steps were taken: 

1. Once the thermal sleeve fails, assume that a crack of the “initiation size” immediately appears. This 

size is a depth of 3.0 mm (0.12 inches). The WinPRAISE default distribution of the aspect ratio is used, 

as in other components. 
F-20

2. A WinPRAISE run with this initial crack is performed, with the stresses that were present before the 

crack initiated (Table F.14), plus a uniform cyclic stress cycling each hour of sufficient amplitude to 

result in a high leak probability at not long times. This defines the uniform stress. 

3. Use WinPRAISE to compute the leak frequencies for larger leak rates. 

This procedure provides the results shown in Table F.16. 

Table F.16 Cumulative PRAISE Results for HPI/Make Up Nozzle with 

Failed Thermal Sleeve and Additional Uniform Cyclic Stress, σu

 Intact 

Sleeve 

With Initial 

Crack and 

Original 

Stresses, 

σu=0 

With Initial 

Crack and 

σu = 8 ksi 

With Initial 

Crack and 

σu = 12 ksi 

With Initial 

Crack and 

σu = 25 ksi 

5 -- 5.67x10-5 <10-2 0.047 0.18 

25 1.004x10-5 3.69x10-3 0.032 0.14 0.727 

40 6.08x10-4 1.26x10-2 0.129 0.33 0.909 

60 1.04x10-2 2.98x10-2 0.161 0.47 0.909 

>0 

 

25 4.5x10-8 6.49x10-4 <10-5

40 4.9x10-7 2.68x10-3 <10-5

60 1.79x10-5 5.31x10-3 <10-5

>100 

 

25 2.0x10-8 -- 

40 2.10x10-7 -- 

60 4.56x10-6 -- 

>1500 

 

25 -- 6.49x10-4 

40 -- 2.68x10-3

break 

60 -- 5.31x10-3

Table F.16 shows that a uniform stress of some 170 MPa (25 ksi) is needed to result in an appreciable 

leak probability within 25 years. However, the frequency of larger leak rates is actually reduced by 

imposing the uniform stress that is necessary to produce the high leak probabilities seen in service. This 

uniform stress grows cracks to leaks, so that the larger leak rate frequencies are reduced. The least

favorable condition for larger leaks is a failed thermal sleeve with the original stresses (σu = 0). 

F.3.5 Recirculation Line – 12 inch 

The recirculation line is one of the base case systems for a BWR. This system has developed leaks in the 

past due to intergranular stress corrosion cracking (IGSCC). The 12 inch line has some of the highest

stresses, so is considered here. The recirculation system also has 28 inch lines, which can contribute to 

larger flow rate failures than possible from a 12 inch line. Hence, the 28 inch line is also considered in 

subsequent sections. 

F.3.5.1 Dimensions and Welds - The layout of the recirculation system is given in isometrics made 

available to panel members. There are two recirculation loops, which are very similar to one another. 

There are 121 welds in this system, including field welds, shop welds and safe ends. The piping is 
F-21

fabricated from A-358 Class 1 Type 304, and the piping is of diameters 12, 22 and 28 inches – all 

schedule 80. 

F.3.5.2 Stresses and Cycles - IGSCC will be the dominant degradation mechanism. Hence, time at 

stress is of major concern, and the number of stress cycles is of secondary importance. Estimated stresses 

at the highest stressed locations for the two pipe sizes of interest are given in Table F.17. 

Table F.17 Stress Information for Two Recirculation Joints 

OD, inch Thickness, inch σNO, ksi Seismic σ, ksi 

12.75 0.687 20.41 20.41 

28 1.201 9.48 10.60 

The normal operating stress (σNO) is the sum of the pressure stress, deadweight stress and restraint of 

thermal expansion stress. A value of 14 MPa (2 ksi) for the deadweight stress is assumed. The seismic 

stress is the normal operating stress plus the seismic-induced stress. Note that the seismic stresses are 

small in this case. The magnitude of the seismic event is unknown. 

The time at stress is important for this case, with the cycling frequency being of less importance. 

Consistent with what is used for the PWR, the cycling is considered to be composed of heat up and cool

down at 3 per year. The parameters related to stress corrosion cracking are summarized in Table F.18.

Table F.18 Stress Corrosion Cracking Parameters

Oxygen at startup (PPM) = 8.0 

Oxygen at steady state (PPM) = 0.20 

Heat up (100-550F) time (hrs) = 5.00 

Coolant conductivity (μs/cm) = 0.20 

Degree of sensitization (C/cm2) = 7.04

Residual stresses will be important, and the default residual stress distributions in pcPRAISE, which are 

documented in Reference F.2, are used when no remedial treatments are performed. In order to include 

remedial treatments that have been performed in service, a weld overlay at 20 years will be considered. 

This alters the thickness, crack growth kinetics (post-treatment analyses use Type 316NG crack growth 

defaults in PRAISE) and residual stresses. The axisymmetric through-wall residual stress distribution of 

Figure F.4 is employed. This figure is from Reference F.15. PRAISE can not treat the actual gradient, so 

the linear approximation in this figure is used. The linear gradient employed underestimates the 

beneficial effect of the weld overlay. 
F-22

Figure F.4 Through-Wall Residual Axial Stress Distribution from Weld Overlay [F.15] 

F.3.5.3 Results - Table F.19 summarizes the results obtained for the 12 inch weld in the recirculation 

system. 

Table F.19 Cumulative Probability PRAISE Results for the 12 inch Recirculation Line Weld, with 

and without Weld Overlay at 20 Years (σno = 141 MPa [20.41 ksi]) 

OD=12.75 inches, h=0.687 inches, wrought austenitic, 

stress corrosion crack initiation and growth 

 Base Overlay at 

20 years 

Overlay & 

σDL @ 39 years 

25 0.3674 0.2967 0.2968

40 0.5986 0.3803 0.3872

60 0.7435 0.4241 0.4253

>0 

 

25 0.1682 0.1427 0.1429

40 0.2452 0.1622 0.1632

60 0.2872 0.1693 0.1708

>100 

 

25 0.1529 0.1066 0.1078

40 0.2193 0.1250 0.1276

60 0.2534 0.1312 0.1343

>1500 

 

25 0.1529 0.0490 0.0502

40 0.2193 0.0674 0.0700

60 0.2535 0.0736 0.0767

break 

 

5000 trials 304 full residual stress 

σdw=2.0 ksi σte=13.32 ksi σDL=11.67ksi 3 HU-CD/yr p=1125 psi 
F-23

The beneficial effect of the weld overlay at 20 years is not readily apparent from the results in Table F.19; 

such benefits are shown more clearly in Figure F.5, which provides a plot of the cumulative probability of 

a leak exceeding 380 lpm (100 gpm) as a function of time. 

overlay

no overlay

12 in recirc, overlay at 20 years

0 10 20 30 40 50 60

time, years

0.00

0.05

0.10

0.15

0.20

0.25

0.30

cumulative probability of leak >100 gpm

Figure F.5 Cumulative Probability of a Leak Exceeding 100 gpm as Functions of Time for the 12 

inch Recirculation Line Weld with and without Weld Overlay at 20 Years 

The slopes of the lines in Figure F.5 are the leak frequencies, and the slope at 40 years with no overlay is 

about 7 times that with overlay. 

F.3.5.4 Summary of Observations from Service - Leak frequencies due to IGSCC in recirculation 

lines were estimated from service experience and reported in Reference F.16. Figure F.6 is Figure F.12 

from that reference. With some exceptions, the results in Figure F.6 are between 10-4 and 10-3 per weldyear. The results are for times up to 15 years and do not include remedial actions. No strong 

dependencies on time or line size are apparent, but the smaller diameter lines appear to have a somewhat 

higher failure frequency. 

Table F.20, which is from Charts 2 and 3 of Reference F.17, summarizes the depth distribution of 

observed cracks per weld-year for various pipe sizes in recirculation lines in BWRs. The remedial action 

of Reference F.17 is considered to consist of a weld overlay at 20 years. Observed crack sizes without 

remedial action, as reported in Reference F.16, are shown in Figure F.7. 
F-24

1.00E-05

1.00E-04

1.00E-03

1.00E-02

0 2 4 6 8 10 12 14

Year(s) of Operation

λ [1/weld-yr.]

3 < ∅ ≤ 6

6 < ∅ ≤ 12

12 < ∅ ≤ 22

∅ > 22

Figure F.6 Leak Frequencies as a Function of Time and Pipe Size (from Reference F.16) 

Table F.20 Observed Crack Depth Frequencies in Various Line Sizes in Recirculation Lines as 

Percentages of the Wall Thickness (from Reference F.17) 

No Remedial Action 

Size > 10% > 20% > 30% > 40% > 50% > 60% > 70% > 80% > 90% 

NPS12 2.06E-03 1.62E-03 7.28E-04 3.64E-04 2.00E-04 1.46E-04 1.09E-04 7.28E-05 3.64E-05

NPS22 1.63E-03 1.11E-03 6.48E-04 3.21E-04 1.90E-04 1.24E-04 9.81E-05 6.54E-05 3.27E-05

NPS28 2.12E-03 1.50E-03 1.04E-03 5.99E-04 2.57E-04 1.84E-04 6.12E-05 3.67E-05 1.22E-05

With Remedial Action 

Size > 10% > 20% > 30% > 40% > 50% > 60% > 70% > 80% > 90% 

NPS12 1.95E-04 1.60E-04 1.04E-04 8.31E-05 6.73E-05 4.61E-05 2.78E-05 1.90E-05 1.03E-05

NPS22 3.29E-04 2.74E-04 1.70E-04 1.32E-04 1.01E-04 8.62E-05 4.43E-05 2.95E-05 1.48E-05

NPS28 3.95E-04 2.84E-04 1.77E-04 9.15E-05 6.66E-05 3.82E-05 2.08E-05 1.24E-05 3.95E-06
F-25

0.00%

10.00%

20.00%

30.00%

40.00%

50.00%

60.00%

70.00%

80.00%

90.00%

100.00%

0.0% 10.0% 20.0% 30.0% 40.0% 50.0% 60.0% 70.0% 80.0% 90.0% 100.0%

Crack Length / Pipe Circumference

a/t-ratio

Figure F.7 Observed Crack Sizes as Reported in Reference F.16 

F.3.5.5 Comparisons with PRAISE - The normal operating stress in Table F.17 of 20.41 ksi is for the 

highest stressed joint in the 12 inch recirculation line, whereas the observations are for all joints,

including lower stressed locations. In order to generate PRAISE results that would be more 

representative of the population, runs were made with various stresses. Table F.21 summarizes the 

results. 

Table F.21 Cumulative PRAISE Results for a 12 inch Recirculation Line Weld 

for Various Normal Operating Stresses (Remedial Action at 20 Years) 

 Cumulative 

Mean σno 10 12 15 20 

 COV 0.0 0.3 0.0 0.0 

Mean σte 3.32 5.32 8.32 13.32 

25 1.42x10-3 1.54x10-2 9.36x10-2 0.2967

40 1.46x10-3 1.89x10-2 0.1473 0.3803

60 1.46x10-3 2.08x10-2 0.1781 0.4241

>0 

 

25 4.90x10-4 7.59x10-3 3.90x10-2 0.1427

40 4.90x10-4 8.86x10-3 5.35x10-2 0.1622

60 4.90x10-4 9.48x10-3 6.11x10-2 0.1693

>100 

 

25 3.50x10-4 5.80x10-3 3.19x10-2 0.1066

40 3.50x10-4 7.06x10-3 4.53x10-2 0.1250

60 3.50x10-4 7.684x10-3 5.27x10-2 0.1312

>5000 

25 1.00x10-4 2.70x10-3 2.12x10-2 0.0490

40 1.00x10-4 2.96x10-3 3.46x10-2 0.0674

DEPB 

60 1.00x10-4 4.58x10-3 4.20x10-2 0.0736
F-26

The results for 138 MPa (20 ksi) correspond to those in Table F.19. The normal operating stress was 

taken to be deterministic, except for the case of 83 MPa (12 ksi), in which case the normal operating 

stress is normally distributed with a mean of 83 MPa (12 ksi) and a standard deviation of 0.3x(36.7+13.8) 

MPa (0.3x(5.32+2.00) ksi) = 15.2 MPa (2.20 ksi). 

The cumulative results from Table F.21 can be compared with the observed frequencies in Table F.20 by 

converting the cumulative results to a frequency by dividing by the time increment involved. In the 

current case, the increase in the cumulative following the remedial action is relatively small, as seen from 

Figure F.5. Hence, the cumulative results at 25 years from Table F.21 should be divided by 20 to provide 

frequencies for comparison purposes. This provides the results in Table F.22. 

Table F.22 Estimated Leak Frequencies Prior to Remedial Action, from Table F.21 

Mean σno 10 12 15 20 

COV σno 0.0 0.3 0.0 0.0 

Mean σte 3.32 5.32 8.32 13.32 

Frequency 7.11x10-5 7.69x10-4 4.68x10-3 1.48x10-2 

This table shows that the mean normal operating stress of 83 MPa (12 ksi) with some variance provides 

the best agreement with the results of Figure F.6. This is the case that will be used for benchmarking 

against observed cracks. 

The following steps were followed in order to provide PRAISE results for comparison with observations 

of part-through cracks: 

 1. The WinPRAISE software was modified to print out the sizes of cracks present at each time step in

the analysis. The depth and length of the deepest crack and the longest crack at that time step are printed 

into a file, along with the number of cracks present at that time. This file contains at most a number of 

lines equal to the number of Monte Carlo trials times the number of time steps (which can be a lot of 

lines). 

2. The WinPRAISE file from step 1 is then processed to provide another file that includes only the sizes 

of part-through cracks present at the time of interest (25 years in this case). (Cracks of zero depth, leaks 

and other times are eliminated.) 

3. The crack size file from step 2 is then loaded into a histogram, which provides the number of cracks 

present at 25 years that fall within a certain depth range. 

4. Since Reference F.17 reports detected cracks, the detection probability (Equation F.1) must be 

accounted for. This is accomplished by multiplying the number of cracks in each bin by the detection 

probability for a crack of depth equal to the midpoint of the bin. This provides the number of detected 

cracks in this bin. The contents of each bin are then divided by the number of trials times the time (25 

years) to provide the crack sizes per weld year. 

5. The histogram is then converted to a complementary cumulative form, which is then directly 

comparable to results from Reference F.17. 

Figure F.8 presents the crack size results for the benchmark case. Once again, not many deep cracks are 

observed. A pattern is observed in Figure F.8 which shows a preponderance of cracks below about 2.5 

mm (0.1 inches). This pattern is due to cracks growing to a depth of 2.5 mm (0.1 inch) and then slowing 
F-27

down or arresting, which is most likely due to the transitioning from growth of “initiating cracks” to

“fracture mechanics cracks” that occurs in the PRAISE modeling of initiation and growth. The 

transitioning criteria are discussed on page 42 of Reference F.2, and one of the criteria is “If the depth of 

the crack is greater than 2.5 mm (0.1 inch), its growth will always be by fracture mechanics velocity”. 

Figure F.9 is a plot of the predicted complementary cumulative number of observed cracks for the 

benchmark case, along with a comparison with reported observations. The outstanding inspection 

parameters of Table F.1 were employed. In this figure, Reference F.17 results from Chart 1 (prior) and 

Chart 2 (posterior) are both shown, since the analysis mixed with and without remedial action (weld 

overlay at 20 years). The analysis results fall midway between the two results, except for shallow cracks. 

0.0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0

half surface crack length, b, inches

0.0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

crack depth, a, inches

Figure F.8 Crack Sizes After 25 Years for the Benchmark Case (Mean σNO = 12 ksi) with Weld 

Overlay at 20 Years
F-28

mean sig NO=12

0.0 0.2 0.4 0.6 0.8 1.0

a/h

10-5

10-4

10-3

10-2

cracks >a/h, per weld year

prior

post

Figure F.9 Comparison of Results for the Benchmark Case Predicted for Outstanding Inspection 

Quality with Reported Prior and Post Observations [F.17] 

The agreement shown in Figure F.9 is felt to be quite good, and indicates that the PRAISE model best fits 

the observed crack depths when the mean stress of 83 MPa (12 ksi) is used. Figure F.10 shows that the

stress has an important effect, because the agreement is not so good when a stress of 103 MPa (15 ksi) is 

employed. 

outstanding

good

0.0 0.2 0.4 0.6 0.8 1.0

a/h

10-6

10-5

10-4

10-3

10-2

cracks >a/h, per weld year

12 in, med stress random

median σte=8.32 ksi, std dev = 3.33 ksi 

Figure F.10 Comparison of Results for the High-Stress Case Predicted for Two Inspection 

Qualities with Reported Observations (the Square Data Points are the Observations [F.17], no 

Weld Overlay) 
F-29

The results of Figure F.7 are for observed (detected) cracks, whereas Figure F.8 has not had the 

nondetection probability applied. A direct comparison is therefore not possible, and the two figures are 

plotted on much different scales. However, the PRAISE predictions contain a much greater proportion of 

short and shallow cracks than the observations. This could be somewhat affected by the nondetection 

probabilities, but the differences would not be removed by applying the nondetection probabilities to the 

cracks predicted by PRAISE. 

The question immediately arises regarding the results to be used in the estimation of the recirculation 

system reliability; the results of Table F.20 for the highly stressed joint or the results of Table F.21 for the 

benchmarked stress of 83 MPa (12 ksi). Interestingly, the system and average weld leak frequencies are 

nearly the same whether the 83 MPa (12 ksi) or 140 MPa (20 ksi) weld is used, because the number of 

joints involved also depends on the stress. Table F.23 summarizes this comparison. This table is in terms 

of the leak frequency per year, which is obtained from the cumulative results given above by use of 

Equation F.2. 

The system frequency and average frequency per weld are nearly the same for both cases. 

Table F.23 System and Average Weld Leak Frequencies for Two Cases 

of the 12 inch Recirculation Line

Mean σNO, ksi 12 20 

Mean σte, ksi 5.32 13.32 

COV 0.3 0 

Std Dev of σte, ksi 1.6 0 

 per weld joint 

0 - 25 years 6.15x10-4 1.19x10-2

25 – 40 years 2.36x10-4 5.57x10-3

40 – 60 years 9.25x10-4 2.19x10-3

 

Number dominant. joints 49 2 

Number in system 49 49 

 System (times number dom. joints) 

0-25 3.01x10-2 2.38x10-2

25-40 1.16x10-2 1.11x10-2

40-60 4.53x10-3 4.38x10-3

 Average per joint ( ÷49) 

0-25 6.15x10-4 4.86x10-4

25-40 2.36x10-4 2.24x10-4

40-60 9.25x10-4 8.94x10-5

Of the two cases in Table F.23, the case of a mean stress of 83 MPa (12 ksi) and coefficient of variation 

of 0.3 (on σte + σdw) is more representative of the population of joints as a whole, so is preferred for 

comparisons with observations of part-through cracks. 

F.3.6 Recirculation Line– 28 inch 

Stresses and dimensions are given in the corresponding sections for the 12 inch line. IGSCC crack 

initiation and growth are the dominant degradation mechanisms. Table F.24 summarizes the results for 

this weld. 
F-30

Table F.24 Cumulative PRAISE Results for the Weld 

 in the 28 inch Recirculation Line

OD=28 inches t = 1.201 inches 

 Time Probability 

25 6.23x10-3 

40 1.02x10-2 

60- 1.46x10-2

>0 

25 6x10-4

40 8x10-4

60- 8x10-4

>100 

25 6.66x10-5 

40 6.66x10-5

60- 1.25x10-4 

>1500 

25 6.00x10-5

40 6.87x10-5

60- 9.79x10-5

>5000 

25 3.3x10-5

40 3.3x10-5

60- 6.7x10-5

break 

σdw=2.0 ksi 

σte=1.75 ksi 

P = 1,125 psi 

Type 304 full 

residual stress 

3 HU-CD/yr

F.3.7 Feedwater Elbow 

The feedwater elbow is one of the base case systems. This system is subject to flow accelerated corrosion 

(FAC), which can be a serious degradation mechanism if left unchecked. PRAISE can not model FAC, 

but some analyses are provided for fatigue crack initiation and growth. 

F.3.7.1 Dimensions and Welds - The layout of the feedwater system is given in the piping isometrics 

made available to the panel members. There are some 123 welds in the two loops of the feedwater 

systems, all but 6 of them in 12 and 20 inch piping. The 12 inch lines are schedule 100 (17.4 mm [0.687 

inches] thick) and the 20 inch lines are schedule 80 (32.5 mm [1.281 inches] thick). The material is A333 Grade 6 (which is a carbon steel). 

F.3.7.2 Stresses and Cycles - The feedwater line elbow is considered in Reference F.5, so this is 

evidently the high stress point in the system. Note that there are at least 6 such elbows in a feedwater 

system. (There are many more elbows, but they are likely to not be so highly stressed). The degradation 

mechanism is fatigue and flow accelerated corrosion (FAC). Stresses do not contribute to FAC, so are 

not needed for this mechanism. For fatigue, there are a considerable number of cycles of high stress 

amplitude. They are available from Reference F.5. Table F.25, which (except for the column of 

temperatures) is page A.25 of Reference F.5, summarizes the stresses. These stresses are “decomposed”

according to the procedure discussed above for the surge line. The analysis reported in Reference F.5 
F-31

used a temperature of 590°F (310°C), as indicated in the text at the top of Table F.25. However, Table 5-

123 of Reference F.18 provides the temperatures for these transients, and it is suggested that these 

temperatures be used, because their use is more realistic and less conservative. They are included as the 

right-hand column of Table F.25. The temperature influences the strain-life curve, and has a noticeable 

effect on the computed failure probabilities because of its influence on the initiation probabilities.

The values of the deadweight and restraint of thermal expansion under normal operation that Reference 

F.5 uses for this location are 

σdw = 0 

σte = 115 MPa (16.68 ksi). 

The stress history in Table F.25 most likely contains seismic events. It is not possible to eliminate them 

from the list using information currently available, but their influence on the calculated failure 

probabilities is expected to be minimal. 
F-32

Table F.25 Summary of Stress Cycles for Feedwater Line Elbow 

(from Page A.25 of NUREG/CR-6674 [F.5]) 

NAME OF PLANT = GE-NEW 

NAME OF COMPONENT = FEEDWATER LINE ELBOW 

NUM OF LOAD PAIRS = 28 

MATERIAL = LAS 

WALL THICK (INCH) = 1.000 

INNER DIAMETER = 12.000 

AIR/WATER = WATER 

TEMPERATURE(F) = 590.000 

SULFUR(WHT%) = .015 

DISOL O2 (PPM) = .100 

STR RATE (%/SEC) = 0.00100 

USEAGE(DETERM.) = 3.68800 

P-INITIATION@40 = 1.59E-01 

P-INITIATION@60 = 3.65E-01 

 P-TWC @40 = 1.01E-03 

 P-TWC @60 = 1.46E-02 

LOAD PAIR AMP(KSI) NUM/40 YR EDOT(%/S) USEAGE TEMP, °C

HIGH 18/LOW 21 106.040 5.0 .117000 .025000 200

HIGH 18/LOW 21 103.960 5.0 .114000 .024000 200

HIGH 18/LOW 21 102.610 5.0 .113000 .024000 200

HIGH 14/LOW 17 91.590 8.0 .001000 .123000 200

HIGH 8/LOW 17 89.400 10.0 .095000 .037000 200

HIGH 3/LOW 16 88.270 5.0 .094000 .018000 200

HIGH 8/HIGH 7 83.760 126.0 .041000 .519000 200

HIGH 7/HIGH 7 81.430 10.0 .086000 .033000 215

HIGH 7/LOW 13 67.930 97.0 .001000 .740000 200

HIGH 7/LOW 13 66.710 14.0 .001000 .101000 200

HIGH 7/LOW 15 61.290 6.0 .001000 .035000 200

HIGH 7/LOW 15 61.160 64.0 .001000 .451000 212

HIGH 8/LOW 12 55.500 92.0 .001000 .391000 200

HIGH 3/LOW 12 46.630 88.0 .001000 .254000 215

HIGH 7/LOW 22 42.880 15.0 .001000 .029000 212

HIGH 3/HIGH 7 39.440 212.0 .001000 .315000 215

HIGH 3/HIGH 7 38.130 69.0 .001000 .104000 224

HIGH 3/LOW 20 36.800 11.0 .001000 .014000 224

HIGH 4/LOW 20 34.320 60.0 .001000 .053000 215

LOW 11/LOW 20 32.950 203.0 .001000 .122000 200

HIGH 7/LOW 11 32.530 360.0 .001000 .203000 200

HIGH 6/LOW 11 29.770 222.0 .025000 .035000 200

HIGH 2/HIGH 19 26.090 30.0 .028000 .003000 212

HIGH 5/HIGH 19 26.040 81.0 .028000 .007000 200

HIGH 5/HIGH 9 21.640 96.0 .001000 .012000 212

HIGH 1/HIGH 11 20.560 40.0 .001000 .003000 200

LOW 10/LOW 11 14.180 30.0 .001000 .001000 200

HIGH 5/LOW 11 11.220 11515.0 .001000 .008000 200

F.3.7.3 Results - PRAISE runs for this component were made using the version that can treat fatigue 

crack initiation with details of the circumferential variation of the stresses. The feedwater system is 
F-33

relatively more likely to experience water hammer, so the influence of an overload event with a stress of 

0.42σflo = 128 MPa (18.5 ksi) above that normally present was considered. This stress is denoted as σDL, 

and results were generated for one cycle of this stress at 24, 39, or 59 years. The results are summarized 

in Table F.26, which includes the effects of σDL (columns D & F). 

Table F.26 Cumulative PRAISE Results for Feedwater Line Elbow 

 

A B C D E F G 

Stresses Ref. 

F.5 

Table F.25 Table F.25 Table F.25 Table F.25 Table F.25 80% of 

Table F.25 

Failure 

Criterion 

σflow σflow σflow σflow σflow & J-T σflow & J-T σflow 

σDL no no no σDL@(t-1) no σDL@(t-1) no 

25 -- -- <10-8 2.5x10-8 1.0x10-7 3.10x10-6 <10-7 

40 0.001 2x10-6 5.69x10-6 7.19x10-6 1.54x10-5 1.43x10-4 <10-7 

60 0.0146 1.8x10-4 2.57x10-4 2.59x10-4 ~5x10-4 2.9x10-3 4.6x10-7 

>0 

 Ref F.6 

Table 4-8 

108 trials GEN6TWA4 

25 -- -- <10-8 1.5x10-6 * <10-7 1.70x10-6*

40 -- -- <10-8 1.5x10-6 * <10-7 1.70x10-6*

60 -- -- <10-8 1.50x10-6* -- 2.1x10-6 * 

>100 

 GENC6TW4 

25 -- -- <10-7

40 -- -- <10-7

60 -- -- <10-7

>1500 

 

 axisymmetric 

actual T 

 reduced 

stresses 

* also a break

Case A is directly from Reference F.5, and Case B is directly from Table 4-8 of Reference F.6. Case C is 

Case B rerun with 108 trials. Cases D-G are variations of C with different failure criteria, overloads and 

reduction of stresses. The results for various failure criteria (critical net section stress only or critical net 

section stress and tearing instability) show that consideration of tearing instability noticeably increases the 

computed failure probability (compare, for instance, cases C&E). Consideration of an overload event 

also has a noticeable effect (E&F). The use of lower stresses markedly reduces the computed failure 

probabilities (G & C). In the case of an overload event, the probability of a 100 gpm failure is the same 

as a complete pipe break. 

F.3.7.4 Alternate Procedure - The results of Table F.26 show that the probability of a large leak was 

obtainable from the Monte Carlo procedure only when a large overload occurred. When this did not 

occur, there were no leaks of even 380 lpm (100 gpm) in 107

 or 108 trials. In order to obtain estimates for 

the larger leak probabilities, the alternate procedure discussed for the surge line was also applied to Case 

C of Table F.26 for the feedwater elbow. 

As before, the crack length for a given leak rate, b( q& ), was obtained from a pcPRAISE run, along with 

the half-crack length of any cracks that become through-wall. Figure F.11 provides a plot of the leak rate 

as a function of b for the feedwater elbow. 
F-34

0 2 4 6 8 10 12

half crack length, b, inches

0

2000

4000

6000

8000

10000

12000

leak rate, gpm

Figure F.11 Leak Rate as a Function of Half Crack Length for Feedwater Elbow Base Case C 

The results in Table F.27 are obtained from this figure and the corresponding pcPRAISE results. This 

table also includes the portion of the circumference that is cracked and the proportion of the crack 

opening area to the flow area of the pipe. It is seen that the opening area of the crack is nearly equal to 

the flow area of the pipe when the leak rate is 19,000 lpm (5,000 gpm). The value of b for a complete 

pipe break, as obtained from Equation E.7 is also included. Table F.29 defines b( q& ). 

Table F.27 Half Crack Lengths and Areas for a Given Leak Rate 

(Feedwater Elbow Base Case C) 

q& , 

gpm 

b, 

inches 

b

πRI 

A, 

in2 

A

Apipe 

100 5.737 0.32 1.837 0.02 

1500 9.743 0.55 27.554 0.27 

5000 11.095 0.62 90.877 0.91 

DEGB 15.925 0.89 -- -- 

As before, the next step is to estimate the probability of having a through-wall crack exceeding a given 

length as a function of time. The modified version of pcPRAISE was used to generate a table of values of 

b and the time at which the leak first occurred. A run was made with 107 trials, with 2,607 cracks 

becoming through-wall within 60 years. This corresponds to a leak probability of 2.607x10-4 at 60 years, 

which agrees closely with the leak probability obtained earlier. Of these 2,607 cracks, none appeared

before 25 years, and 64 occurred between 25 and 40 years. The statistical distribution of these 64 cracks 

at 40 years provides the probability of having a through-wall crack greater than a given length within 40 

years. Extrapolation is required to obtain results for the crack lengths included in Table F.27. Figure 

F.12 shows the complementary cumulative distribution of b at 40 years, along with the curve fit of 

Equation F.10. 

34.5 ( )1

( )

− −

> =

b P b e (40 years) [F.10] 
F-35

Note that the plot starts at a half-crack length of 25 mm (1 inch), and that the data are closely 

approximated by a straight line on log-linear scales. 

1.0 1.2 1.4 1.6 1.8 2.0

half crack length, inches

10-3

10-2

10-1

100

probability of a crack longer than b

Figure F.12 Complementary Cumulative Distribution of Half-Crack Length of Through-Wall 

Cracks in Feedwater Elbow within 40 Years, Along with Fit 

Figure F.13 provides a similar plot for the 2,607 through-wall cracks that occurred within 60 years. 

Equation F.11 is the fit of the distribution at 60 years within the range of interest. 

25.2 ( )1

( ) .0 0274 − −> =

b P b e (60 years) [F.11] 

Note that in this case the data appear bilinear and are not well approximated by a straight line on loglinear scales. To represent the data at the longer crack lengths of interest, a straight line was assumed 

beyond a crack length of 50 mm (2 inches). This corresponds to a probability below about 0.003.
F-36

1.0 1.5 2.0 2.5 3.0 3.5

half crack length, inches

10-4

10-3

10-2

10-1

100

probability of having a crack longer than b

Figure F.13 Complementary Cumulative Distribution of Half-Crack Length of Through-Wall 

Cracks in Feedwater Elbow within 60 Years, Along with Fit 

The probability of a leak exceeding a given size within 40 and 60 years is then obtained by taking using 

the value of b for a given leak rate from Table F.27 in conjunction with Equations F.10 and F.11, 

respectively. Table F.28 summarizes the results. 

Table F.28 Cumulative Results for Feedwater Elbow Case C 

time 

years 

P(> q& ) 

25 <10-8 

40 5.69x10-6 

60 2.57x10-4 

>0 

 

25 --

40 1.03x10-11 

60 6.44x10-7 * 

>100 

 

25 --

40 5.29x10-21 

60 7.84x10-11 

>1500 

 

25 --

40 3.88x10-24 

>5k 

60 3.74x10-12

25 --

40 2.44x10-35 

DEPB 

60 7.14x10-17

* direct Monte Carlo gave <10-8
F-37

The leak (>0) results in Table F.28 came directly from the Monte Carlo simulation. With 108 trials, no 

leaks exceeding 380 lpm (100 gpm) were obtained. Hence, the Monte Carlo simulation predicts <10-8 

probability of a leak exceeding 380 lpm (100 gpm) within 60 years. The alternative procedure gave a 

corresponding value of 6.44x10-7. This suggests that the alternative procedure overestimates the 

probability of a given leak, as was also the case for the surge line elbow. 

F.4 Selection of Reference Cases and Extension to System Frequencies 

The earlier sections of this document contain many sets of results for each base case component. The 

multiple cases were generated primarily as a series of sensitivity studies. For these results are to be useful 

in the LOCA elicitation, a reference case must be selected for each component as being representative for 

that component. This section briefly discusses which case for each component is suggested as the 

reference case, and system leak frequencies are presented for each reference case. 

The joint frequency is calculated from the cumulative results reported above by use of Equation F.2. The 

system frequencies are then obtained by multiplying by the number of highly stressed joints in the system 

(this approximation works because the failure probabilities are generally small). 

Each component is discussed, with a summary table provided after all components are discussed. 

F.4.1 Hot Leg Pressure Vessel 

As shown in Tables F.5 – F.7, the large leak (> 380 lpm [100 gpm] and larger) probabilities for this 

component varied considerably, depending on the crack growth mechanism (cycle-dependent fatigue or 

time-dependent stress corrosion cracking [PWSCC]), and whether crack initiation or growth from preexisting defects was considered. The fatigue crack growth results (Table F.5) were very low (~10-18), and 

the PWSCC crack initiation results (Table F.6) were quite large (~10-5). Since it is expected that this 

component will totally dominate the very large (> 380,000 lpm [100,000 gpm]) leak category, the 

selection of the reference case is critical for very large leak estimates. The PWSCC with fabrication

defects has intermediate failure probability results (~10-10), and is recommended as the reference case. 

The case without residual stresses is selected. Table F.6 shows that residual stresses do not have a large 

influence. The time dependency of the large leak cumulative probability is very small, which suggests

that the leak frequency is very small. For estimation purposes, the leak frequencies are estimated by

taking the value of the cumulative at 60 years, dividing it by 60, and assuming the value to be applicable 

independently of time. This will overestimate the leak frequency at long time and underestimate it at short 

time. 

For extension to system failure frequency, it is assumed that there are three comparably stressed joints in 

the large main coolant piping. 

F.4.2 Surge Line Elbow 

The surge line elbow result identified as “axisymmetric nonseismic” in Table F.12 is suggested as the 

reference case. Table F.13 summarizes the cumulative results for the larger flow rates, which were 

obtained by the alternative procedure. 

Two of these highly stresses elbows are considered to be present in the surge line system 

F.4.3 HPI Makeup Nozzle 

Probability analyses were performed with and without failure of the thermal sleeve, which has been 

observed to fail in service. The least favorable large leak probabilities were for a failed thermal sleeve, 
F-38

which immediately resulting in fatigue crack initiation, but with the same stresses as before. This is 

suggested as the reference case, with the column labeled σu = 0 in Table F.16 being the results of interest. 

Three such locations are considered to be present in the system. 

F.4.4 Recirculation Line – 12 inch 

Analyses were performed for this component for a range of applied stresses, with predictions compared to 

field experience of leaks and observed part-through cracks. Analyses were performed for no remedial 

action, and for a weld overlay at 20 years. The weld overlay at 20 years is considered to be the most

realistic. Comparisons with experience led to an estimate of stresses that were considerably below the

peak value used in the original analysis. However, when compensated for the number of weld joints 

involved, the system leak frequencies were nearly the same whether 49 joints with a random stress (mean 

σNO = 83 MPa [12 ksi]) or 2 joints with a high stress (σNO = 140 MPa [20 ksi]) were considered (see Table 

F.25). The case of weld overlay at 20 years with the high stress is recommended as the reference case. 

Table F.19 contains the cumulative results. 

Two of the highly stresses joints are considered to be present in the recirculation system. 

F.4.5 Recirculation Line – 28 inch 

The recirculation line with no remedial action and a high stress representing the dominant joints was the 

only case considered, and is summarized in Table F.24. 

Two such joints are considered to be present in the system. 

F.4.6 Feedwater Elbow 

Case C in Table F.26 is suggested as the reference case. Results for > 380 lpm (100gpm) and larger were 

generated by the alternative procedure, and are summarized in Table F.28. 

Four such locations were considered to be present in the system. 

F.4.7 Summary Table 

Table F.29 provides an overall summary of the leak flow rate frequencies for the reference cases of the 

base case systems.
F-39

Table F.29 Summary of Results for Reference Systems 

 Recirculation 

 

Hot Leg Surge 

Line 

HPI 

12 28 

Feedwater 

OD, in 34 14 3.44 12.75 28 12.75

t, in 2.5 1.406 0.4375 0.687 1.201 0.687

A, in2 661 98.3 5.167 102 515 102

Qmax 423 63 3.6 38 193 38

matl cast SS SS SS SS SS CS

Degr 

Mech 

PWSCC 

growth 

fatigue 

init&gro

fatigue SCC 

init&gro 

SCC 

init&gro 

fatigue 

init&gro 

Table F.6 F.12 F.16 F.21 F.26 F.26, F.28

Case PWSCC 

no σres 

Table 

F.9 

stresses

failed 

slv σu=0 

overlay 

@ 20 yrs 

 C 

Insp 0,20,40 none none 0,20,40 0,20,40 none 

0-25 -- 9.3x10-3 1.48x10-4 1.19x10-2 2.5x10-4 <4x10-10 

25-40 -- 0.024 5.94x10-4 5.57x10-3 2.6x10-4 3.8x10-7

>0 

40-60 -- 0.015 8.60x10-4 2.19x10-3 2.2x10-4 1.3x10-5

0-25 1.33x10-8 3.0x10-7 2.60x10-5 5.71x10-3 2.4x10-5 --

25-40 1.33x10-8 4.2x10-6 1.35x10-4 1.30x10-3 1.3x10-5 6.9x10-13

>0.1 

40-60 1.33x10-8 9.0x10-6 1.32x10-4 3.55x10-4 <5x10-6 3.2x10-8

0-25 1.6x10-11 1.4x10-9 2.60x10-5 4.26x10-3 2.7x10-6 --

25-40 1.6x10-11 2.2x10-8 1.35x10-4 1.23x10-3 -- 3.5x10-22

>1.5 

40-60 1.6x10-11 5.8x10-8 1.32x10-4 3.10x10-4 3.0x10-6 3.9x10-12

0-25 4.6x10-13 9.7x10-11 3x10-3 2.4x10-6 --

25-40 4.6x10-13 2.3x10-9 1.23x10-3 5.8x10-7 2.6x10-25

>5 

40-60 4.6x10-13 8.3x10-9 3.10x10-4 1.5x10-6 1.9x10-13

0-25 4.6x10-13 3.9x10-14 1.96x10-3 1.3x10-6 --

25-40 4.6x10-13 3.2x10-12 1.23x10-3 ~2x10-6 1.6x10-36 

>25 

40-60 4.6x10-13 2.5x10-11 3.10x10-4 1.7x10-6 3.6x10-18 

0-25 3.6x10-16 1.6x10-6

25-40 3.6x10-16 ~2x10-6

dominant joint freq 

>100

**

40-60 3.6x10-16 1.7x10-6

 field 22 3 20 22 29 

shop 12 9 20 30 22 

safe end 16 1 9 3 12 

dominant 3 2 3 2 2 4 

0-25 -- 0.019 4.44x10-4 2.43x10-2 <1.6x10-9 

25-40 -- 0.048 1.78x10-3 1.17x10-2 1.5x10-6

>0 

40-60 -- 0.030 2.58x10-3 4.82x10-3 5.2x10-5

0-25 4.0x10-8 6.0x10-7 7.80x10-5 1.15x10-2 --

25-40 4.0x10-8 8.5x10-6 4.05x10-4 2.62x10-3 2.8x10-12

>0.1 

40-60 4.0x10-8 1.8x10-5 3.96x10-4 7.10x10-4 1.3x10-7

0-25 4.8x10-11 2.8x10-9 7.80x10-5 8.52x10-3 --

25-40 4.8x10-11 4.4x10-8 4.05x10-4 2.46x10-3 1.4x10-21

>1.5 

40-60 4.8x10-11 1.2x10-7 3.96x10-4 6.20x10-4 1.6x10-11

0-25 1.4x10-12 1.9x10-10 6x10-3 --

25-40 1.4x10-12 4.6x10-9 2.46x10-3 1.0x10-24

>5 

40-60 1.4x10-12 1.7x10-8 6.20x10-4 7.6x10-13

0-25 1.4x10-12 7.9x10-14 3.92x10-3 --

25-40 1.4x10-12 6.4x10-12 2.46x10-3 6.5x10-36 

>25 

40-60 1.4x10-12 5.0x10-11 6.20x10-4 1.4x10-17 

0-25 1.1x10-15 2.6x10-6 

25-40 1.1x10-15 4x10-6 

system frequencies 

>100

**

40-60 1.1x10-15 3.7x10-6 

times in reactor years, 1 calendar year ~ 0.8 reactor years 

shaded areas are estimates based on alternative procedure 

leak rates in thousands of gallons per minute 

cross-hatched cells are beyond maximum leak capability for that pipe size 

** also applicable to > 1,900,000 lpm (500 kgpm) for hot leg if sufficient diameter 
F-40

F.5 References

F.1. D. O. Harris, E. Y. Lim and D. Dedhia, Probability of Pipe Fracture in the Primary Coolant

Loop of a PWR Plant, Vol. 5: Probabilistic Fracture Mechanics Analysis, U.S. Nuclear 

Regulatory Commission Report NUREG/CR-2189, Vol. 5, Washington, D.C., August 1981 

F.2. D. O. Harris, D. Dedhia, E.D. Eason and S.D. Patterson, Probability of Failure in BWR 

Reactor Coolant Piping: Probabilistic Treatment of Stress Corrosion Cracking in 304 and 

316NG BWR Piping Weldments, U.S. Nuclear Regulatory Commission Report NUREG/CR4792, Vol. 3, Washington, D.C., December 1986 

F.3. D. O. Harris, D. Dedhia and S. C. Lu, Theoretical and User’s Manual for pc-PRAISE, A 

Probabilistic Fracture Mechanics Code for Piping Reliability Analysis, U.S. Nuclear Regulatory 

Commission Report NUREG/CR-5864, Washington, D.C., July 1992 

F.4. D.O. Harris and D. Dedhia, WinPRAISE: PRAISE Code in Windows, Engineering 

Mechanics Technology, Inc. San Jose, California, Technical Report TR-98-4-1, 1998 

F.5. M.A. Khaleel, F.A. Simonen, H.K. Phan, D.O. Harris and D. Dedhia, Fatigue Analysis of 

Components for 60-Year Plant Life, U.S. Nuclear Regulatory Commission Report NUREG/CR6674, Washington, D.C., June 2000 

F.6. A. Deardorff, D. Harris and D. Dedhia, Materials Reliability Program: Re-Evaluation of 

Results in NUREG/CR-6774 for Carbon and Low-Alloy Steel Components”, Electric Power 

Research Institute Report 1003667, Palo Alto, California, 2002 

F.7. J. Keisler, O.K. Chopra and W.J. Shack, Fatigue Strain-Life behavior of Carbon, Low-Alloy 

Steels, Austenitic Stainless Steels, and Alloy 600 in LWR Environments, U.S. Nuclear Regulatory 

Commission Report NUREG/CR-6335, Washington, D.C., 1995 

F.8. O.K. Chopra and W.J. Shack, Effects of LWR Coolant Environments on Fatigue Design 

Curves of Carbon and Low-Alloy Steels, U.S. Nuclear Regulatory Commission Report 

NUREG/CR-6583, Washington, D.C., March 1998 

F.9. Technical Elements of Risk-Informed Inservice Inspection Programs for Piping, U.S. Nuclear 

Regulatory Commission Draft Report NUREG-1661, Washington, D.C., January 1999 

F.10. M.A. Khaleel, O.J.V. Chapman, D.O. Harris and F.A. Simonen, “Flaw Size Distribution 

and Flaw Existence Frequencies in Nuclear Piping”, Probabilistic and Environmental Aspects of 

Fracture and Fatigue, ASME PVP-Vol. 386, 1999, pp. 127-144 

F.11. ASME Boiler and Pressure Vessel Code, Section XI, Appendix C, 1992 

F.12. P. Ricardella, “Probabilistic Fracture Mechanics Analysis of CRDM Nozzles”, presented at 

ACRS Meeting, Rockville, Maryland, June 5, 2002 

F.13. e-mail from Gery Wilkowski to David Harris, “Material Property Inputs for Base Cases”, 

June 10, 2003 

F.14. Personal communication, Art Deardorff, Structural Integrity Associates, San Jose, 

California, to David Harris, Engineering Mechanics Technology, Inc., San Jose, California 
F-41

F.15. T.C. Chapman, et al., Assessment of Remedies for Degraded Piping, Electric Power 

Research Institute Report NP-5881-LD, Palo Alto, California, 1988 

F.16. B.O.Y. Lydell, An Application of the Parametric Attribute/Influence Methodology to 

Determine Loss of Coolant Accident (LOCA) Frequency Distributions, Document No. R2003-02, 

May 2003, provided to members of the NRC LOCA Frequency Expert Elicitation Panel. 

F.17. Attachment (Action Item 45R1.xls) to e-mail from Bengt Lydell to base case panel 

members, June 20, 2003 

F.18. A.G. Ware, D.K. Morton and M.E. Nitzel, Application of NUREG/CR-5999 Interim Fatigue 

Curves to Selected Nuclear Power Plant Components, U.S. Nuclear Regulatory Commission 

Report NUREG/CR-6260, Washington, D.C., 1995 

 

APPENDIX G 

PIPING BASE CASE RESULTS OF 

VIC CHAPMAN 

G-1

APPENDIX G 

PIPING BASE CASE RESULTS OF VIC CHAPMAN 

Summary of Benchmarking Analysis 

Carried out Using ‘RR-PRODIGAL’ 

G.1 General Background to RR-PRODIGAL

RR PRODIGAL is a basic fatigue failure probability model developed by Rolls Royce for the Naval 

Nuclear program. When analysing a weld, it first simulates the weld construction in order to 

determine a start of life defect distribution and density for both buried and surface breaking defects. 

A failure probability using standard linear elastic fracture mechanics methods is then evaluated for 

both the buried and surface breaking defects (assumptions about break through of buried defects to 

surface defects are based on the ASME criteria). Failure is achieved when the defect either exceeds 

the R6 failure criteria or simply grows through to the full thickness. The failure probability for all 

initial defects is then combined to form the total failure probability. 

For non-weld areas, a probabilistic crack initiation analysis is carried out with a correlated crack 

growth analysis to failure. This correlation means that short times to crack initiation imply that a fast 

crack growth follows this initiation. There is no positive data to confirm or deny this proposition. It 

was chosen simply because it is pessimistic. 

The modelling contains a routine to assess the growth of the defect around a welded pipe at the same 

time as the defect grows through the weld thickness, however, this part was not used in this 

assessment. RR-PRODIGAL does not, at present, contain a routine to evaluate the crack growth of a 

through wall defect around the outer surface of a pipe weld. 

At failure, the model evaluates a critical through wall defect size based again on the R6 criteria. 

At present RR-PRODIGAL does not contain a verified and validated assessment of the PWSCC 

degradation mechanism. 

There are several publications that describe RR-PRODIGAL, which include a recent benchmarking 

exercise as part of a European initiative. References G.1 and G.2 should provide sufficient 

information for any readers wishing to obtain further information on this code. 

G.2 Leak Rate Evaluation

When estimating RR-PRODIGAL leak rates through the final through wall defect in a pipe weld, 

evaluations were made using an elastic crack opening displacement (COD) analysis. However, it was 

felt that the uncertainties associated with assessing both the defect length around the pipe 

circumference as well as the COD needed for estimating the flow rate through the crack, were too 

great and too subject to ongoing development, to allow a suitable analysis of the leak rate. Thus, RRPRODIGAL does not contain, within itself, a routine for evaluating the flow rate from the final defect

size. 

Instead, it was concluded that the leak rate from a through wall defect could be considered 

independently of the probability of the breach, i.e. the leak rate from the defect is not dependent on

the probability of the defect cracking through the pipe wall. Note, however, that the COD, crack 

length, and hence leak rate is not independent of the mechanism that led to the failure, only the 

probability of the failure itself. 
G-2

Within the Naval Nuclear program, computer programs have been developed to assess the leak rate 

from different defects based primarily on the ‘SQUIRT’ model. However, for consistency within this 

program, the data on leak rate against defect area provided by the USNRC were used, as shown in 

Figure G.1. 

G.3 Procedure

The procedures used to develop the base case numbers are as follows: 

1 Evaluate the basic fatigue failure probability using RR-PRODIGAL code using the transient 

data supplied1

. 

2 Evaluate an elastic COD as a function of defect size. 

3 Use expert judgement to extend this COD beyond the elastic limit. 

4 Evaluate a mean defect cross-sectional area for a given defect size using its associated COD. 

5 Evaluate the mean leak rate from a given defect size using the data supplied by the USNRC, 

see Figure G.1. 

[Note for Steps 2, 3, 4 and 5 above a defect length is given. Thus, Steps 2, 3, 4 and 5 provide a 

mapping from a given defect size at failure to the mean leak rate in gpm, given this defect exists.] 

6 Use expert judgement to assess the distribution of the defect length at failure. 

7 Combine Steps 5 and 6 to obtain the conditional probability of a leak rate greater than the 

given leak rates for Categories 1 through 6. These categories being as follows; 

Table G.1 Leak Category Leak Rates 

 Leak Rate Greater than 

(gpm) 

Log 

Leak Rate 

Leak Category 1 100 2 

Leak Category 2 1,500 3.2 

Leak Category 3 5,000 3.7 

Leak Category 4 25,000 4.4 

Leak Category 5 100,000 5.0 

Leak Category 6 500,000 5.7 

8 Combining the conditional probability of Step 7 with the basic fatigue failure probability in 

Step 1 gives the required final probability of a leak greater than each of the categories. 

G.4 Example Base Case Analysis

As a way of demonstrating the procedure given above, the results for the 14-inch Surge Line elbow 

are reproduced in this section. Two situations are considered, the elbow and the adjacent weld. The 

transients were based on data supplied and are reproduced in Attachment G.1 

G.4.1 Probability of Failure Surge Line Elbow – Base Case 

This is a failure from base material and so the analysis assumed a fatigue based crack initiation 

followed by crack growth to failure. As stated earlier, the crack initiation and crack growth are 

assumed to be positively correlated. This assumption assumes that if the properties of the base 

material are such as to lead to an early crack initiation, it is very possible that these same properties 

 

1

This information needed to be supplied because the transient experience for the Naval Nuclear program is a) confidential 

and b) not applicable to commercial plants.
G-3

could result in a subsequently fast crack growth rate. The results of this analysis are shown in the 

following table: 

Table G.2 Results for PWR Surge Line Elbow Base Case Analysis 

Time 

(years) 

Cumulative Probability 

of Failure 

25 6.1x10-6

40 7.8x10-6

60 9.4x10-6

RR-PRODIGAL gave the critical through wall defect length, based on the R6 criterion, as 14 inches. 

G.4.2 Probability of Failure Surge Line Weld 

The surge line elbow weld was analysed at a 60-year life assuming the same cyclic conditions as for 

the elbow itself, but with the stresses factored down by 20 percent as suggested at the Elicitation Base 

Case Review Meeting on June 4 and 5, 2003 in Bethesda, Maryland. The two hydro cases were, 

however, maintained at their original values. 

In this analysis, RR-PRODIGAL first simulates the weld construction, including any build 

inspections, to establish the start of life defect density and distribution for both buried and surface 

breaking defects. As stated earlier, conditional failure probabilities are assessed for both situations 

and combined to give the final failure probability.

The failure probability evaluated for this case was: 

Table G.3 Results for PWR Surge Line Weld Analysis

Cumulative Probability 

of Failure at 60 years 1.3 x 10-4

It can be seen that this failure probability is over an order of magnitude higher than the base case. 

This is due to the difference between having to initiate a defect and then grow this defect to failure, 

and having the probability of pre-existing defects in the weld. The base case values from the base 

material failure as reflected in Table G.2, i.e. crack initiation leading to failure, have been used in 

Table D.1 in the main body of this report. Note, however, that the values reported in Table G.2 are 

cumulative probabilities of failure in 25, 40, and 60 years whereas the values reported in Table D.1 of 

the main body are frequencies. Consequently, the Table G.2 values need to be divided by 25, 40, and 

60 years, respectively to facilitate any comparisons. Furthermore, the values in Table D.1 are for leak 

rates greater than the threshold leak rates, i.e., 380 lpm (100 gpm) while the values in Table G.2 

reflect the totals. 

G.4.3 COD and Leak Rate for a Given Defect Size 

Having established a basic failure probability, the COD can be evaluated independently of this 

probability. Once this is established, the leakage area of the defect follows, and given this leak area, 

the flow rate can be evaluated using the information from Figure G.1. A mean power law was then 

used to calculate the mean flow rate given a leakage area. The table below gives the elastic COD 

values evaluated for this case and the resultant flow rate. 
G-4

Table G.4 Elastic COD and Resultant Leak Rates for a Given Defect Length 

Defect Length 

(inches) 

Elastic COD 

(inches) 

Flow rate 

(gpm) 

1.98 0.0025 17 

3.96 0.0049 48 

5.94 0.0074 92 

7.41 0.01 145 

9.89 0.012 200 

11.87 0.015 

(Invalid Result) 

270 

Interpolating between the results in Table G.4, it can be seen that a defect approximately 160 mm (6.2

inches) long, which is approximately 15 percent of the pipe circumference, results in the first leakage 

category of 380 lpm (100 gpm). 

Clearly it is the behaviour of the defect beyond the elastic range that is of interest for the larger leak 

categories. If it were to be assumed that at the critical defect size the pipe would simply tear, in an 

unstable manner, to result in a Double Ended Guillotine Break (DEGB) failure, then the leak rate 

would simply jump from a Category 1 failure to the gpm associated with the DEGB. In this case that 

would be 250,000 lpm (65,000 gpm) or a Category 4 leak. The probability of a Category 2 leak rate 

would then be the same as a Category 3, which would be the same as the Category 4! 

Such an assumption could be considered valid. However, in this work, it was assumed that the defect 

would continue opening in a stable, but plastic manner. Whilst models do exist to evaluate the plastic 

deformation of defective pipes, no such model was used in this analysis. Instead expert judgement 

was used to assess how the COD would develop beyond this elastic point, and at what defect size the 

pipe would finally tear into a DEGB failure. The results of this judgement are shown in Figure G.2. 

The area of leakage can then be calculated, and the leak rate, given a defect length also follows. The 

resulting gallon per minute flow rate, for this example, is shown in Figure G.3. 

The failure probability gives the basic probability of a breach of the pressure boundary. Figure G.3 

shows the leak rate in gallons per minute, given a defect of a given length. In order to obtain the 

probability of a leak rate greater than ‘X’ gallons per minute, it only remains to provide a distribution 

of the defect size at the moment of failure. 

G.4.4 Defect Distribution and Leak Rate at Failure – No Leak Detection 

First consider the case with no leak detection. For this case the instantaneous size of the defect, and 

its associated COD, at the moment of snap through to a breach of containment is required. As an 

example, if the aspect ratio were of the order of 8/1 at snap through, then given a pipe wall thickness 

of about 36 mm (1.4 inches), the defect length would be approximately ten or eleven inches long. If it 

were then pessimistically assumed that this was the full through wall defect length, then the 

instantaneous leak rate would be just above (actually about twice) our ‘Category 1’ failure criteria of 

380 lpm (100 gpm). Thus, the probability of a leak rate greater than Category 1 becomes the basic 

probability of failure times the probability that the defect at snap through was greater than 250 mm 

(10 inches), i.e., the defect had an aspect ratio at snap through of about 8/1 or greater. It then follows, 

from Figure G.3, that in order to exceed the Category 2 leak rate, the instantaneous defect size at snap 

through would have to be greater than 380 or 405 mm (15 or 16 inches), i.e., the defect had an aspect 

ration of about 11/1. Furthermore, the defect snapped straight open to the fully plastic COD. 

As stated earlier, RR-PRODIGAL has the capability of simulating the crack growth both around and 

through the pipe wall. However, this is not generally used as the solutions require a detailed 

knowledge of the stress distribution around the pipe, including any weld residual stress, and generally 

such knowledge is not well enough defined. Thus, expert judgement was again used. The expert 
G-5

judgement required is to generate a defect distribution at the moment the defect snaps to the COD of 

Figure G.3, assuming no leak detection. 

This base case is for the surge line elbow and it has been assumed that most of the deformation and 

high stress will result from large bending moments at the elbow. It was felt that this would initiate a 

defect preferential on the hogging side of the elbow, and promote a crack to grow through the wall 

thickness on this side of the elbow. This would then imply that the crack growth around the pipe 

diameter would be restricted. Figure G.4 represents the distribution decided upon for this analysis. 

This distribution shows the most likely defect length to be up to about 250 mm (10 inches), which is 

about a quarter of the way around the pipe circumference. The probability of the defect being over 

halfway around the pipe is seen as a rare event, being about 0.025 or a 1 in 40 chance. If the loading 

were not dominated by bending, then this distribution would probably be judged to be flatter, with 

perhaps a 1 in 10 chance of being greater than halfway round the pipe circumference. 

Combining Figures G.3 and G.4 gives the conditional probability of a leak greater than a given leak 

rate. This final plot is given in Figure G.5 and is combined with the basic failure probability to derive 

the values given in Table D.1 in Section D of the main body of this report. 

G.4.5 Defect Distribution and Leak Rate at Failure – With Leak Detection 

In the previous section it was assumed that the defect would instantaneously snap open to the full 

COD associated with its length at the moment the pressure boundary was breached. In reality this 

will probably not happen. Instead, the very large defects, which are those of interest, will probably

grow to different through wall depths at different points around the length. Thus, much smaller 

surface defects would begin to breach the boundary at different points around the defect. The COD of 

these small defects would then remain elastic until the whole defect progressed to the surface. In this 

scenario, the leak from the defect would start very small and grow, slowly at first and then probably 

very quickly before snapping open to the fully plastic COD. 

During this time of surface crack combination, the leak rate may exceed the value at which the 

operators shut the reactor down to a safe state in order to investigate the leak. Provided this occurs 

before the crack reaches a critical size, i.e., before the leak rate moves very quickly to the final leak 

state. Whilst the high leak rate may still occur, the plant would be in a safe condition. This can be 

seen as leak detection. 

This probability of leak detection is almost certainly associated with the length of the defect that is 

itself related to the rate of leakage in the previous section. Thus, expert judgement was again used to 

introduce a factor, based on the leak rate, which would represent this probability of leak detection. 

Figure G.6 shows this plot as a function of leak rate. 

From this plot it can be seen that the reduction factor for Category 1 (380 lpm [100 gpm]) is about 

five, rising to a factor of about fifty at Category 6 (1,900,000 lpm [500,000 gpm]). 

G.5 Effect of In Service Inspection

An assessment of the effect of ISI was carried out for the surge line elbow weld, the defect 

distribution and density being those generated by RR-PRODIGAL, see section G.4.2. A Probability 

of Detection (POD) curve was defined by the following equation: 

POD 1 2 ln a

f c c

t

    = Φ +        

 

where c1 = 1.526 and c2 = 0.533 (G.1) 

This POD is shown in Figure G.7, and it can be seen that this sets the probability of detection at about 

90 percent for defects 70 percent of the way through the wall thickness. This was felt to be 
G-6

representative of inspections carried out to date, but for future inspections that conform to modern 

standards, this POD could be much better. 

The results are shown in the table below and in Figure G.8 for various ISI intervals. 

Table G.5 Reduction Factors Due to ISI 

ISI case Cumulative Probability 

of Failure at 60 years 

Factor for 

General Use 

No ISI 1.3 x 10-4 1 

0 years (PSI) 4.2 x 10-5 3 

10 years 3.8 x 10-5 3.4 

10, 20 years 1.3 x 10-5 10 

10, 20, 30 years 6.5 x 10-6 20 

10, 20, 30, 40 years 4.8 x 10-6 27 

10, 20, 30, 40, 50 years 4.7 x 10-6 28 

These results suggest that even with this quite low inspection capability, and for a weld with a high 

failure probability, reductions of a decade can be achieved with two or three inspections during the 

life of the plant. It also indicates that going beyond three inspections gives little extra return. 

An interesting conclusion from this figure would be that if a fourth inspection is carried out at the end 

of a forty year period, then, provided this inspection was clear, there would be little gain from an 

inspection at fifty years for a total life of sixty years! However, at this stage such a conclusion can 

only be taken as tentative and would require more investigation. 

G.6 References

G.1 NUREG/CR-5505 PNNL-11898 ‘RR-PRODIGAL – A Model for Estimating the Probability of 

Defects in Reactor Pressure Vessel Welds. 

G.2 NURBIM (Nuclear Risk-Based Inspection Methodology) WP4. Published by the European 

Commission under the EURATOM programme. 
G-7

Figure G.1 Leak Rate as a Function of Leakage Area (Data Supplied by USNRC) 

Figure G.2 Estimated Defect Separation (COD) Based on Expert Judgment 

as a Function of Defect Length Assuming Plastic Deformation 

Leak R ate = A*(area^n)

1.5

2

2.5

3

3.5

4

4.5

5

-1.5 -0.5 0.5 1.5 2.5

Log A rea

Log Leak Rate (gpm)

S ource D ata

M ean

Length Separation 

0

0.5

1

1.5

2

2.5

0 5 10 15 20 25 30

Full Defect Length (Inches)

Defect Separation (Inches)

Onset of

 Plastic Hinge

Elastic COD

DEGB 

Assumed 

at 

28 inches
G-8

Figure G.3 Estimated Leak Rate Versus Defect Length Based on Expert Judgment 

Figure G.4 Probability of the Existence of a Defect of a Certain Length for Surge Line Base 

Case 

 Defect Size - Given Failure (Large BM) 

0 5 10 15 20 25 30 35

Full Defect Length (Inches)

PDF

Most likely length of defect at 

snap through

Rare Events

Defect half 

way round 

pipe

Leak Rate

0.E+00

1.E+04

2.E+04

3.E+04

4.E+04

5.E+04

6.E+04

7.E+04

0 5 10 15 20 25 30

Defect Size (Inches)

Leak Rate (gpm)

DEGB
G-9

Figure G.5 Conditional Probability of a Leak of a Given Size 

Figure G.6 Reduction Factors for Leak Detection Based on Expert Judgment 

 Probability of Leak Greater Than 'X'

(No Leak Detection)

-3.0E+00

-2.5E+00

-2.0E+00

-1.5E+00

-1.0E+00

-5.0E-01

0.0E+00

0 0.5 1 1.5 2 2.5 3 3.5 4 4.5 5

Log Leak Rate 'X' (gpm)

Log Probability

Reduction Factor for Leak Detection

0

10

20

30

40

50

60

0.E+00 1.E+05 2.E+05 3.E+05 4.E+05 5.E+05

Leak Rate (gpm)

Reduction Factor

Figure 6
G-10

Figure G.7 Probability of Detection Curve 

Figure G.8 Reduction Factors Due to ISI as a Function of the Number of Inspections 

Reduction Factor against Number of ISI's

0

5

10

15

20

25

30

0 1 2 3 4 5

Number of ISI's over 60 Years

Reduction Factor

Figure 8

ISI Probability of Detection

0

0.1

0.2

0.3

0.4

0.5

0.6

0.7

0.8

0.9

1

0 0.2 0.4 0.6 0.8 1

a/t

Probability
G-11

ATTACHMENT 1 TO APPENDIX G (ATTACHEMENT G.1) FROM 

NOTE ‘SUMMARY OF BASE CASES STRESSES’ APRIL 2003 

Summary of Stress Cycles for Surge Line Elbow 

 (No seismic stresses) 

Load Pair Amplitude 

(ksi) 

Number/40 

years 

HYDRO-EXTREME 190.17 6 

9B-HYDRO 149.86 4 

8A-UPSET 4 140.42 14 

9B-UPSET4 139.43 10 

8B-UPSET4 105.89 14 

9A-UPSET4 105.13 2 

9A-LEAK 103.86 12 

8F-18 63.40 68 

9C-11 63.38 68 

9F-LEAK 63.37 68 

8C-LEAK 63.37 35 

2A-8C 62.30 33 

8G-18 52.38 22 

8G-17 52.35 90 

9D-11 52.35 22 

2A-8D 51.20 72 

8H-9G 51.18 400 

8G-UPSET3 51.00 30 

9D-12 50.96 50 

8G-12 50.96 40 

8G-16 50.93 90 

8G-9H 50.92 128 

2A-8E 40.10 90 

8H-9H 40.09 100 

9H-10A 40.09 272 

9E-13 39.82 90 

3A-10A 33.10 4120 

6-10A 33.10 200 

3B-10A 33.10 4120 

7-10A 33.10 4580 

2B-SLUG1 32.87 100 

2B-SLUG2 32.87 500 

5-10A 29.90 9400 

4A-10A 29.90 17040 

4B-10A 29.90 17040 

2B-10A 20.60 14400 

2A-10A 20.60 14805 

10A-UPSET1 20.59 70 

10A-UPSET5 20.59 30 

10A-UPSET6 20.59 5 

10A-UPSET2 20.59 95 

1B-10A 20.59 1533 

1B-10B 20.00 87710 
G-12

Many of the high stress contributors in Tables 2 and 3 are from rapid excursions of the coolant 

temperature. The largest stress amplitude (half the peak-to-peak) is 1,310 MPa (190 ksi), so the 

stresses are large (but localised). These are the stresses at the peak stress location, which is not at 

weld. The spatial stress gradients (both along the surface and into the pipe wall) are required for a

thorough analysis. The radial gradient (into the pipe wall) can be estimated by the following 

procedure: 

l Cyclic stresses associated with seismic loads were treated as 

100 percent uniform stress. 

l Cyclic stresses greater than 310 MPa (45 ksi) were treated 

as having a uniform component of 310 MPa (45 ksi), and 

the remainder were assigned to the gradient category. 

l For those transients with more than 1,000 cycles over a 40-

year life, it was assumed that 50% of the stress was uniform 

stress and 50% a through-wall gradient stress. In addition, 

for these transients, the uniform stress component was not 

permitted to exceed 70 MPa (10 ksi). 

 

The gradient stress mentioned above is assumed to vary through the thickness as 













= − +

2

2

3

σ (ξ ) σ o 1 3ξ ξ (G.2) 

In this equation, σo is the stress at the inner wall of the pipe, ξ = x / h , x is the distance into the pipe 

wall from the inner surface, and h is the wall thickness. The stresses and cycles are high enough that 

fatigue crack initiation is important, which has been considered in Reference 6. Reference 6 shows a 

probability of 0.981 of a leak in 40 years for this component. The LOCA probabilities will be less. 

The use of the gradient along the surface will reduce this. 

As mentioned above, the more thorough results that include a better estimate of the radial gradient and 

also consider the spatial gradient along the surface are available, and could be used for the base case 

calculations. 
APPENDIX H 

DESCRIPTION OF NON-PIPING DATABASE 

H-1

APPENDIX H 

DESCRIPTION OF NON-PIPING DATABASE

H.1 Background

The non-piping database has been compiled with the intention that it will serve as one source of 

information supporting the development of estimates of LOCA frequencies attributable to non-pipe 

components. The data has been obtained from two primary sources. First, a search of LERs was made to

identify those instances where failures1

 of non-pipe components of the primary reactor coolant pressure 

boundary were reported to the NRC. The second source of information is data that has been incidentally 

collected on non-pipe components during the development and maintenance of the pipe-based OECD and 

SLAP databases. LER events compose the majority of the events in the database (see Attachment A of 

this appendix for a description of the LER reporting requirements). 

The database is accessible in two formats, Table and Forms. The Table named “Failure Data” lists the 

data in a spreadsheet type of format where each line of the table contains one data record and each 

column contains the various fields that make-up the records. In the Forms format, only one record of the 

“Failure Data” is displayed at a time, but in a manner that allows all of the fields to be view at the same 

time. Both formats display the same data, only the presentation is different. Also, sorting and filtering of 

the data can be done in both views. 

H.2 Approach

A search of LERs was performed (see Attachment B for the specific search criteria) using the SCSS. This 

search returned 1,036 LERs. Each LER was reviewed and coded in the Non-Pipe database. The database 

structure is based on information generated during the elicitation meetings. In particular, the component, 

piece part, and degradation mechanism are all identified using the tables documented in the elicitation 

meeting notes. Other fields of the database were developed and defined as judged appropriate. 

The initial screening of the 1,036 LERs to remove those that were judged to not be applicable reduced the 

total number to 213. As discussed in Attachment B, the data search simply looked for leak and crack 

events associated with primary coolant systems. This conservative search included LERs that identified 

potential and possible leak and crack events (e.g., a problem with ECCS such that the plant would not 

respond as designed to a loss of coolant accident. Screening out the potential failures resulted in a

reduction to 213 records. A further 34 records were removed since they identified problems with pipes or 

seals. Then 37 records were added that had been collected previously during the development of the 

OECD and SLAP databases. This results in a current total of 216 records. 

H.3 Description of Data

This process results in 216 data records that document crack (both partial and full) and leak events 

associated with non-pipe primary coolant system components. This dataset can be considered complete 

for U.S. NPP operation from 1990 through 2002, inclusive, in as much as the LER reporting requirements

(Attachment A) can be relied upon to generate complete reporting. Additionally, the dataset does include 

a limited amount of data from outside this time frame and from non-U.S. plants. Nevertheless, the dataset 

can be considered to be internally consistent, that is, the various components, failures and degradation 

 

1

 Failures are classified using four categories: partial through-wall cracks, through-wall cracks without a significant leak rate 

(typically indicated by a boric acid deposit), leaks, and joint failures (i.e., non-welded connection). 
H-2

mechanisms are believed to be represented equally such that relative ratios (if not the absolute 

frequencies) can be assumed to be reasonably accurate. Several of the database records represent multiple 

cases of degradation or failure. Attachment C includes a sample of multiple event records, including a

discussion on how to estimate flaw frequencies from the observed events as recorded in the database. 

The figure below (Forms view of the database) identifies the various fields maintained by the database. 

For additional detail on those records based on LERs, the LER hyperlink can be clicked to retrieve the

full LER (internet access is require for this). 
H-3

H.4 Limitations

As with many reliability databases, the completeness of the data is always an issue. While relative 

frequencies (e.g., percentage distribution of events by component or degradation mechanism) might be 

reasonably accurate, the accuracy of any absolute frequency (e.g., events per year) calculations will 

depend directly on how complete the data are. That is, have all events that have occurred been included 

in the database? In the current situation this question has two parts. First, have all relevant event been 

discovered? Second, of the discovered events, have they been reported (via LER)? 

The completeness issue is probably more of an issue for the partial through-wall cracks than it is for the 

more severe failures. There are two causes for this concern. One is ambiguity in the interpretation of the 

LER reporting requirement (Attachment A), and the second and probably primary cause is simple lack of 

detection. While effort is made to make the LER reporting requirements as clear as possible, the 

“seriously degraded” aspect of 50.73(a)(2)(ii)(A) is difficult to quantify. How far does a crack have to 

extend to seriously degrade the primary pressure boundary? It is possible that some cracks are being 

detected and repaired (which might be considered normal plant maintenance rather than corrective 

action), without being reported as a LER. These events have not been captured in this search. However, 

detection likelihood is probably a bigger reason for coverage deficiencies of part-through wall flaws. A 

leak (or a non-leaking through-wall crack) is simply more likely to be detected. This issue is clearly 

illustrated by events in the data in which a detected leak prompted the plant to do a thorough inspection 

that found partial through-wall cracks. If the leak had not occurred and motivated the inspection, the 

partial through-wall cracks would not have been found. 

H.5 Selected Non-Pipe Database Summary Tables

The following tables list some summary data from the non-pipe database. 
H-4

Table H.1 Non-Pipe Event Counts by Component and Degradation Mechanism (for Each Plant 

Type) 

 Degradation Mechanism (see legend) 

Plant Type Component MA FDR SCC LC MF TF FS UNK 

BWR RPV 10 1 9 

 100% 0% 10% 90% 0% 0% 0% 0% 0% 

 Valve 1 1 

 100% 0% 0% 0% 0% 100% 0% 0% 0% 

 Pump 2 2 

 100% 0% 100% 0% 0% 0% 0% 0% 0% 

 Totals 13 0 3 9 0 1 0 0 0 

 Adjusted* 17 0.5 3.5 9.5 0.5 1.5 0.5 0.5 0.5 

 100% 3% 21% 56% 3% 9% 3% 3% 3% 

 

 MA FDR SCC LC MF TF FS UNK 

PWR Pzr 28 1 3 23 1 

 100% 4% 11% 82% 4% 0% 0% 0% 0% 

 RPV 42 5 27 5 5 

 100% 0% 12% 64% 12% 0% 0% 0% 12% 

 Valve 3 1 1 1 

 100% 33% 0% 33% 33% 0% 0% 0% 0% 

 SG 124 2 29 85 3 5 

 100% 2% 23% 69% 0% 0% 0% 2% 4% 

 Pump 2 2 

 100% 0% 100% 0% 0% 0% 0% 0% 0% 

 Instr nozzles 4 4 

 100% 0% 0% 100% 0% 0% 0% 0% 0% 

 Totals 203 4 39 140 7 0 0 3 10 

 Adjusted* 207 4.5 39.5 140.5 7.5 0.5 0.5 3.5 10.5 

 100% 2% 19% 68% 4% 0% 0% 2% 5% 

 

DM DM Description 

MA Material Aging 

FDR Fabrication Defect and Repair 

SCC Stress Corrosion Cracking 

LC Local Corrosion 

MF Mechanical Fatigue 

TF Thermal Fatigue 

FS Flow Sensitive (includes FAC and E/C) 

UNK Unknown 

* A half failure (0.5) was added to all degradation mechanism (DM) totals to force the representation of 

all DMs. 
mary of Non-Pipe Database by Plant Type and Piece Part (see Table H.3 for Legend of Piece Part Acronyms) 

No. Calendar Year 

Piece Part Records 78 84 86 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 

Pipe-w 1 1 

RCP-bdy 1 1 

RecP-hx 1 1 

RPV-crc 4 1 1 1 1 

RPV-crd 4 1 2 1 RPV-hbt 1 1 

RPV-noz 1 1 

FLG-fbs 1 1 

LIV-bon 2 1 1 

Pipe-w 4 1 1 1 1 

Pzr-bbs 4 1 1 1 1 

Pzr-hsl 8 2 1 1 2 1 1 

Pzr-noz 16 1 1 1 2 3 1 1 1 2 1 1 1 

RCP-noz 2 2 

RPV-crc 13 1 1 2 7 2 

RPV-crd 13 1 1 1 1 1 1 1 1 3 1 1 RPV-hdt 3 1 1 1 

RPV-ici 2 1 1 

RPV-noz 2 2 

RPV-pen 9 1 1 1 1 1 1 1 1 1 

Pipe-w 1 1 

SG-mwb 1 1 

SG-noz 3 1 1 1 

SG-tub 119 1 1 1 1 10 13 9 10 10 13 11 11 10 6 5 4 3 

Totals: 216 1 2 2 4 3 4 15 18 14 15 18 16 12 12 21 11 12 20 11 5 
H-6

Table H.3 Piece Part Legend 

PP-ID Piece Part 

FLG-fbs Flange Bolts 

LIV-bbs Bonnet Bolts 

LIV-bdy Valve Body 

LIV-bon Bonnet 

MSIV-bbs Bonnet Bolts 

MSIV-bdy Valve Body 

MSIV-bon Bonnet 

Pipe-w Weld 

PIV-bbs Bonnet Bolts 

PIV-bdy Valve Body 

PIV-bon Bonnet 

Pzr-bbs Pzr valve bonnet bolts 

Pzr-brv Bolted Relief Valves 

Pzr-hsl Heater Sleeves 

Pzr-mwb Manway Bolts 

Pzr-mwy Manway 

Pzr-noz Pzr Nozzles 

Pzr-rvb Relief Valve Bolts 

Pzr-shl Shell 

RCP-bdy Pump Body 

RCP-fwh Flywheel 

RCP-noz Pump Nozzles 

RCP-sel Pump Seals 

RecP-bbs Bonnet Bolts 

RecP-bdy Pump Body 

RecP-hx Pump cooler 

RecP-noz Pump Nozzles 

RecP-sel Pump Seals 

RPV-crc CRDM connections 

RPV-crd CRDM 

RPV-hbt Head Bolts 

RPV-hdb Head (bottom) 

RPV-hdt Head (top) 

RPV-ici In-Core Instru. 

RPV-noz RPV Nozzles (incl. Instr.) 

RPV-pen Penetrations 

SG-mwb Manway Bolts 

SG-mwy Manway 

SG-noz SG Nozzles 

SG-shl Shell 

SG-tbs Tube Sheet 

SG-tub Tube 
Table H.4 Summary of Non-Pipe Database by Plant Type, Piece Part and Failure Mode 

re No. Calendar Year 

e

a Records 78 84 86 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 

Part 1 1 

Part 1 1 

Part 1 1 

Part 1 1 

total: 4 1 1 1 1 

ilure 1 1 

ilure 1 1 

total: 2 1 1 

k 1 1 

k 1 1 k 2 1 1 

k 3 1 1 1 

total: 7 1 1 1 1 2 1 

otals: 13 3 1 2 1 2 1 2 1 

Part 1 1 

Part 3 1 1 1 

Part 2 1 1 Part 1 1 

Part 99 7 13 5 9 9 12 9 10 8 6 4 4 3 

total: 106 1 7 13 7 10 9 12 9 10 8 6 4 5 4 1 

ilure 2 1 1 

ilure 2 1 1 

ilure 2 1 1 

total: 6 1 2 1 1 1 

Full 4 1 2 1 Full 4 1 2 1 

Full 8 2 1 1 1 1 1 1 

Full 1 1 Full 4 4 

Full 2 1 1 

Full 3 1 1 1 
re No. Calendar Year 

ea Records 78 84 86 87 88 89 90 91 92 93 94 95 96 97 98 99 00 01 02 03 

Full 1 1 Full 2 1 1 

total: 29 3 3 3 3 3 3 8 3 

k 1 1 

k 1 1

k 2 1 1 

k 4 2 1 1 

k 8 1 1 1 3 1 1 

k 1 1 k 8 1 1 2 2 2 

k 11 1 1 1 1 1 1 1 2 1 1 k 2 2 k 4 1 1 1 1 k 18 1 1 1 1 3 3 1 1 2 1 2 1 

k 1 1 

total: 61 1 2 2 3 2 4 7 2 3 5 2 3 2 7 2 3 4 3 4 

 1 1 

otals: 203 1 2 2 4 3 4 15 15 13 15 16 15 12 12 19 11 11 18 10 5 

ugh-wall crack 

hrough-wall crack, but no active leak, typically indicated by boric acid deposit 

f a bolted connection
H-9

ATTACHMENT A TO APPENDIX H – LER REPORTING 

REQUIREMENTS 

The database relies upon LERs submitted by plants under the requirement of 10 CFR 50.73. Of the LERs 

reviewed for this effort, the two most commonly cited reporting requirements (each LER must reference 

the requirement that necessitates the LER.) are 50.73(a)(2)(i)(B) and 50.73(a)(2)(ii)(A). These are 

described below. 

50.73(a)(2)(i)(B) – Any operation or condition which was prohibited by the plant’s Technical 

Specifications. Westinghouse Standard Tech Specs (NUREG-1431, Vol. 1, Rev. 2, June 2001, Section 

3.4.13) related to RCS leakage are as follows. 

RCS operational leakage shall be limited to: 

a. No pressure boundary leakage 

b. 1 gpm (3.8 lpm) unidentified leakage 

c. 10 gpm (38 lpm) identified leakage 

d. 1 gpm (3.8 lpm ) total primary to secondary leakage through all steam generators (SGs), and 

e. 500 gallons (1,900 liters) per day primary to secondary leakage through any one SG 

Pressure Boundary Leakage is defined as leakage through a non-isolable fault in an RCS component 

body, pipe wall, or vessel wall (except SG leakage). Leakage past seals and gaskets is not considered

pressure boundary leakage. 

50.73(a)(2)(ii) – Any event or condition that resulted in: (A) The condition of the nuclear power plant, 

including its principal safety barriers, being seriously degraded; or (B) The nuclear power plant being in 

an unanalyzed condition that significantly degraded plant safety. 

NUREG-1022, Rev. 2 clarifies statement (A) as: 

This criterion applies to material (e.g., metallurgical or chemical) problems that cause abnormal 

degradation of or stress upon the principal safety barriers (i.e., the fuel cladding, RCS pressure boundary, 

or the containment). Abnormal degradation of a barrier may be indicated by the necessity of taking 

corrective action to restore the barrier’s capability . . . 

PWR tech specs also contain reporting guidance (via LERs) associated with the plants SG tube 

surveillance program. Typically, this reporting requirement is triggered when an inspection reveals that 

greater than 1% of the tubes in a SG are found to be defective (i.e., greater than 40% thru-wall crack). 
H-10

ATTACHMENT B TO APPENDIX H – SEQUENCE CODING AND 

SEARCH SYSTEM 

The SCSS is an NRC-sponsored database maintained by Oak Ridge National Laboratory (ORNL). It is a 

web-accessible database of LERs that can execute searches using a variety of criteria. It can be accessed 

at: 

http://scss.ornl.gov/

The following search criteria were used to generate the LER portions of the non-pipe database. 

LER SYSTEM EVENT SEARCH CRITERIA

Primary System(s) = SAB, SAF, SAE, SAA, SAD, SAI, SAH 

Interfacing System(s) = Any 

Include Trains/Channels = Yes 

Include Components = Yes 

Happening(s) = Any 

Event Cause(s) = Any 

Event Effect(s) = BH, BF, BE, BI, DE, BN, BL, BK, BP, BX, BC, BB, BA, BD 

Event Timing(s) = Any 

Detection Methods(s) = Any 

Nuclear Plant = Any 

Beginning Event Date = 01/01/1990 

Ending Event Date = 1/1/2003 

Maximum LERs = 2000 

This search returned 1,036 LERs. Basically, this search criteria looks for any leaking or cracking event 

associated with any primary coolant related system. The above search criteria rely upon the coding effort 

performed by the staff at ORNL as part of the SCSS program. In that effort, each LER is reviewed and 

characterized for possible relevance to each related system. This characterization includes both actual 

and possible system failures. Therefore, these search criteria returned both pipe and non-pipe failures, as 

well as many “non-failure” events. Each of the returned 1,036 LERs was reviewed and approximately 

80% (823 LERs) were judged to be non-failures and coded as not-applicable (NA). Most of these NA 

events were of the type where an engineering review or some other analysis was performed by the plant,

and it was found that a pipe was inadequately (compared to the design requirements) constrained such 

that if an earthquake were to occur, there was an increased chance that the pipe might fail. Another 

common “non-failure” example is of a problem unrelated to the integrity of the primary coolant system,

which would have adversely affected the ability of the plant to respond to a loss of coolant accident (i.e., a 

failure of the primary coolant system). These potential or possible issues were judged to not be actual 

failures and hence were deleted from the list. A further 34 LERs were removed from the set of LERs 

when they were found to document problems associated with pipe defects (or pipe-weld defects). 
H-11

ATTACHMENT C TO APPENDIX H - ESTIMATION OF FLAW 

FREQUENCY 

In general, a point estimate of the frequency of pipe failure (where ‘failure’ includes both small and large 

leaks and through-wall cracks, but excludes partial-through wall cracks), λ, is given by the following 

expression: 

NT

nF λ = (H.C.1) 

Where: 

nF = the number of failure events including both small and large leaks in the operational 

experience data; 

T = the total time over which failure events were collected; 

N = the number of components that provided the observed pipe failures. 

A point estimate of the total frequency of flaws (cracks and leaks), φ, is given by the following 

expression: 

φ + λ

⋅ ⋅ ⋅

+ =

⋅ ⋅ ⋅

=

FD

F C

FD

C

N T f P

n

NT

n

N T f P

n

 (H.C.2) 

Where: 

nC = the number of crack or flaw events 

f = the fraction of welds inspected for cracks or flaws

PFD = the probability that an inspected weld will find an existing flaw 

Nearly all through-wall leaks are found from independent observations such as routine leak inspections

and not from NDE inspections. However, part-through cracks are only typically found by NDE and thus 

the number is a function of the number of inspection locations. In Equation H.C.2 we account for the 

observed cracks in the data base and the fact that only a fraction (f) of the welds in the database are 

inspected according to ISI programs looking for cracks. The number of flaws actually discovered in ISI is 

subject to a finite NDE reliability, which is characterized by the factor PFD. 

If we now take the ratio of φ toλ, we get an expression for the factor by which to multiply the pipe failure 

rate to obtain the flaw (non-through wall crack) rate: 

1 / +

⋅ ⋅

= =

F FD

C

C F

n f P

n

R

λ

φ

 (H.C.3) 

Where: 

RC/F = Number of non-through wall cracks per leak event: 
H-12

One approach to assess the RC/F ratio is to evaluate those records where both cracks and leaks were found 

during a single inspection of a component of interest. Ideally, the best data would be found in those

instances where the component was 100% inspected. Without complete inspection, some assumptions 

about the inspection coverage, f, are required to assess this ratio. An example of this approach and the 

effect of the inspection coverage and POD is provided by analyzing the database for CRDM nozzle 

failures. For the component type ‘CRDM Nozzles’ in B&W PWR plants the database includes 6 LERs (= 

6 database records) as identified in Table H.C.1. A detailed review of each of these LERs revealed 

multiple failures and degradation. Equation (C.3) together with an assumption about the inspection scope 

(f) makes it possible to estimate RC/F.

Table H.C.1 B&W CRDM Nozzle Failures in ‘Non-Pipe’ Database 

Plant Date 

LER 

Number 

No. 

Components 

Leaking 

No. 

Components 

Cracked Population Comment 

Oconee-3 2/18/2001 2001-001 9 N/A2 69 

Expanded inspection 

of an additional 9 

nozzles. No recordable 

flaws. 

Oconee-3 5/2/2003 2003-001 2 N/A 69 RVH replaced 

Crystal 

River-3 10/1/2001 2001-004 1 N/A 69 

5 flaws found in 

CRDM Nozzle #32. 

Expanded inspection 

of 8 nozzles found no 

flaws. 

Three Mile 

Island-1 10/12/2001 2001-002 5 7 69 Inspection scope included 12 nozzles 

ANO-1 3/24/2001 2001-002 1 N/A 69 Visual inspection only 

of remaining nozzles. 

ANO-1 10/7/2002 2002-003 1 6 69 NDE of all nozzles 

 Totals: 19 13 

Estimates of RC/F for the data set in Table H.C.1 are presented in Table H.C.2 for different assumptions 

about the POD and fraction of welds inspected. This analysis also assumes that the inspection criteria and 

the cracking characteristics of the events listed in Table H.C.1 are representative of the entire population. 

The fraction of welds inspected is a function of the ISI program requirements. In Table H.C.2, a LB for f

is calculated using insights from piping reliability studies. This low f estimate results in the high RC/F

estimate presented in the table. 

The current ASME Section XI requirements are to inspect 25% of the Class 1 pipe welds and 7.5% of the 

Class 2 pipe welds. The current inspection practice for most if not all plants calls for the same welds to be 

inspected each inspection interval as opposed to randomly selecting a different set of welds for each 

interval. When cracks or significant flaws are found, the ASME code requires that an expanded search be 

made; however, the frequency of flaws and failures is so rare that this requirement adds very few 

additional inspections to the total population of inspected welds. Using data from an operating 4-loop

Westinghouse PWR unit on the number of Class 1 and Class 2 welds of 1,605 and 1,800, the following 

estimate of the parameter f is obtained for Westinghouse PWR plants: 

 

2 N/A in Table H.C.1 means that a full-scope NDE was not pursued. 
H-13

( ) ( )

( ) .0 157,1 605 ,1800

,1 605 25.0 ,1800 .0 075

=

+

+

f = (H.C.4) 

We assume this estimate of f to be representative of the non-pipe components. With additional 

assumptions about the reliability of the NDE we get the results as indicated in Table H.C.2. 

Table H.C.2 Estimates of RC/F for B&W CRDM Nozzles 

 PWSCC 

Parameter Data Source High Est. Median 

Est.

Low Est.

Number of cracks Table H.C.1 NA 13 NA 

Number of leaks Table H.C.1 NA 19 NA 

Fraction of components inspected, f Equation (H.C.4) – LB 1.0 0.5 0.157 

RC/F with PFD = 0.5 -- 9.72 3.74 2.37 

RC/F with PFD = 0.75 -- 6.81 2.82 1.91 

RC/F with PFD = 0.9 -- 5.84 2.52 1.76 

Hence, the relative number of flaws and leaks observed does not predict the relative frequency of flaws 

and leaks at a given weld. The estimate for the ratio of cracks to leaks obtained in Table H.C.2 reflects the 

degree to which components are exposed to PWSCC and inspection coverage. 

APPENDIX I 

REACTOR VESSEL LOCA PROBABILITY 

BASE CASE ANALYSES 

(BWR VESSELS AND PWR TOP HEAD NOZZLES) 

I-1

APPENDIX I 

REACTOR VESSEL LOCA PROBABILITY BASE CASE ANALYSES 

(BWR VESSELS AND PWR TOP HEAD NOZZLES)

I.1 Introduction

The LOCA expert panel elicitation team charter includes estimating the contribution to LOCA frequency 

from reactor vessels and other non-piping components. Extensive analyses were performed by members 

of the elicitation panel to develop LOCA frequencies for five piping “base cases” that were formulated by 

the panel in early meetings (documented as Appendices D, E, F and G to this NUREG). The piping base 

cases include failures on the piping side of vessel nozzles, including safe-ends. However, they do not 

include small diameter, partial penetration welded nozzles such as CRDM penetrations and other small 

nozzles, such as instrument nozzles, that aren’t connected to piping systems. In addition, the piping base 

cases do not include consideration of a leak from or rupture of other regions of the reactor vessel, such as 

the irradiated reactor vessel beltline or the low alloy steel portions of large vessel nozzles. LOCA 

frequency estimates for these cases are presented in this appendix, based on prior PFM analyses 

performed for PWR top head nozzles [I.1, I.2], the BWR Reactor Vessel Beltline Region [I.3, I.4], and 

BWR reactor vessel feedwater nozzles [I.5]. These estimates are used to construct a complete set of 

LOCA frequency tables for BWR and PWR reactor vessels, for all LOCA categories defined in the 

elicitation, a comparison of them to the aforementioned piping base cases is also presented. 

I.2 PWR Reactor Vessel Top Head Nozzles

Extensive PFM analyses have been conducted over the past several years to estimate the probability of 

leakage and rupture associated with the PWR CRDM penetration PWSCC problem [I.1, I.2]. The 

analysis model incorporates the following major elements: 

• computation of applied stress intensity factors for circumferential cracks in various nozzle 

geometries as a function of crack length, 

• determination of critical circumferential flaw sizes for nozzle failure, 

• an empirical (Weibull) analysis of the probability of nozzle cracking or leakage as a function 

of operating time and temperature of the RPV head, 

• statistical analysis of PWSCC crack growth rates in the PWR primary water environment as a 

function of applied stress intensity factor and service temperature, and 

• modeling of the effects of inspections, including inspection type, frequency and effectiveness. 

The model has been benchmarked with respect to field experience, considering the occurrence of cracking 

and leakage and of circumferential cracks of various sizes. Figures I.1 and I.2 illustrate the 

benchmarking. Figure I.1 presents a Weibull analysis of inspection results at thirty plants, of which 14 

detected leakage or cracking (data points in the figure). The remaining plants that were inspected and 

found clean were treated as “suspended tests” according to standard Weibull analysis theory [I.2]. The 

data are plotted in terms of effective degradation years (EDYs) which are equivalent operating years at 

600oF (315oC), using an activation energy (Arrhenius) model [I.1] to adjust for different head operating 

temperatures. For plants in which multiple cracked nozzles were detected in the inspections, the data

were extrapolated back to the expected time of first cracking or leakage, using an assumed Weibull slope 

of 3. The straight line through the data represents a medium rank Weibull regression (also with a slope of 

3) upon which the probability of leakage predictions in the model are based. Figure I.2 illustrates the 

benchmarking process used for the crack growth analysis algorithm in the model with respect to CRDM 

nozzles that exhibited circumferential cracks of various sizes. (Eleven (11) nozzles out of a total of 881 

inspected nondestructively through the spring of 2003 exhibited circumferential cracking. No additional 
I-2

circumferential cracking has been detected in more recent inspections.) The figure shows that, when 

using original analysis parameters, the crack growth model under-predicted the probability of 

circumferential cracking somewhat, but after adjustment of selected analytical parameters, the PFM 

model was “benchmarked” so as to very accurately predict the field results, especially for the most 

important, larger crack sizes. 

The benchmarked model was then used to evaluate the probabilities of nozzle failure and leakage in 

actual plants. A sample of the results is presented in Figures I.3 and I.4. Figure I.3 illustrates the 

probability of nozzle failure (ejection of a nozzle) for a head operating temperature of 580oF (304oC), the 

approximate average of U.S. PWRs. No inspections were assumed to be performed during the first 25 

years of plant operation, resulting in the probability of nozzle failure constantly increasing with time 

during that period. The analysis then assumed that inspections begin after 25 years, at intervals and 

detection levels representative of current requirements [I.6]. It is seen from the figure that the current 

inspection regimen reduces the nozzle failure probability significantly. 

Ejection of a 4 inch CRDM nozzle [2.75 inch (~70 mm) ID] due to a circumferential crack would yield a 

one-sided LOCA corresponding approximately to Category 2 LOCA [>1,500 gpm (5,700 lpm) but < 

5,000 gpm (19,000 lpm)]. If periodic inspections are continued, with any nozzles in which leakage or 

cracking are detected repaired or the heads replaced (as is common practice), the nozzle ejection 

probability will be even lower in the future. Table I.1 below provides a summary of the average failure 

probabilities from Figure I.3, between 0 and 25 years, and from 25 to 40 years. The probability of failure 

for 40 to 60 years was not calculated, but was assumed to be the same as 25 to 40 years, on the basis that 

the current inspection regimen will be maintained, or the heads replaced. A Category 3 break was 

assumed to require multiple nozzle failures, the probability of which was computed via a binomial 

distribution for the typical number of nozzles in a head. As seen in Table I.1, the probabilities of 

simultaneous multiple nozzle failures is quite low.

 

Figure I.4 illustrates similar PFM results (based on the above Weibull model) for the probability of small 

amounts of leakage from a top head CRDM nozzle. The same inspection regimen was assumed as in the 

nozzle ejection analysis (no inspections from 0 to 25 years, inspections in accordance with current 

requirements thereafter). A small leak from a CRDM nozzle was assigned as a Category 0 break [less 

than 1 gpm (3.8 lpm)] in Table I.1, and the intermediate, Category 1 break size was obtained by 

logarithmic interpolation between Categories 0 and 2. 
I-3

Table I.1 Summary of PWR CRDM Nozzle PFM Results 

Average LOCA Probabilities 

During Operating Years: 

Break 

Category 

Leak Rate 

>(gpm) 

0-25 25-40 40-60 

0 1 2.00E-02 5.00E-03 5.00E-03

1 100 1.27E-03 2.75E-04 2.75E-04

2 1,500 2.50E-04 5.00E-05 5.00E-05

3 5,000 4.00E-08 2.00E-09 2.00E-09

4 25,000 - - - 

5 100,000 - - - 

6 500,000 - - - 

I.3 BWR Reactor Vessels

Analyses have been previously submitted and approved [I.3, I.4] that establish reduced inspection 

requirements for BWR reactor vessels relative to ASME Section XI requirements. Specifically, 

BWRVIP-05 [I.3] justifies that only axially-oriented welds in the vessel beltline region need be examined 

on a ten year interval, versus the Section XI requirement to inspect all axial and circumferential welds on 

this interval. This relief was based on PFM calculations demonstrating, for the BWR fleet, that 

circumferential weld inspections contribute negligibly to reduction in the already small failure probability 

of a BWR vessel. The methodology used for the PFM analysis is a computer program (VIPER [I.7]) 

developed by Structural Integrity Associates for EPRI and the BWRVIP. To address this LOCA 

frequency contributor, the VIPER software was run for a typical BWR vessel, extending the analysis 

period from 40 to 60 years. A modification to the software (VIPER-NOZ) was also used to estimate 

leakage and failure probabilities for BWR Reactor Vessel feedwater nozzles. Feedwater nozzles were 

selected because they are subject to thermal fatigue cycling, which caused serious nozzle cracking in the 

1970s [I.5]. Both analyses take credit for routine ISI programs that are conducted on these components 

on ten-year inspection intervals. The feedwater nozzle analysis also takes credit for nozzle modifications 

and thermal sleeve improvements that were installed in all U.S. BWRs to reduce the severity of the 

thermal fatigue cycling. 

I.3.1 BWR Vessel Beltline Region

In the VIPER software, cracks are assumed to exist in BWR vessel welds due to two causes – original 

manufacturing defects and service-induced cracks which initiate in the stainless steel cladding. These 

cracks are assumed to grow as a function of operating time due to fatigue crack growth and SCC of the 

low alloy steel vessel material. Simultaneously, the vessel beltline region is assumed to embrittle due to 

irradiation. Monte Carlo simulations of these processes are employed in VIPER, which include fracture

mechanics crack growth calculations due to fatigue and SCC, and a comparison of predicted crack sizes 

to the critical crack size due to normal operation as well as possible transient conditions. The governing 

transient condition was determined to be a LTOP event, since BWRs are not subject to PTS. 

The effects of ISI are imposed at appropriate inspection intervals, assuming a POD curve for the 

inspections. Flaws that are detected during ISI are assumed to be repaired, and thus eliminated from the 

population, such that they can no longer grow to a leak or vessel failure. 

The axial vessel beltline welds are divided into a series of segments, and each segment is analyzed 

separately to account for axial gradients of irradiation fluence in the welds, which peaks at the core

centerline, and decays at elevations above and below that location. The failure frequencies from each
I-4

segment are weighted by their respective weld volume, and summed to determine failure frequency for 

the entire vessel. 

Modes of failure considered are: 

1. Vessel fracture during normal operation (KI

 > KIc) 

2. Vessel fracture during an assumed LTOP event. The LTOP event considered is pressurization to 

1,150 psi (7.93 MPa) at 88oF (31oC), which is assumed to occur at a frequency of 1E-3. 

3. Predicted crack growth to 80% of wall thickness before failure modes 1 or 2 occurs (LBB) 

The results for a typical BWR are given in the following table: 

Table I.2 Summary of BWR RPV Beltline PFM Results 

Average LOCA Probabilities 

During Operating Years:

Break 

Category 

Leak Rate 

>(gpm) 

0-25 25-40 40-60 

1 100 1.00E-08 2.98E-08 4.57E-08

2 1,500 2.32E-09 4.31E-09 2.84E-08

3 5,000 1.21E-09 1.83E-09 2.30E-08

4 25,000 5.04E-10 5.79E-10 1.73E-08

5 100,000 2.38E-10 2.15E-10 1.36E-08

6 500,000 9.86E-11 6.79E-11 1.02E-08

These provide an estimate of the probability (per vessel year) of breaks of various sizes due to vessel 

beltline failures. To complete this table, it was assumed that a leak (LBB mode failure) corresponds to a 

crack of length = 60 inches (1525 mm) that breaks through and begins leaking as a through-wall crack of 

this length (since the wall thickness is approximately 6 inches (150 mm), and cracks in VIPER are 

assumed to have a ten to one aspect ratio). Dave Harris ran this case using the PRAISE code leakage rate 

prediction capability (See Appendix F for description), and computed a leak rate of 193 gpm (733 lpm) 

for an axial crack of this size in a BWR vessel. Thus, predicted LBB mode failures from the RPV beltline 

were treated as Category 1 breaks. Predicted vessel fractures, either during normal operation or due to 

LTOP events were treated as complete RPV ruptures, which were assumed to result in very large, 

Category 6 breaks. Intermediate break sizes were then determined by log-log interpolation between these 

two extremes. The sharp increase in large break probability between years 40 and 60 is attributable to the 

combined effect of two aging mechanisms – crack growth and RPV embrittlement. 

I.3.2 BWR Feedwater Nozzles 

RPV nozzles constitute another potential non-piping LOCA concern. The example used to address this 

concern here is BWR feedwater nozzles, which have in the past been subject to thermal fatigue cracking

[I.5]. The thermal fatigue problem was caused by mixing of hot reactor water and relatively cold 

feedwater (see Figure I.5) during reactor startups, shutdowns and other periods of low power operation, 

when feedwater heating is generally unavailable. Cracking of various depths, up to 1.5 inches (38 mm), 

was detected in a number of BWRs in the 1970s (see Figure I.6). At that time, the standard feedwater 

nozzle design incorporated a loose-fitting thermal sleeve/sparger configuration, as shown in Figure I.5. 

Since then, all U.S. BWRs have installed some type of fix, employing either welded-in spargers or 

multiple-sleeve designs with shrink fits and piston rings to protect the nozzle from the effects of the cold 

feedwater. No subsequent cracking has been discovered since the improved thermal sleeves were 

installed. 
I-5

In order to perform a base case analysis of this problem, a modification to the software (VIPER-NOZ) 

was developed to estimate leakage and failure probabilities for BWR Reactor Vessel feedwater nozzles. 

The substantive changes to the VIPER software in VIPER-NOZ were the addition of thermal fatigue 

crack initiation and growth algorithms specific to the feedwater nozzle thermal cycling phenomenon, and 

zeroing out the effects of irradiation embrittlement, since feedwater nozzles are far enough from the 

reactor core region that neutron fluence effects are small. The VIPER-NOZ software was run for 

conditions representative of the original nozzle/sparger designs, to confirm that cracking probabilities 

consistent with early field experience (Figure I.3) are predicted. The boundary conditions were then 

modified to represent improved nozzle/sparger designs, which reduce the effects of thermal fatigue on the 

nozzle. The analyses were conducted for a 60 year operating lifetime, and included the effects of periodic 

ISI, which are performed for these nozzles on ten-year intervals. The results are given in the following 

table: 

Table I.3 Summary of BWR Feedwater Nozzle PFM Results 

Average LOCA Probabilities 

During Operating Years:

Break 

Category 

Leak Rate 

>(gpm) 

0-25 25-40 40-60 

0 1 <1.00E-06 1.47E-06 1.25E-06

1 100 <1.00E-06 <1.00E-06 <1.00E-06

2 1,500 <1.00E-06 <1.00E-06 <1.00E-06

3 5,000 <1.00E-06 <1.00E-06 <1.00E-06

4 25,000 <1.00E-06 <1.00E-06 <1.00E-06

The predicted leakage cases were treated as Category 0 breaks in this case, and since the nozzle is 

attached to a 12 inch diameter pipe, the maximum credible break size was assumed to correspond to 

single ended rupture of a 12 inch pipe, which corresponds to a Category 4 break. A total of 1 million 

simulations were run, and except for LBB type failures at 40 and 60 years, no other failures were 

predicted. Thus a failure frequency of less than 1E-6 is given for most entries in the above table. 

I.4 Combined LOCA Frequencies due to Reactor Vessel Failures

Tables I.4 and I.5 summarize and combine the above RPV LOCA frequency results for BWRs and PWRs, 

respectively. For BWRs, three RPV LOCA contributors are addressed: RPV vessel beltline region, large 

nozzles (6 through 28 inch diameter), and small penetrations (partial penetration welded nozzles, 4 inch in 

diameter or less, such as CRDs). The individual LOCA probability contributions for each of these are 

provided in the top three sections of Table I.4, and they are summed in the bottom section of the table. 

These address all LOCA categories as well as the three time periods under consideration (0-25 yrs, 25-40 

yrs, and 40-60 yrs). Note, in Table 4.4 in Section 4 of the main body of this report, only the results for 

the BWR beltline region and the large feedwater nozzles are presented. The results for the BWR CRDs 

and other small penetrations listed in Table I.4 of this appendix were not included in Table 4.4 since the 

estimates for the BWR CRDs and other penetrations are not based on analysis, but instead, were based on 

engineering judgment, i.e., BWR CRDs and other penetrations LOCA frequencies were simply assumed 

to be a factor of 10 less than the PWR CRDM LOCA frequencies, which were based on analysis. 

The first contributor is the BWR shell region, the failure probabilities for which are dominated by the 

irradiated reactor vessel beltline region. The upper section of Table I.4 summarizes the results of this 

analysis from Table I.2, in terms of the probability of leaks of various sizes due to degradation or failure 

of the RPV beltline region. (Note that these were modified slightly relative to those in Table I.2 to

eliminate the negative time factor for Category 5 and 6 LOCAs in the 25 – 40 year period.) 
I-6

The second section of Table I.4 addresses large nozzle contributions to LOCA probability, which in 

BWRs are assumed to correspond to the 12 inch diameter Feedwater Nozzles that experienced thermal 

fatigue cracking in the 1980s [I.5]. LOCA probabilities due to this contributor are given for break 

Categories 1 through 4, taken from Table I.3 as this nozzle size could not lead to larger break sizes. For 

the Category 2 through 4 LOCAs for 0-25 year time frame, an assumption was made as to the size factor 

that each successive LOCA size greater than Category 1 was 5 times less likely to occur than the previous 

size LOCA. Then, for the 25-40 and 40-60 year time frames, the same time factor as determined for the 

Category 0 LOCAs in Table I.3 was assumed for the larger size LOCAs. Breaks of the other, larger 

diameter nozzles, such as recirculation outlet nozzles, are considered to be adequately encompassed by

the vessel beltline case. 

Finally, the third section of Table I.4 lists LOCA probabilities due to failures of CRDs and other small 

penetrations in the BWR vessel. These were estimated from the detailed analysis of PWR CRDM 

penetrations described above (Table I.1) but they assume that the BWR penetrations have about an order

of magnitude lower LOCA probability than similar penetrations in a PWR. The order of magnitude 

reduction is deemed appropriate, because problems in small vessel penetrations in BWRs have occurred 

at a much lower frequency than the recent PWSCC experience in PWRs, upon which Table I.1 is based. 

The problems in BWR penetrations have also been attributed to a fairly well-understood phenomenon 

(IGSCC) and in most cases the nozzles of concern have been mitigated by design and materials changes. 

Table I.5 provides a similar summary for PWR RPVs. In this case, LOCA probabilities are reported for 

only two categories of LOCA contributors, the shell region (RPV beltline) and small penetrations. Again, 

as was the case for BWRs, Table 4.5 in Section 4 only includes the results for PWR CRDMs. It does not

include the results for the PWR beltline region as reported in Table I.5 of this appendix. As was the case 

for BWR CRDs and other penetrations, the PWR beltline results in Table I.5 are not based on analysis. 

Again, they are based on engineering judgment, i.e., the PWR beltline LOCA frequencies in Table I.5 of

this appendix were simply assumed to be a factor of ten greater than the BWR beltline LOCA frequencies

from Table I.4. It was judged that the large nozzles in a PWR RPV do not pose a significant LOCA risk 

because they are not subject to significant thermal cycles such as the BWR Feedwater nozzles, and except 

for the safe-ends (which are covered in the piping elicitation), they have not experienced any degradation 

mechanisms to date. The contributions for the two PWR RPV LOCA contributors are summed in the 

bottom section of Table I.5. 

 

For the PWR beltline region, results from a prior analysis of a PWR vessel using a third version of the 

VIPER software (VIPER-PWR) were reviewed. Based on this review, it was estimated that the PWR 

RPV beltline region presents about an order of magnitude increase in large rupture probability relative to 

that of a BWR, because PWR beltlines are more highly irradiation embrittled, and because they are 

potentially subject to PTS transients. Thus, the BWR RPV beltline region LOCA frequency entries in 

Table I.4 were multiplied by a factor of ten and entered in the upper section of Table I.5. 

PWR CRDM penetrations results were entered directly from the above PFM analysis results in Table I.1. 

I.5 Summary and Comparison to Piping Base Cases

Figures I.7 and I.8 present plots of these RPV base cases, compared to the piping base cases from 

Appendices D, E, F, and G. For purposes of this comparison, a single set of piping base case LOCA 

frequencies were derived that are a composite of the results from the four appendices. Plots are presented 

for the 0-25 year (Figure I.7) and the 25-40 year (Figure I.8) periods. Since the RPV LOCA frequencies 

for the 40-60 year period are not significantly different than the 25-40 year results, a separate plot for that 

case is not included. It is seen from these figures that the RPV base cases are at the low end of the piping 
I-7

LOCA probabilities for the large break Categories 5 and 6, but are at the high end for small, Category 1 

and 2 breaks, due largely to the small penetration (CRDM) contributions discussed above. Note also that 

the small LOCA probability estimates are substantially lower in the outlying years (25-40 and 40-60) 

because of inspection programs implemented as a result of these issues. In general, small break LOCA 

frequency contributors (Categories 1 and 2) from PWR RPVs are seen to be greater than those for BWRs, 

due to the PWSCC concern in CRDM and other small penetrations. Large break LOCA contributors 

(Categories 5 and 6) are also estimated to be greater for PWR RPVs due to higher irradiation 

embrittlement and the potential for PTS transients.
I-8

Table I.4 LOCA Frequencies for BWR Reactor Pressure Vessel Base Case

BWR RPV Beltline 

Break 

Cat. Break Size 

Average LOCA Probabilities 

During Operating Years:

 gpm NPS 0-25 yrs 25-40 yrs 40-60 yrs 

 TimeFactor TimeFactor 

1 100 0.5 1.00E-08 2.98E-08 2.98 4.57E-08 4.57

2 1,500 1.5 2.32E-09 6.19E-09 2.67 2.84E-08 12.24

3 5,000 3.5 1.21E-09 3.12E-09 2.58 2.30E-08 19.01

4 25,000 7 5.04E-10 1.25E-09 2.47 1.73E-08 34.33

5 100,000 16 2.38E-10 5.65E-10 2.37 1.36E-08 57.14

6 500,000 30 9.86E-11 2.32E-10 2.35 1.02E-08 103.45

BWR FW Nozzles

Break 

Cat. Break Size 

Average LOCA Probabilities 

during Operating Years:

 gpm NPS 0-25 yrs 25-40 yrs 40-60 yrs 

 TimeFactor TimeFactor 

1 100 0.5 1.00E-06 1.47E-06 1.47 1.25E-06 1.25

2 1,500 1.5 2.00E-07 2.94E-07 1.47 2.50E-07 1.25

3 5,000 3.5 4.00E-08 5.88E-08 1.47 5.00E-08 1.25

4 25,000 7 8.00E-09 1.18E-08 1.47 1.00E-08 1.25

5 100,000 16 

6 500,000 30

BWR CRDs & Other Small Penetrations

Break 

Cat. Break Size 

Average LOCA Probabilities 

during Operating Years:

 gpm NPS 0-25 yrs 25-40 yrs 40-60 yrs 

0 2.00E-03 5.00E-04 Factor 5.00E-04 Factor 

1 100 0.5 1.27E-04 2.75E-05 0.22 2.75E-05 0.22

2 1,500 1.5 2.50E-05 5.00E-06 0.20 5.00E-06 0.20

3 5,000 3.5 4.00E-09 2.00E-10 0.05 2.00E-10 0.05

4 25,000 7 

5 100,000 16 

6 500,000 30

BWR Vessel - Totals

Break 

Cat. Break Size 

Average LOCA Probabilities 

during Operating Years:

 gpm NPS 0-25 yrs 25-40 yrs 40-60 yrs 

 TimeFactor TimeFactor 

1 100 0.5 1.28E-04 2.90E-05 0.23 2.88E-05 0.23

2 1,500 1.5 2.52E-05 5.30E-06 0.21 5.28E-06 0.21

3 5,000 3.5 4.52E-08 6.21E-08 1.37 7.32E-08 1.62

4 25,000 7 8.50E-09 1.30E-08 1.53 2.73E-08 3.21

5 100,000 16 2.38E-10 5.65E-10 2.37 1.36E-08 57.14

6 500,000 30 9.86E-11 2.32E-10 2.35 1.02E-08 103.45
I-9

Table I.5 LOCA Frequencies for PWR Reactor Pressure Vessel Base Case

PWR RPV Beltline 

Break Cat. Break Size Pete Riccardella Estimate 

 gpm NPS 0-25 yrs 25-40 yrs 40-60 yrs 

 Factor Factor 

1 100 0.5 1.00E-07 2.98E-07 2.98 4.57E-07 4.57

2 1,500 1.5 2.32E-08 6.19E-08 2.67 2.84E-07 12.24

3 5,000 3 1.21E-08 3.12E-08 2.58 2.30E-07 19.01

4 25,000 7 5.04E-09 1.25E-08 2.47 1.73E-07 34.33

5 100,000 14 2.38E-09 5.65E-09 2.37 1.36E-07 57.14

6 500,000 30 9.86E-10 2.32E-09 2.35 1.02E-07 103.45

PWR CRDMs

Break Cat. Break Size Pete Riccardella Estimate 

 gpm NPS 0-25 yrs 25-40 yrs 40-60 yrs 

0 2.00E-02 5.00E-03 Factor 5.00E-03 Factor 

1 100 0.5 1.27E-03 2.75E-04 0.22 2.75E-04 0.22

2 1,500 1.5 2.50E-04 5.00E-05 0.20 5.00E-05 0.20

3 5,000 3.5 4.00E-08 2.00E-09 0.05 2.00E-09 0.05

4 25,000 7 

5 100,000 16 

6 500,000 30 

PWR Vessel - Totals

Break Cat. Break Size Pete Riccardella Estimate 

 gpm NPS 0-25 yrs 25-40 yrs 40-60 yrs 

 Factor Factor 

1 100 0.5 1.27E-03 2.75E-04 0.22 2.76E-04 0.22

2 1,500 1.5 2.50E-04 5.01E-05 0.20 5.03E-05 0.20

3 5,000 3.5 5.21E-08 3.32E-08 0.64 2.32E-07 4.45

4 25,000 7 5.04E-09 1.25E-08 2.47 1.73E-07 34.33

5 100,000 16 2.38E-09 5.65E-09 2.37 1.36E-07 57.14

6 500,000 30 9.86E-10 2.32E-09 2.35 1.02E-07 103.45

 

I.6 References

I.1 Peter Riccardella, Nathaniel Cofie, Angah Miessi, Stan Tang, Bryan Templeton, “Probabilistic 

Fracture Mechanics Analysis to Support Inspection Intervals for RPV Top Head Nozzles” U.S. 

Nuclear Regulatory Commission / Argonne National Laboratory Conference on Vessel Head 

Penetration Inspection, Cracking, and Repairs, September 29 – October 2, 2003, Gaithersburg, 

Maryland. 

I.2 Materials Reliability Program, MRP-105, “Probabilistic Fracture Mechanics Analysis of PWR 

Reactor Pressure Vessel Top Head Nozzle Cracking,” EPRI Report 1007834 (EPRI Licensed 

Material), May, 2004. 

I.3 EPRI Report, “BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP05),” TR-105697, September 1995. 
I-10

I.4 NRC Report, “Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report 

(TAC No. M93925),” Division of Engineering, Office of Nuclear Reactor Regulation, May 1998. 

I.5 NUREG-0619, “BWR Feedwater Nozzle and CRD Return Line Nozzle Cracking, Resolution of 

Generic Tech Activity A-10,” November 1980. 

I.6 U.S. NRC Order EA-03-009, “Interim Inspection Requirements for Reactor Pressure Vessel Heads at 

Pressurized Water Reactors”, issued on February 11, 2003. 

I.7 VIPER Version 1.2, Structural Integrity Associates, Report # SIR-95-098 Rev. 1, Feb. 1999. 
I-11

Oconee 1

Crystal River 3

North Anna 2

 Millstone 2

Beaver Valley 1

TMI-1

Oconee 3

Cook 2

ANO-1

Surry 1

Davis-Besse

St. Lucie 2

North Anna 1

Oconee 2

0.02

0.01

0.05

0.10

0.20

0.50

0.63

0.90

1 10 100

EDYs

Cumulative Fraction of Units with Leakage

EDYs at detection of leakage or cracking

EDYs at first leak or crack extrapolated

back using slope b = 3

All inspection data adjusted to 600 °F (Q = 50 kcal/mole)

Weibull slope of

b = 3 assumed for fit

to extrapolated data 

Median Rank Regression yields 

Weibull characteristic

time θ = 15.21 EDYs

Figure I.1 Weibull Plot of Plant Inspection Data Showing Extrapolation Back to Time of First 

Leakage or Cracking. Plants that Performed NDE and were Found Clean are 

Treated as Suspensions
I-12

11

7

6

3

2

1.00E-03

1.00E-02

1.00E-01

30 60 90 120 150

Circ. Crack Length (degrees)

Prob. of Circ. Crack Exceeding

Plant Cum. Size Dist.

Base Case Parameters

Benchmarked Parameters

Figure I.2 Benchmarking of PFM Crack Growth Analyses with Respect to Field-Observed 

Circumferential Cracking of Various Lengths
I-13

0.0E+00

5.0E-04

1.0E-03

1.5E-03

2.0E-03

2.5E-03

0 5 10 15 20 25 30 35 40

Operating Years

Prob. of Nozzle Ejection (per year)

 w/ NRC Inspection Plan 

w/ MRP Inspection Plan

Baseline

Inspection

Subsequent

Inspections

Figure I.3 RPV Top Head PFM Analysis Results for Plant with 580oF (304oC) Head Temperature – 

Probability of CRDM Nozzle Failure (i.e. Ejection of Nozzle from Vessel Head) 
I-14

0.00%

0.50%

1.00%

1.50%

2.00%

2.50%

3.00%

3.50%

4.00%

0 5 10 15 20 25 30 35 40

Operating Years

Prob. of Leak (per year)

Insp. Per NRC Order

Insp. per MRP Plan

Baseline

NDE

Subsequent

Inspections

Figure I.4 RPV Top Head PFM Analysis Results for Plant with 580oF (304oC) Head Temperature – 

Probability of Leakage from CRDM Nozzle 
I-15

Figure I.5 Schematic of Thermal Fatigue Cracking in BWR Feedwater Nozzles 
I-16

Figure I.6 Historical BWR Feedwater Nozzle Cracking Experience (circa 1980) 
I-17

Average LOCA Frequencies; 0-25 Years 

1.00E-10

1.00E-09

1.00E-08

1.00E-07

1.00E-06

1.00E-05

1.00E-04

1.00E-03

100 1500 5000 25000 100000 500000

Break Size (GPM)

Probability

BWR-1 (RR)

BWR-2 (FW)

PWR-1 (RCL-HL)

PWR-2 (Surge)

PWR-3 (HPI/MU)

PWR-Vessel

BWR-Vessel

Figure I.7 Comparison of RPV and Piping Base Case LOCA Frequencies Versus Break Size 

(0 to 25 Years) 
I-18

Average LOCA Frequencies; 25-40 Years

1.00E-10

1.00E-09

1.00E-08

1.00E-07

1.00E-06

1.00E-05

1.00E-04

1.00E-03

100 1500 5000 25000 100000 500000

Break Size (GPM)

Probability

BWR-1 (RR)

BWR-2 (FW)

PWR-1 (RCL-HL)

PWR-2 (Surge)

PWR-3 (HPI/MU)

PWR-Vessel

BWR-Vessel

Figure I.8 Comparison of RPV and Piping Base Case LOCA Frequencies Versus Break Size 

(25 to 40 Years) 
APPENDIX J 

ELICITATION QUESTIONS 

J-1

APPENDIX J 

ELICITATION QUESTIONS

J.1 Instructions

There are four basic quantities that are the ultimate focus of this exercise: the LOCA frequencies of

piping components, the LOCA frequencies of non-piping components, the LOCA probabilities of piping 

components after emergency faulted loading, and the LOCA probabilities of non-piping components after 

emergency faulted loading. The elicitation will be structured so that each of these questions can be 

answered using one of two question sets. The question sets are structured to decompose the underlying

issues using different approaches so increase your flexibility. 

The bottom-up approaches (3A, 4A, 5A, 6A) could entail significantly more work if every piping 

and non-piping system is evaluated. It is recommended that people choosing this approach focus on 

significant contributing issues in only significant piping and non-piping systems to reduce the 

workload. Similarly the people choosing the top-down approaches (3B, 4B, 5B, 6B) may want to 

ensure that their significant issues are manifested correctly within relevant systems. These 

strategies allow you to combine features of each methodology. 

Only a few additional examples of these questions are provided in this document. Many examples will be 

similar to those included in the Elicitation Question Development document. Please refer to this 

document and the Kick-off Meeting Notes document as indicated within the notes section for the 

questions. 

J.1.1 Specific Instructions: Minimum Requirements Prior to Your Elicitation 

A1. Provide answers to the questions in the “Base Case Evaluation” area. 

A2. Provide MV estimates for the question set within the “Regulatory and Utility Safety Culture” 

area. 

A3. Provide MV estimates for at least one question set within the “LOCA frequencies of Piping 

Components” area. 

A4. Provide MV estimates for at least one question set within the “LOCA frequencies of Non-Piping 

Components” area. 

A5. Provide MV estimates for at least one question set within the “LOCA Probabilities of Piping 

Components under an Emergency Faulted Load” area. 

A6. Provide MV estimates for at least one question set within the “LOCA Probabilities of Non-Piping 

Components under an Emergency Faulted Load” area. 

A7. Categorize the uncertainty ranges (90% coverage intervals) associated with your MV estimates in 

A2 – A6 as low, medium, or high. 

J.1.2 Specific Instructions: Additional Questions During Your Elicitation 

We will be asking for your response within the following general areas. 

B1. Provide rationale and discuss those important issues that you identified and quantified in 

questions A2 – A6. 

B2. Quantify the uncertainty ranges (90% coverage intervals) associated with estimates provided for 

A2 – A6. This will quantify the initial responses in A7. 

B3. Provide MV estimates for the question sets that you did not initially answer in A2 – A6. 

B4. Quantify uncertainties associated with answers in B3. 

B5. Ensure that the critical issues for LOCA frequencies are captured. 
J-2

J.2 Elicitation Questions

J.2.1 Elicitation Question 1: Base Case Evaluation 

The following questions will be asked to solicit your opinion about the base case evaluation. These 

questions are necessary to determine how the rest of your responses will be anchored, i.e., if the base case 

conditions will be used for anchoring or if you prefer to anchor using a specific set of results. Therefore, 

you should only consider the general approach used by each base case team member and not specific 

results. You will be given additional opportunity later to provide your assessment of the specific results. 

Of course, you can also provide additional information that can be used to provide anchoring to the rest of 

your estimates, including your own set of results for the base cases. All the questions below will be asked 

for each of the LOCA size categories and the evaluation time periods. 

1A.1. Do you think the base case results reflect the same conditions in each of the four team 

member’s calculations? 

1A.2. If not, which panelists’ results best describe the base case conditions? 

1B.1. Do you think that the differences in the four base case team members’ results are a reasonable 

reflection of the range of variability in the true LOCA probability? 

1B.2. If not, do the results under or over estimate your opinion of the true uncertainty? 

1C. Do you think that any particular base case results are more accurate? 

1D. Do you wish to anchor your responses on either the base case conditions, or on a specific team 

member’s results? 

J.2.2 Elicitation Question 2: Safety Culture 

2A.1. Consider the current utility safety culture that exists after approximately 25 years (current-day) 

of plant operation and how it influences Category 1 LOCAs. Express the relative change, or 

ratio, in the utility safety culture’s effect on LOCA frequencies after 15 additional years (40 

years of operation) compared to its current-day effect. Next, express the ratio of the utility 

safety culture’s effect on LOCA frequencies ratio in 35 years (60 years of plant operation) to its 

current-day effect. Include the 90% coverage interval for all estimates. 

2A.2. Repeat 2A.1 but now considering the effect of the regulatory safety culture on LOCA 

frequencies. 

2A.3. If you believe that safety culture effects are a function of leak rate category, repeat 2A.1 and 

2A.2 for Category 2 through Category 6 LOCA frequencies. 

2A.4. Do you believe that the utility safety culture and regulatory safety culture are correlated? If so, 

is the correlation high, medium, or low? 

Notes: 

a. Some aspects of regulatory and utility safety culture are discussed in the Kick-off Meeting 

Notes. These aspects can be considered independently and then combined or the aspects can 

be considered in the aggregate in question 2A. 

b. Please see EQ 9 in the Elicitation Question Development document for additional information 

and an example for this question. 

c. A ratio greater than 1 indicates that the safety culture will result in an proportional increase in

the future LOCA frequency compared to the current LOCA frequency. Similarly, ratios less 

than 1 indicate that the LOCA frequencies will decrease as a function of safety culture. 
J-3

J.2.3 Elicitation Question 3: LOCA Frequencies of Piping Components 

Question Set 3A 

3A.1.1. Consider Category 1 LOCAs for the PWR cold leg reference case. Choose a base case to 

compare with this reference case for this LOCA size at 25, 40, and 60 years of plant operation. 

Determine the ratio of LOCAs in the cold leg reference case to the chosen base case at each 

time period. Also, estimate the 90% coverage interval for these ratios. 

3A.1.2. Repeat 3A.1.1 for each LOCA size category for the cold leg. 

3A.1.3. Repeat 3A.1.1 and 3A1.2 for all other PWR and BWR reference cases. 

Notes: 

a. Piping base and reference case conditions are described in the Kick-off Meeting Notes

document. 

b. Please see EQ 10 in the Elicitation Question Development document for additional 

information and an example for this question. 

c. Any base case can be chosen at any specified period of time (25, 40, and 60 years) for 

anchoring. Please note the time period of your chosen base case time period is different than 

the time period being analyzed. 

3A.2.1. List the specific combinations of the variables (i.e., material, geometry, degradation 

mechanism, loading, and mitigation/maintenance) which are the most significant contributors to 

PWR cold leg Category 1 LOCA frequency as a function of plant operating time (25, 40, 60 

years). The list should total at least 80% of the total contribution to all cold leg Category 1 

LOCAs. Estimate the MV contribution of these systems (> 80%). Also, provide the 90% 

coverage interval for the total contribution estimate of these systems. 

3A.2.2. Repeat 3A.2.1 for each LOCA size category for the cold leg. 

3A.2.3. Repeat 3A.2.1 and 3A.2.2 for all other PWR and BWR LOCA-susceptible piping systems. 

Notes: 

a. The list of possible values for each variable class is provided in the Kick-off Meeting Notes

document. 

b. Please see EQ 4 in the Elicitation Question Development document for additional information 

and an example for this question. 

3A.3.1. Estimate the relative LOCA likelihood, or ratio, of each unique variable combination for 

Category 1 cold leg LOCA developed in 3A.2 to the cold leg reference case (or another suitable 

base or reference case) as a function of plant operating time (25, 40, 60 years). 

3A.3.2. Repeat 3A.3.1 for each LOCA size category for the cold leg. 

3A.3.3. Repeat 3A.3.1 and 3A.3.2 for all PWR and BWR LOCA-susceptible piping. 

Notes: 

a. Piping reference case conditions are described in the Kick-off Meeting Notes document. 

b. Please see EQ 6 in the Elicitation Question Development document for additional information 

and an example for a similar question.

Question Set 3B 

3B.1.1. List the PWR piping systems that provide a minimum of 80% of the total contribution for 

Category 1 (leak rates > 100 gpm [380 lpm]) LOCAs in US plants after 25, 40, and 60 years of 

operation. Now estimate the MV contribution of these systems (> 80%). Provide the 90% 

coverage interval for the total contribution estimate of these systems. 

3B.1.2. Repeat 3B.1.1 for Category 2 through 6 LOCAs in PWR piping systems. 

3B.1.3. Repeat 3B.1.1 and 3B.1.2 for BWR piping systems. 

Notes: 

a. Relevant BWR and PWR piping systems are described in the Kick-off Meeting Notes

document. 
J-4

b. Please see EQ 1 in the Elicitation Question Development document for additional information 

and an example for a similar question. 

3B.2.1. Estimate the percentage contribution for each PWR piping system in 3B1.1 for Category 1 

LOCAs as a function of plant operating time. 

3B.2.2. Repeat 3B.2.1 for Category 2 through 6 LOCAs in PWR piping systems. 

3B.2.3. Repeat 3B.2.1 and 3B.2.2 for BWR piping systems. 

Notes: 

a. Please see EQ 2 in the Elicitation Question Development document for additional information 

and an example for a similar question. 

3B.3. If a base case piping system(s) is listed within your significant PWR piping systems for 

Category 1 LOCAs (3B.1.1), go to 3B.5. If not, go to 3B.4. 

3B.4.1. Estimate the ratio of the reference case for the Category 1 LOCA contribution of your most 

important BWR piping system to a suitable base case as a function of plant operating time. 

Also, provide the 90% coverage range for this ratio. 

3B.4.2. Repeat 3B.3 for Category 2 through 6 LOCAs in PWR piping systems. 

3B.4.3. Repeat 3B.3 for BWR piping systems. 

Notes: 

a. Base and reference case conditions for piping systems are defined within the Kick-off Meeting 

Notes document. 

b. Please see EQ 10 in the Elicitation Question Development document for additional 

information and an example for a similar question. 

3B.5.1. Estimate the ratio of all Category 1 LOCA contributions for this piping system to those 

contributions represented by the base (or reference) case conditions as a function of plant 

operating time. Provide the 90% coverage range for this ratio. 

3B.5.2. Repeat 3B.3 for Category 2 through 6 LOCAs in PWR piping systems. 

3B.5.3. Repeat 3B.3 for BWR piping systems. 

Notes: 

a. Please see the appendix for an example for this question.

J.2.4 Elicitation Question 4: LOCA Frequencies of Non-Piping Components 

Question Set 4A 

4A.1.1. Examine the failure scenarios for each of the five PWR non-piping components (pressurizer, 

valves, pumps, RPV, and steam generator). For each component, list the failure scenarios that 

provide a minimum of 80% of the total contribution for Category 1 (leak rates > 100 gpm [380 

lpm ]) LOCAs in US plants after 25, 40, and 60 years of operation. Estimate the MV 

contribution of these failure scenarios (> 80%). Also, provide the 90% coverage interval for 

the total contribution estimate of these systems. 

4A.1.2. Repeat 4A.1.1 for Category 2 through 6 LOCAs for the non-piping PWR components. 

4A.1.3. Repeat 4A.1.1 and 4A.1.2 for BWR non-piping components (valves, pumps, RPV). 

Notes: 

a. A failure scenario is associated with a specific non-piping component, material, degradation 

mechanism, etc. 

b. Relevant BWR and PWR non-piping failure scenarios and components are discussed in the 

kick-off meeting notes document (called failure modes instead of scenarios in this document). 

These are also summarized in the elicitation summary tables. 

c. Please see EQ 1 in the Elicitation Question Development document for additional information 

and an example for a similar question. 

4A.2.1. Choose a piping or non-piping base case which results in the most natural comparison for each 

of the failure scenarios described in 4A.1.1 for all five PWR non-piping component classes. 
J-5

Provide a MV estimate of the ratio for the Category 1 LOCA contribution of the chosen nonpiping failure scenario to the chosen base case. 

4A.2.2. Repeat 4A.2.1 for Category 2 through 6 LOCAs for the non-piping PWR components. 

4A.2.3. Repeat 4A.2.1 and 4A.2.2 for BWR non-piping components (valves, pumps, RPV). 

Notes: 

a. Please see EQ 6 in the Elicitation Question Development document for additional information 

and an example for a similar question. 

b. Non-piping base cases are currently being quantified to determine the leaking frequencies due 

to all degradation mechanisms for each non-piping component listed in the kick-off meeting 

notes document. There will also be non-piping base cases frequencies for items that have 

failed such as SGTRs. Additionally, non-piping base cases can still be chosen for making 

relative comparisons. For instance if a panelist considers valve body failure due to cavitation 

erosion to be significant for Category 1 PWR LOCAs, then valve body leakage can be chosen 

as the base case. 

Question Set 4B 

4B.1.1. List the PWR non-piping failure scenarios that provide a minimum of 80% of the total 

contribution for Category 1 (leak rates > 100 gpm [380 lpm]) LOCAs in US plants after 25, 40, 

and 60 years of operation. Now estimate the MV contribution of these failure scenarios (> 

80%). Also, provide the 90% coverage interval for the total contribution estimate. 

4B.1.2. Repeat 4B.1.1 for Category 2 through 6 LOCAs for the non-piping PWR failure scenarios. 

4B.1.3. Repeat 4B.1.1 and 4B.1.2 for BWR non-piping failure scenarios. 

Notes: 

a. This question differs from Elicitation Question 4A in that only the significant failure 

scenarios, regardless of component, need to be considered. 

b. A failure scenario is associated with a specific non-piping component, material, degradation 

mechanism, etc. 

c. Relevant BWR and PWR non-piping failure scenarios are discussed in the kick-off meeting 

notes document (called failure modes instead of scenarios in this document). These are also 

summarized in the elicitation summary tables. 

d. Please see EQ 1 in the Elicitation Question Development document for additional information 

and an example for a similar question. 

4B.2.1. Estimate the percentage contribution for each PWR non-piping failure scenario in 4B1.1 for 

Category 1 LOCAs after 25, 40, and 60 years of operation. 

4B.2.2. Repeat 4B.2.1 for Category 2 through 6 LOCAs for the non-piping PWR scenarios. 

4B.2.3. Repeat 4B.2.1 and 4B.2.2 for BWR non-piping failure scenarios. 

Notes: 

a. Please see EQ 2 in the Elicitation Question Development document for additional information 

and an example for a similar question. 

4B.3.1 Pick either a piping or a non-piping base case (or a piping reference case) for comparison with 

one or more of your significant non-piping failure scenarios from 4B1.1 for Category 1 

LOCAs. The comparison should be natural, but if possible, should be made with one of the 

most significant failure scenarios that you listed. Provide a MV estimate of the ratio of the nonpiping failure scenario to the chosen base case as a function of operating time (40 and 60 

years). Also, provide the 90% coverage range for this ratio. 

4B.3.2 Repeat 4B.3.1 for Category 2 through 6 LOCAs for the non-piping PWR failure scenarios. 

4B.3.3 Repeat 4B.3.1 and 4B.3.2 for BWR non-piping failure scenarios. 

Notes: 

a. Base case conditions for piping systems and are defined within the kick-off meeting notes

document. Base case conditions for non-piping components are being developed as discussed 

in the notes to Elicitation Question 4A.2 
J-6

b. Please see EQ 6 in the Elicitation Question Development document for additional information 

and an example for a similar question. 

J.2.5 Elicitation Question 5: LOCA Probabilities of Piping Components under an 

Emergency Faulted Load 

An emergency faulted load represents an initial design consideration for a possible large transient load 

that was not expected to occur over any particular plant’s operating life of 40 years (rare event), or a 

frequency less than approximately 0.025 yr-1. These loads could be due to seismic loading or any other 

large pressure transients. Base cases have been developed which examine the conditional failure 

probability for ASME Service Level B loading. This loading level was estimated for several plants to 

conservatively approximate a 1*SSE event on a pipe which is flawed up to the allowable technical 

specification leakage rate for the given piping system and degradation mechanism. An SSE event was 

initially a design-level earthquake amplitude that was thought to occur once in 40 years; however, 

operating experience to date suggests that the frequency of an SSE event occurring is less than that. 

This question will ask you to list and quantify the effect of the most significant piping systems and 

degradation mechanisms that contribute to each LOCA category. The quantification will be done for two

emergency faulted load sizes (ASME Service Levels B and D) for three assumed damage states. The 

damage states will consist of tech. spec. leakage rates, the onset of leakage through a slow drip 

(perceptible leak), and a surface crack with a/t = 0.5. The surface crack length will be assumed by each 

panelist and may be a function of degradation mechanism and material. A relationship between the 

failure loads for a circumferential through-wall-crack and surface cracks with a/t = 0.5 and different

lengths is provided in the “Piping Seismic Base Cases” document. The likelihood of each damage state 

will also be ascertained by each panelist relative to the operational experience data for the leak-rate 

frequencies corresponding to each system listed, regardless of degradation mechanism. This assessment

will require nine different relative comparisons for each LOCA size category and plant type (BWR or 

PWR). 

The appendix of this document and the “Piping Seismic Base Cases” document provide the philosophy 

behind the seismic piping elicitation questions and detail the seismic piping base case calculations. Both 

documents should be read prior to answering this elicitation question. 

5A.1.1. List the piping systems and degradation mechanisms which most significantly contribute to 

Category 1 LOCAs given that an assumed emergency faulted load occurs with an equivalent 

magnitude of an ASME Service Level B (SLB) event for PWRs. This total list should 

summarize at least the top 80% contributing factors to Category 1 LOCAs under assumed 

faulted loading conditions. Also, for each system, list the loads which may result in SLB 

loading and indicate if they are primary (loading-controlled) or secondary (displacementcontrolled). Provide the total contribution and also the 90% coverage interval for this estimate. 

5A.1.2. Repeat 5A.1.1 for ASME Service Level D (SLD) loading 

5A.1.3. Repeat 5A.1.1 and 5A.1.2 for each PWR LOCA size category. 

5A.1.4. Repeat 5A.1.1 - 5A.1.3 for BWRs. 

Notes: 

a. Information on relevant piping systems, degradation mechanisms, and piping sizes is 

contained in the “Elicitation Meeting Notes” from the kick-off meeting. 

b. In this question, pick your piping systems assuming that the pipes will completely fail. 

Therefore, the LOCA size category will be directly related to the pipe size. 

5A.2.1. Pick a representative set of seismic base-case conditions to use for comparison for each of your 

significant contributors to Category 1 LOCAs in PWRs. 

5A.2.2. Repeat 5A.2.1 for each PWR LOCA size category. 

5A.2.3. Repeat 5A.2.1 and 5A.2.2 for BWRs. 
J-7

Notes: 

a. A PWR and BWR base case have been defined in the “Piping Seismic Base Cases” document 

for a specific degradation mechanism, pipe size, and material. Additionally, figures are 

available which show the effects of changing materials, piping size, and service level loading 

with respect to the base case definitions. 

b. Comparisons to the selected base cases will be made in subsequent questions. 

c. A relationship between the failure loads for a circumferential through-wall-crack and surface 

cracks with a/t = 0.5 and different lengths is given at the end of the Piping Seismic Base 

Case/Background section. 

5A.3.1. Consider a single piping system and degradation mechanism combination identified for 

Category 1 PWR LOCAs in 5.A.1 and the associated seismic base case identified in 5A.2. 

Determine the ratio of the CFP for this system/degradation mechanism combination (PTSL or 

PTSL@SLB) to the CFP for the chosen seismic piping base case (PBC). Assume that a SLB 

emergency faulted load occurs and that the piping system is degraded by a through-wall crack 

that is leaking at the technical specification limit. Also provide the 90% coverage interval of 

this ratio. 

5A.3.2. Consider the same piping system and degradation mechanism combination identified for 

Category 1 PWR LOCAs in 5.A.3.1. Next, determine the ratio of the CFP for a crack that has 

just formed a perceptible leak (PPL) to the CFP for a crack leaking at the technical specification 

limit assuming (PTSL) a SLB load. Also provide the 90% coverage interval of this ratio. 

5A.3.3. Again, consider a single piping system and degradation mechanism combination identified for 

Category 1 PWR LOCAs in 5.A.3.1. Next, determine the ratio of the conditional failure 

probability for a crack with a maximum depth of 50% of the wall thickness (P50) to the CFP for 

a crack that has just formed a perceptible leak (PPL) assuming a SLB load. Also provide the 

90% coverage interval of this ratio. 

5A.3.4. Repeat 5A.3.1 – 5A.3.3 for each significant piping system/degradation mechanism combination 

listed for PWR Category 1 LOCAs in 5A.1. 

5A.3.5. Repeat 5A.3.1 - 5A.3.4 for each PWR LOCA size category. 

5A.3.6. Repeat 5A.3.1 - 5A.3.5 for BWRs. 

Notes: 

a. The leaking crack size is a function of the degradation mechanism and is the major 

contributor to the differences with the base-case conditional failure probabilities. 

b. A perceptible leak is a leak which has just initiated. 

5A.4.1. Again consider a single piping system and degradation mechanism combination identified for 

Category 1 PWR LOCAs in 5.A.1. Next, determine the ratio of the CFP for a SLD event 

(PTSL@SLD) to the CFP for a SLB event (PTSL@SLB) assuming a crack leaking at the technical 

specification limit in both cases. Also provide the 90% coverage interval of this ratio. 

5A.4.2. Consider the same piping system and degradation mechanism combination identified for 

Category 1 PWR LOCAs in 5.A.4.1. Next, determine the ratio of the CFP for a crack that has 

just formed a perceptible leak (PPL) to the CFP for a crack leaking at the technical specification 

limit (PTSL or PTSL@SLD) assuming a SLD load. Also provide the 90% coverage interval of this 

ratio. 

5A.4.3. Again, consider a single piping system and degradation mechanism combination identified for 

Category 1 PWR LOCAs in 5.A.4.1. Next, determine the ratio of the conditional failure 

probability for a crack with a maximum depth of 50% of the wall thickness (P50) to the CFP for 

a crack that has just formed a perceptible leak (PPL) assuming a SLD load. Also provide the 

90% coverage interval of this ratio. 

5A.4.4. Repeat 5A.4.1 – 5A.4.3 for each significant piping system/degradation mechanism combination 

listed for PWR Category 1 LOCAs in 5A.1. 

5A.4.5. Repeat 5A.6.1 - 5A.6.4 for each PWR LOCA size category. 

5A.4.6. Repeat 5A.6.1 - 5A.6.5 for all BWRs. 
J-8

Notes: 

a. If your system and degradation mechanism list in 5A.1.2 for SLD loading is different from 

that in 5A.1.1 for SLB loading, pick a seismic base for reference in 5A.4.1 instead of 

referencing with respect to the SLB loading magnitude. 

J.2.5.1 Estimation of Piping Damage Likelihood: Now consider the relative likelihood of the 

occurrence of a particular level of damage (50% through-wall, perceptible leak, tech. spec. leakage) due 

to the piping system/degradation mechanism combination chosen in 5A.1. All answers will be ultimately

referenced to a piping base-case damage probability as in earlier questions. However, there are no 

numbers assigned to the base-case damage probabilities at this time, so the comparisons should be made

with respect to a piping base-case damage condition. A separate piping base-case condition is defined for 

each piping system and LOCA size category identified in 5A.1, as the operational experience frequency 

of all leaks regardless of the degradation mechanism. This frequency will be determined from 

operational experience data. 

5A.5.1 Consider a single piping system and degradation mechanism combination identified for 

Category 1 PWR LOCAs in 5.A.1. Next, determine the ratio of the likelihood of a pipe having 

a perceptible leak due to that degradation mechanism in that piping system (LPL) after 25 years 

of operation (LPL) to the base case (LBC), which is the likelihood of a leak due to any 

degradation mechanism. Also provide the 90% coverage interval for this estimate.

5A.5.2 Consider the same single piping system and degradation mechanism as in 5A.5.1. Next, 

determine the ratio of the likelihood of a technical specification leak (LTSL) to a perceptible leak 

(LPL) due to that degradation mechanism after 25 years of operation. Also provide the 90% 

coverage interval for this estimate.

5A.5.3 Consider the same single piping system and degradation mechanism as in 5A.5.1. Next, 

determine the ratio of the likelihood of a 50% through-wall leak (L50) to a perceptible leak (LPL) 

due to that degradation mechanism after 25 years of operation (current-day). Also provide the 

90% coverage interval for this estimate.

5A.5.4 Now determine if you believe the relative likelihood ratios in 5A.5.1 – 5A.5.3 will increase, 

decrease, or remain constant with continued operating time. First consider all three likelihood 

estimates (LPL/LBC, LTSL/LPL, and L50/LPL) at 40 years and then 60 years of continued operation. 

Determine the ratio of these estimates at 40 years of operation to the current-day estimates. 

Next, determine the ratio these estimates at 60 years of operation to current-day estimates. 

5A.5.5 Repeat 5A.5.1- 5A.5.4 for each significant piping system/degradation mechanism combination 

listed for PWR Category 1 LOCAs in 5A.1. 

5A.5.6 Repeat 5A.5.1 - 5A.5.5 for each PWR LOCA size category. 

5A.5.7 Repeat 5A.5.1 - 5A.5.6 for all BWRs.

J.2.6 Elicitation Question 6: LOCA Probabilities of Non-Piping Components under an 

Emergency Faulted Load 

An emergency faulted load represents an initial design consideration for a large transient load that was not 

expected to occur over any particular plant’s operating life of 40 years (rare event). These loads could be 

due to seismic loading or any other large pressure transients. Similar to the piping evaluation, base cases 

will be used for anchoring on the conditional failure probability. However, the actual base cases will not 

be developed until after the panelists’ identify the non-piping components which provide significant 

LOCA contributions. In the interim, each panelist should use a particular set of base case conditions for 

anchoring. More information on this selection will follow in Elicitation Question 6A.2. 

This question will ask you to list and quantify the effect of the most significant non-piping systems and 

degradation mechanisms that contribute to each LOCA category. The quantification will be done for two

emergency faulted load sizes (SLB and SLD) for three assumed damage states. The damage states will 
J-9

consist of tech. spec. leakage rates, the onset of leakage through a slow drip (perceptible leak), and a 

surface crack with a/t = 0.5. The surface crack length will be assumed by each panelist and may be a 

function of degradation mechanism and material. The likelihood of each damage state will also be 

ascertained by each panelist relative to the operational experience data for the leak-rate frequencies

corresponding to each non-piping component listed, regardless of degradation mechanism. This 

assessment will require nine different relative comparisons for each LOCA size category and plant type

(BWR or PWR). 

The structure of this question is almost identical to Elicitation Question 5. The appendix contains 

information on the philosophy behind these two questions. 

6A.1.1. List the non-piping components and degradation mechanisms (or failure scenarios) which most 

significantly contribute to Category 1 LOCAs given that an assumed emergency faulted load 

occurs with an equivalent SLB magnitude for PWRs. This total list should summarize at least 

the top 80% contributing factors to Category 1 LOCAs under assumed faulted loading 

conditions. Also, for each component, list the loads which may result in SLB loading and 

indicate if these loads are primary (load-controlled) or secondary (displacement-controlled). 

Provide the total contribution and also the 90% coverage interval for this estimate. 

6A.1.2. Repeat 6A.1.1 for Service Level D (SLD) loading 

6A.1.3. Repeat 6A.1.1 and 6A.1.2 for each PWR LOCA size category. 

6A.1.4. Repeat 6A.1.1 - 6A.1.3 for BWR non-piping components. 

Notes: 

a. Information on relevant non-piping components and degradation mechanisms, are contained 

in the “Elicitation Meeting Notes” from the kick-off meeting and subsequent revisions to 

Tables B.1.13 – B.1.17 in this document. 

b. In this question, pick your non-piping component assuming that it will completely fail. 

Therefore, the LOCA size category will be directly related to the component size. 

6A.2.1. Pick a representative set of seismic base-case conditions to use for comparison for each of your 

significant contributors to Category 1 LOCAs in PWRs. 

6A.2.2. Repeat 6A.2.1 for each PWR LOCA size category. 

6A.2.3. Repeat 6A.2.1 and 6A.2.2 for BWR non-piping components. 

Notes: 

a. The base case conditions should correspond to at least one (or several, or all) of the 

significant non-piping LOCA contributors identified in 6A.1. Assume that the component is 

damaged with a fatigue flaw which results in technical specification leakage. Assume that 

the base case loading magnitude is an SLB load. Assume that absolute size of this flaw and 

the actual conditional failure probability to a SLB load magnitude will be quantified at a later 

date. 

b. Comparisons to the selected base cases will be made in subsequent questions. 

6A.3.1. Consider a single non-piping component and degradation mechanism combination identified 

for Category 1 PWR LOCAs in 6.A.1 and the associated seismic base case identified in 6A.2. 

Determine the ratio of the CFP for this system/degradation mechanism combination (PTSL or 

PTSL@SLB) to the CFP for the chosen seismic non-piping base case assuming (PBC) that an SLB 

emergency faulted load occurs and that the non-piping component contains a through-wall 

crack that is leaking at the technical specification limit. Also provide the 90% coverage 

interval of this ratio. 

6A.3.2. Consider the same non-piping component and degradation mechanism combination identified 

for Category 1 PWR LOCAs in 6.A.3.1. Next, determine the ratio of the CFP for a crack that 

has just formed a perceptible leak (PPL) to the CFP for a crack leaking at the technical 

specification limit (PTSL) assuming a SLB load magnitude. Also provide the 90% coverage 

interval of this ratio. 
J-10

6A.3.3. Again, consider the single non-piping component and degradation mechanism combination 

identified for Category 1 PWR LOCAs in 6.A.3.1. Next, determine the ratio of the CFP for a 

crack with a maximum depth of 50% of the wall thickness (P50) to the CFP for a crack that has 

just formed a perceptible leak (PPL) assuming a SLB load. Also provide the 90% coverage 

interval of this ratio. 

6A.3.4. Repeat 6A.3.1 – 6A.3.3 for each significant non-piping component/degradation mechanism 

combination listed for PWR Category 1 LOCAs in 6A.1. 

6A.3.5. Repeat 6A.3.1 - 6A.3.4 for each PWR LOCA size category. 

6A.3.6. Repeat 6A.3.1 - 6A.3.5 for BWR non-piping components. 

Notes: 

a. The leaking crack size is a function of the degradation mechanism and is the major 

contributor to the differences with the base-case conditional failure probabilities. 

6A.4.1. Again consider a single non-piping component and degradation mechanism combination as 

identified for Category 1 PWR LOCAs in 6.A.1. Next, determine the ratio of the CFP for a 

SLD event (PTSL@SLD) to the CFP for a SLB event (PTSL@SLB). Assume that a crack exists which 

is leaking at the technical specification limit in both cases. Also provide the 90% coverage 

interval of this ratio. 

6A.4.2. Consider the same non-piping component and degradation mechanism combination identified 

for Category 1 PWR LOCAs in 6.A.4.1. Next, determine the ratio of the CFP for a crack that 

has just formed a perceptible leak (PPL) to the CFP for a crack leaking at the technical 

specification limit (PTSL or PTSL@SLD). Assume a SLD loading magnitude in each case. Also 

provide the 90% coverage interval of this ratio. 

6A.4.3. Again, consider the same non-piping component and degradation mechanism combination 

identified for Category 1 PWR LOCAs in 6.A.4.1. Next, determine the ratio of the CFP for a 

crack with a maximum depth of 50% of the wall thickness (P50) to the CFP for a crack that has 

just formed a perceptible leak (PPL). Assume a SLD loading magnitude in each case. Also 

provide the 90% coverage interval of this ratio. 

6A.4.4. Repeat 6A.4.1 – 6A.4.3 for each significant non-piping component/degradation mechanism 

combination listed for PWR Category 1 LOCAs in 5A.1. 

6A.4.5. Repeat 6A.6.1 - 6A.6.4 for each PWR LOCA size category. 

6A.4.6. Repeat 6A.6.1 - 6A.6.5 for all BWR non-piping components. 

Notes: 

a. If your system and degradation mechanism list in 6A.1.2 for SLD loading is different from 

that in 6A.1.1 for SLB loading, pick a seismic base for reference in 6A.4.1 instead of 

referencing with respect to the SLB loading magnitude. 

J.2.6.1 Estimation of Piping Damage Likelihood: Now consider the relative likelihood of the 

occurrence of a particular level of damage (50% through-wall, perceptible leak, tech. spec. leakage) due 

to the non-piping component/degradation mechanism combination chosen in 6A.1. All answers will be 

ultimately referenced to a non-piping base-case damage probability. However, there are no numbers 

assigned to the non-base-case damage probabilities at this time. Comparisons should therefore be made

with respect to a non-piping base-case damage condition. A separate non-piping base-case condition is 

defined for each non-piping component identified in 6A.1, as the operational experience frequency of all 

component leaks regardless of the degradation mechanism. 

6A.5.1 Consider a single non-piping component and degradation mechanism combination identified 

for Category 1 PWR LOCAs in 6.A.1. Next, determine the ratio of the likelihood of the nonpiping component having a perceptible leak after 25 years of operation (LPL) due to that 

degradation mechanism to the base case (LBC), which is the likelihood of a leak due to any 

degradation mechanism. Also provide the 90% coverage interval for this estimate.

6A.5.2 Consider the same non-piping component and degradation mechanism as in 6A.5.1. Next, 

determine the ratio of the likelihood of a technical specification leak (LTSL) to a perceptible leak 
J-11

(LPL) due to that degradation mechanism after 25 years of operation. Also provide the 90% 

coverage interval for this estimate.

6A.5.3 Consider the same single non-piping component and degradation mechanism as in 6A.5.1. 

Next, determine the ratio of the likelihood of a 50% through-wall leak (L50) to a perceptible 

leak (LPL) due to that degradation mechanism. Also provide the 90% coverage interval for this 

estimate.

6A.5.4 Now determine if you believe the relative likelihood ratios in 6A.5.1 – 6A.5.3 will increase, 

decrease, or remain constant with continued operating time. First consider all three likelihood 

estimates (LPL/LBC, LTSL/LPL, and L50/LPL) at 40 years and then 60 years of continued operation. 

Determine the ratio of these estimates at 40 years of operation to the current-day estimates. 

Next, determine the ratio these estimates at 60 years of operation to current-day estimates. 

6A.5.5 Repeat 6A.5.1- 6A.5.4 for each significant non-piping component/degradation mechanism 

combination listed for PWR Category 1 LOCAs in 6A.1. 

6A.5.6 Repeat 6A.5.1 - 6A.5.5 for each PWR LOCA size category. 

6A.5.7 Repeat 6A.5.1 - 6A.5.6 for all BWR non-piping components.
J-12

ATTACHMENT A TO APPENDIX J: ADDITIONAL EXAMPLE FOR

ELICITATION QUESTION 3B.5.1 

3B.5.1. Estimate the ratio of all Category 1 LOCA contributions for this piping system to those 

contributions represented by the base (or reference) case conditions as a function of plant 

operating time. Provide the 90% coverage range for this ratio. 

For question set 3B, Panelist A has previously listed the following important PWR piping systems and 

their individual contributions to Category 1 LOCAs after 25 years of plant operation (see Table J.A.1). 

For this example, this panelist does not expect the relative contributions to change after either 40 or 60 

years of plant operation. 

Table J.A.1 PWR Piping System Contributions for Category 1 LOCAs 

Case Piping System Lines Percentage 

Contribution 

1 Instrumentation 50 

2 Drain Lines 10 

3 Reactor Coolant Pressure Hot Leg 10 

4 Chemical Volume Control System 10 

5 Safety Injection System Accumulator 10 

The reactor coolant pressure hot leg has an associated base case. The base case geometry is a 30 inch

diameter pipe, manufactured from Type 304 stainless steel with an Alloy 600 safe end. The safe end to

pipe weld is a nickel-based bimetallic weld. The base case degradation mechanisms are thermal fatigue

and PWSCC. The loading is pressure, thermal, residual stress, dead weight, with a pressure pulse 

transient. Panelist A next needs to estimate the ratio between all Category 1 LOCAs in the hot leg 

compared to those represented solely by the base case conditions. His results are summarized in Table

J.A.2. 

Table J.A.2 Panelist A's Ratio Between Entire Piping System and Base Case Contributions for 

Category 1 PWR Hot Leg LOCAs 

25 Years 40 Years 60 Years Base/ref 

case 

5% LB MV 5% 

UB 

5% LB MV 5% 

UB 

5% LB MV 5% 

UB 

SL BC 1 2 3 1 3 4 1 4 5 

Panelist A believes that the total hot leg Category 1 LOCAs are twice the number represented by those 

captured by the base case conditions after 25 years of service (present day). However, this ratio increases 

with time until after 60 years, the total hot leg Category 1 LOCA probability is 4 times what is predicted 

by the base case conditions. This initial difference is due to the fact that Panelist A believes that an equal 

number of Category 1 LOCAs will be due to thermal fatigue of other piping materials than are 

represented by the base case (specifically cast stainless steel and stainless steel clad carbon). This 

panelist also believes that these mechanisms will become more apparent with time than either thermal 

fatigue or PWSCC of stainless steel material. Future increases are also included for unanticipated 

mechanisms. 
J-13

ATTACHMENT B TO APPENDIX J: PHILOSOPHY BEHIND 

ELICITATION QUESTION 5 

The conditional LOCA probability for a given piping system, degradation mechanism, and emergency 

faulted load can be determined by multiplying the likelihood curve (L, red in Figure J.B.1) by the 

conditional piping failure probability (P, black in Figure J.B.1) and then integrating over all the possible 

damage states. This conditional LOCA probability will likely be a function of LOCA size (pipe size), 

piping system, applied emergency faulted load, and degradation mechanism. Figure J.B.1 below provides

a schematic for a fixed ASME Service Level B (SLB) load, piping system (instrument lines), LOCA size 

(Category 1), and degradation mechanism (PWSCC). The curve shapes/trends in Figure J.B.1 are an 

illustration and do not represent any panelist opinion. A separate set of likelihood and conditional failure 

probability curves exists for each unique combination of these four variables. 

50% TW Perceptible Leak Tech. Spec. Leak .

Conditional Failure Probability

Damage Likelihood

P50

Ppl

Lpl

Ptsl

Ltsl

L50

Figure J.B.1 Conditional Failure Probability for Service Level B Loading for Category 1 LOCAs in 

PWR Instrument Lines due to PWSCC 

The elicitation question first asks each panelist to identify only the significant piping system and 

degradation mechanisms to conditional LOCA for each LOCA size category (EQ 5A.1). Then, the 

panelist will pick seismic base case conditions for either the PWR hot leg or BWR feedwater seismic base 

cases described in the “Seismic Base Case” document (EQ 5A.2). 

Next, the panelist will provide ratios of the relative likelihood and conditional failure probabilities (Figure 

J.B.1) among three different damage states (a 50% through-wall crack, a perceptible leak, and a technical 

specification leak) at a fixed load for each LOCA size category. All the estimates are initially for a 
J-14

Service Level B load. Here are the ratios that will be provided: Ptsl/Pbc (EQ 5.A.3.1), Ppl/Ptsl (EQ 5A.3.2), 

and P50/Ppl (EQ 5A.3.3). The Pbc value is defined in the “Seismic Base Case” document while all other 

variables are defined in Figure J.B.1. In this way, all answers are linked back to the case quantification of 

the base-case conditions. The next elicitation question (5A.4.1) asks for the relationship between the 

Service Level D and Service Level B event for the given degradation mechanism and system which is the 

ratio of Ptsl at SLD/Ptsl at SLB. The other questions ask for the ratio of Ppl/Ptsl (EQ 5A.4.2), and P50/Ppl (EQ 

5A.4.3) for solely the SLD loading event. 

Finally, the likelihood of the 3 damage states is estimated. The likelihood base case to be used for each 

piping system and degradation mechanism listed by each panelist will be the leaking frequency for that

system over the piping size range of interest for all degradation mechanisms (Lbc). Obviously, this 

frequency will be more than the frequency for any single degradation mechanism, but dominant 

degradation mechanisms may provide a ratio close to 1. This frequency has not been quantified, but it

will be after the elicitation once a complete listing of important systems and degradation mechanisms has 

been provided by the panelists. However, leaking rate propensity has been provided as part of the piping 

base case analysis for those systems (i.e., see Bill Galyean and Bengt Lydell’s results). The elicitation 

questions ask each panelist to provide the following three likelihood ratios: Lpl/Lbc (5A.5.1), Ltsl/Lpl

(5A.5.2), and L50/Lpl (5A.5.3). These variables are defined in Figure J.B.1. 

The curve is developed for the 3 damage states and 2 loads in an attempt to capture the most significant 

contributions to the conditional LOCA probability. These events will be interpolated and extrapolated as 

necessary to develop continuous relationships as a function of damage state and loading magnitude. This 

information can then be combined with a plant’s seismic hazard curve as well as knowledge about the 

relationship between the hazard curve and actual piping stresses to determine actual LOCA frequencies 

due to a seismic event. They could also be used to determine the LOCA contribution of other large 

transients knowing both the transient frequency and the relationship between the transient applied loading 

magnitude and the ASME service level loading magnitudes. 
APPENDIX K 

GENERAL APPROACH AND PHILOSOPHY 

OF EACH PANEL MEMBER

K-1

APPENDIX K 

GENERAL APPROACH AND PHILOSOPHY OF EACH PANEL MEMBER

In this appendix the general approach and philosophy that each panelist followed as part of this elicitation 

exercise is presented. 
K-2

BRUCE BISHOP 

For PWR piping frequencies, the median probability of a 5,000 gpm (19.000 lpm) leak after 40 years of 

operation comes from the average point estimate for 7 plants that used the PFM methodology for the 

WOG Piping RI-ISI (WCAP-14257, Rev. 1-NP-A, Supplement 1). These seven plants were selected to 

provide a representative sampling of all plants with a Westinghouse NSSS design. Characteristics 

considered in the sampling included number of primary loops, old and new design vintage and foreign 

and domestic utility operators. The variability in 40-year probability with leak-rate comes from a WOG 

supported sensitivity study that reflected both the decrease in probability with increasing leak rate of one 

pipe size and the number of pipes of a given size that could contribute to a given leak rate. All piping 

leak probabilities consider the effects of LBB with a minimum detectable leak of 1 gpm (3.8 lpm) per 

typical plant tech-spec requirements. The increase in failure probability in going from 40 years to 60 

years of operation is based upon another WOG sensitivity study. This study and its results are described 

in a paper presented at the 1999 Pressure Vessel and Piping Conference of ASME and included in PVPVolume 383. 

Non-Piping Frequencies are based upon the degradation mechanism of PWSCC initiation and throughwall growth, which is currently the primary cause of unexpected leaks in non-piping components in the 

primary system. Most other degradation mechanisms are being effectively mitigated. The relative 

frequencies by component type are based upon a proprietary best-estimate of PWSCC susceptibility by 

Westinghouse experts for unmitigated Alloy 600/182 base/weld metal. The uncertainties are based upon 

the variability between the best-estimate susceptibility for PWSCC and observed leak experience. 
K-3

VIC CHAPMAN 

In order to derive a basic set of failure probabilities, the values generated by the ‘Base Cases’ analysis 

were initially considered. However, in the end, a decision was made to use the results from some 

previous work that involved a ‘Risk-Informed ISI’ application. That work considered a full plant 

assessment using fatigue as the basic degradation mechanism. Initially, the results from this full plant 

assessment were compared with the appropriate base cases in order to ascertain whether they were in 

general agreement with each other. Once it was decided that the two sets of results were in agreement, it 

was decided to proceed with using the full plant assessment results. These results provided a set of pipe 

weld failures over a full range of pipe weld sizes that could be considered as a form of global values for 

each weld size. Factoring would then be from this base set. 

Since the leak rate, given a failure, is independent of the failure probability, this can be evaluated

separately to obtain a conditional probability. The basic method developed for the base case was 

expanded to include lower and upper estimates at each step. These basic steps are as follows: 

1 Use expert judgment to estimate the COD, up to full rupture, as a function of defect size. 

2 Evaluate the defect cross-sectional area for a given defect size using its associated COD. 

3 Evaluate the leak rate from a given defect size using some data supplied by the USNRC. 

4 Use expert judgment to assess the distribution of the defect length at failure. 

5 Combine Steps 3 and 4 to obtain the conditional probability of a leak rate greater than the prescribed

leak rate. 

The final probabilities were obtained by combining the conditional probability above with the basic 

fatigue failure probability. 

The effect of leak detection was introduced via a factor that was a function of the leak rate. This 

reduction factor varied from about 5 for a Category 1 leak, up to about 50 for a Category 6 leak. 

For non-weld areas, such as the pump bowls and nozzle crotch corners, the basic probabilities were first 

factored. Next, the basic steps to derive the conditional leak rates as discussed above were followed to 

adjust the distributions as appropriate. 

The effect of PWSCC was introduced as a multiplying factor on the basic fatigue failure rates. It was

assumed that for small pipes, 2 inch diameter, that they would still have a significant contribution from 

fatigue, but that for the largest pipes, the full three orders of magnitude implied by the PWR-1 Base Case 

(i.e., hot leg base case) should be applied. 

Finally, the failure rates for each system were derived by simply summing over all the elements within a 

given system. 
K-4

 WILLIAM (BILL) GALYEAN 

The approach taken by Bill Galyean is based on the total operating experience of U.S. commercial nuclear 

power plants. This experience consists of approximately 2,650 LWR-years with zero category-1 (> 100 

gpm [380 lpm]) loss of coolant accidents. The average age of these plants is approximately 25 years, with 

a number of plants being 30-plus years old. During this time, a number of RCS degradation issues have

arisen and been addressed, for example, IGSCC in BWRs and thermal fatigue in PWRs. The operating 

experience therefore indicates that degradation will occur, but it will likely be detected and corrected 

before it can lead to a catastrophic failure. Consequently, this data is the basis for estimating an average 

LOCA frequency using a Bayesian update of a non-informative prior distribution. Since both PWRs and 

BWRs have zero LOCAs, the reasonable assumption is that the two designs share a similar LOCA 

frequency. The operating experience for the two designs is therefore pooled (i.e., use zero failures and 

the total 2,650 LWR-years of experience). Assuming the LOCA frequency has been (and will be) 

relatively constant over time (again, this seems reasonable given the history of degradation mechanisms 

being detected and subsequently mitigated), the resulting LOCA frequency of 1.9E-4/LWR-year produces 

a probability of one or more LOCA events in the 2,650 LWR years of experience of 39% (again not 

unreasonable, given there have been zero LOCA events). By contrast, separating the PWR and BWR 

experience and analyzing them separately produces LOCA frequencies of 2.8E-4/PWR-yr and 5.6E4/BWR-yr, and a probability of seeing one or more LOCA events (either PWR or BWR) in 2,650 LWRyears experience of 63%. Again, given that there have been no LOCA events, the first (pooled) estimate 

seems to be the more realistic. 

This assumed LOCA frequency (1.9E-4/LWR-yr) was used for the category 1 LOCA (> 100 gpm [380 

lpm]). Note that as defined in the elicitation effort, category 1 LOCA includes all larger size categories. 

So the approach followed by this panel member was to assume a 1⁄2 order of magnitude reduction in 

frequency for each next larger size category. This general approach (if not the precise value of the 

reduction) has been followed by virtually every LOCA frequency estimate ever made, and is supported by

studies on precursor events documented in NUREG/CR-5750, Appendix J. 

The time-independent assumption for the LOCA frequency is also based upon the historical experience, if 

only qualitatively. There seems to be no doubt that the LOCA frequency fluctuates over the age of the

plant, but there is reason to believe it will both increase and decrease over time. The IGSCC experience 

seems to support the assertion that times of increasing frequency will be followed by times of decreasing 

frequency as degradation mechanisms are identified, understood, and mitigated. Indeed, even the recent 

RPV-head corrosion event at Davis Besse supports this model of a LOCA frequency increase as 

degradation occurs undetected, then a decrease as mitigation programs are implemented (e.g., in the case 

of Davis Besse, replacing RPV head). 

The last issue to be addressed is the allocation of the total LOCA frequency among the systems and 

components that compose the RCS. This aspect again relies upon operating experience data, this time in 

the form of the relative frequency of crack and leak events (i.e., precursor events). Basically, these 

precursor data were collected from LER and foreign reactor experience, and then sorted by degradation 

mechanism and RCS subsystem/component. In many cases the information provided on the precursor 

event was somewhat unclear or incomplete. Also, there is little assurance that all precursor events have 

been captured. However, assuming there is no bias in the reporting of the events such that the data 

samples for each subsystem/component are equally representative of the all events for that 

subsystem/component, then the data can be used to support estimates of the relative contribution from 

each subsystem/component. That is, the precursor events do not have to be completely reported, just 

consistently reported. Further, the RCS subsystem/component boundaries have not been clearly defined. 

Hence, the relative contributions to the overall LOCA frequency would likely change somewhat if the 

precursor data were reviewed and categorized by a different analyst. Nevertheless, this aspect of the
K-5

analysis was performed simply to allocate the total LOCA frequency (described above) to the general 

subsystems/components that make-up the RCS. 

In summary, the entire U.S. LWR operating experience is used to estimate an average industry-wide total 

LOCA frequency. This frequency is used for both BWRs and PWRs, not because they are believed to be 

the same, but on the basis that the operating experience does not support different frequencies. Timeindependence is assumed using the rationale that variation (both increases and decreases) in the frequency 

will occur as degradation mechanisms manifest themselves and are subsequently addressed by the 

industry. This total LOCA frequency is allocated by LOCA size categories using the argument that as 

pipe-size increases, the LOCA frequency decreases. This argument is supported by a number of studies 

on precursor data and if nothing else has been reflected in all LOCA frequency estimates since WASH1400 (1975). The total LOCA frequency is also allocated by RCS subsystem/component using data 

collected on primary system crack and leak events (although the details of this allocation are view as

somewhat subjective with respect to the boundaries of the different subsystems/components). 
K-6

KAREN GOTT 

The approach taken by Karen Gott to the elicitation was to first consider how her experience from the 

Swedish nuclear fleet was applicable to the US fleet of nuclear power stations. In this respect she took 

into account the known histories of the various degradation mechanisms which have troubled the two 

fleets as well as the mitigation methods which have been developed. This led her to amongst other things 

to the conclusion that the likelihood of an unexpected mechanism leading to failure is probably larger

than the likelihood of a known mechanism resulting in failure in a region which is inspected on a regular 

basis. In general a new area of concern with regard to component degradation has arisen on a seven to ten 

year cycle over the lifetime of commercial nuclear plants. 

To produce the numbers she used her database of failures and degradation in mechanical components in 

Swedish plants. The degradation mechanisms are the same, but the numerical figures are different 

because of differences in design and construction. She based her elicitation figures on the number of leaks 

in proportion to the number of reported cases (many were detected early) and took these to be the current 

figures for a good safety culture situation. The database includes other mechanical components than 

pipes, but does not cover steam generator tubes, so she was able to generate figures for pump and valve 

housings, for example. She then considered the differences in the philosophies concerning qualification 

and application of inspection programmes between the two countries. This she incorporated into her 

thinking about the safety culture aspects, both for the current time and for the extrapolation to 40 and 60 

years. 
K-7

DAVE HARRIS 

For each plant type and for piping and non-piping Dr. Harris selected a reference system and attempted to 

scale other systems relative to that reference system. He tried to use estimates based on operating 

experience to the maximum extent, and then scaled the relative frequencies for the LOCA categories 

using results from the PFM analyses. In many instances, operating experience was not applicable, so he 

then relied more on the PFM results. If he felt that a given system was not a significant contributor to the 

leak frequency for a given LOCA category, then he was less concerned about the accuracy of the 

frequency estimate for that system. 

PWR Non-seismic LOCA: For the PWR case, he used the hot leg as the reference system for large leak 

flow rates. Operating experience is not readily applicable. The PFM analyses for the hot leg showed a 

very wide range of results depending on the assumptions and input to the analyses. Therefore, he scaled 

the hot leg results by use of the RPV reference case results. Results presented at the wrap-up meeting in 

February 2004 provided an estimate of the RPV > 500,000 gpm (1,900,000 lpm) as 10-10 (per plant-year) 

for the first 25 years. He doubled this value to account for 2 hot legs. He assumed that the leak frequency 

for 100 gpm (380 lpm) LOCA is 3 1⁄2 orders of magnitude higher, and then interpolated on a log-log scale. 

This fixes the frequency-leak rate for the hot leg at 25 years. He assumed that the frequency for 

> 500,000 gpm (1,900,000 lpm) in the time increment 25-40 years is twice that for 0-25 years, and four

times as large in the increment 40-60 years. The leak frequency for 100 gpm (380 lpm) is assumed to be 

independent of time. 

The cold leg is then assumed to have frequencies 1/3 those of the hot leg, because the cold leg operates at 

a somewhat lower temperature. The surge line is assumed to have leak frequencies 100 times as large as 

the hot leg, because the surge line sees a lot more cycles than the hot leg, and is just as hot. These 

estimates then define the very large leak frequencies for the entire plant. 

At the low end of the leak rate scale, he assumed the plant results to be bounded by the past operating 

experience for steam generator tubes. An estimate from the wrap-up meeting for the steam generator 

LOCA frequencies is 3.5x10-3 per plant-year. 

He used the HPI make-up nozzle as a surrogate for all 2 to 6 inch diameter lines. He used the reference 

case results from the PFM results that was presented at the wrap-up meeting, but reduced the leak 

frequencies by an order of magnitude at 5,000 gpm (19,000 lpm). He assumed that the SIS accumulator 

and RHR systems have about an order of magnitude less contribution than the surge line, so they have a

small contribution to the overall plant. 

 

This procedure provides his best estimate. The 5% and 95% estimates are scaled up and down from the 

best estimate. He estimated the 5% to be 1 1⁄2 orders of magnitude below the best estimate (multiply by

0.03), independent of time and leak rate. He varied the multiplier for the 95% estimate, making it larger 

for the larger leak categories. The multiplier varied from 30 to 1000. He believes that we have a better 

handle on the smaller leak rates, because they are bounded by steam generator tubing experience, which is 

plentiful. 

BWR Non-seismic LOCA: He selected the recirculation system for the reference for BWRs. For 

intermediate leak rates 100 to 25,000 gpm (380 to 95,000 lpm), the 12 inch diameter portion of the 

recirculation system dominates. He used the base case results from the wrap-up meeting, but reduced 

them by an order of magnitude, because his PFM analysis underestimated the benefit of the post-remedial 

action residual stress. 
K-8

The feedwater system is also important, because it has lots of welds, and is prone to FAC (which is not 

related to welds). Since this is a dominant system, he assumed it to be comparable to the surge line in a 

PWR (which is a dominant system for that type of plant). The steam line is about the same size and same 

material as the feedwater, but is not prone to FAC, therefore he assumed the steam line to have leak 

frequencies that are two orders of magnitude below the feedwater system. He assumed the RHR line to 

be the same as the PWR surge line, which is about the same size. NUREG/CR-6674 shows very low 

probabilities of through-wall cracks in the HPCS/LPCS system, so the contribution of this system was 

assumed to be negligible. The recirculation, feedwater, steam line and RHR are assumed to be the 

dominant systems, and no estimates were made for other systems. 

The estimated uncertainty bands are generally tighter than the PWR estimates, because they are based 

more on experience for the dominant system (recirculation). They are independent of time, but do vary

somewhat with leak category. 

Non-piping: Dr. Harris felt less confident making estimates for non-piping components, because most of 

his experience is related to piping. He relied heavily on results provided by others in the wrap-up 

meeting, and used CRDM nozzle PWSCC, rector pressure vessel and steam generator tubing data for 

reference purposes. He also scaled relative to piping in many instances. He did not estimate time 

dependencies. For instance, for pumps and valves, he figured that they are less failure prone than the 

piping system in which they are located (passive failures). He estimated probabilities, and then calculated 

relative contributions of failure scenarios. 
K-9

BENGT LYDELL 

The approach is documented in “Base Case Report No. 2.” For systems other than the five Base Case 

Systems, the base case results established anchor distributions for BWR and PWR Code Class 1 reference

piping systems. As an example, the base case results for PWR hot legs were applied to PWR cold legs 

but adjusted to account for insights about the service conditions and degradation susceptibility specific to 

cold legs. For the other BWR and PWR piping systems not covered by the base case study, the base case

results were adjusted downwards or upwards as appropriate by accounting for unique piping design 

features (e.g., size, material, and weld population), service conditions and field experience. For nonpiping passive components the base case report again was used as the main reference (or source of 

calibration parameters) in combination with reviews of relevant operating experience. In summary, this 

Panel Member’s response to the elicitation questions is based on insights from degradation mechanism 

analyses in combination with reviews and statistical evaluations of operating experience. 
K-10

SAM RANGANATH 

One of the challenges in the LOCA frequency estimation is trying to predict the probability of an event 

that has never happened before, but which has enormous consequences if it did. It is important to 

maintain a sense of balance in this effort and aim for a realistic approach that is based strictly on technical 

considerations. As in any probabilistic analysis, the success depends on how realistic the inputs are and 

how the approach reflects actual field experience. Having worked in the BWR industry for almost 30 

years, Dr. Ranganath felt that his most important contribution to the elicitation process was to make sure 

that that frequency estimates reflect BWR field experience. For example, use of probabilistic defect 

distribution data is acceptable as long as the prediction is consistent with actual field behavior. Dr. 

Ranganath’s philosophy was to start from actual field data and to predict future behavior based on his

understanding of failure mechanisms, mitigation measures and BWR systems design. Since his 

knowledge is mainly on BWR systems, he focused his attention on BWRs rather than PWRs. He did not 

want to speculate in areas where he did not necessarily have the expertise. There are other people who 

are more knowledgeable about PWRs and they can do a better job on the estimates. He felt that the 

diversity of the elicitation panel and their expertise and the open mindedness of the NRC team helped in 

coming up with the best estimates. 
K-11

PETE RICCARDELLA 

The first step in the expert panel elicitation was to develop an amalgamated set of base case LOCA 

frequencies upon which the elicitation responses are anchored. The generic base case LOCA frequencies

developed for the panel represented the work of four teams: two teams used an empirical approach based

on operating plant experience with leakage and other precursor events, while two other teams used 

theoretical, PFM analyses. Each of these approaches has different strengths and weaknesses, such that a 

better estimate of base case LOCA probabilities can be achieved by selectively combining the results in a 

manner that optimizes the strengths of both. The method and rationale for combining the base case 

results of the individual teams were documented, ultimately producing a revised set of LOCA frequencies 

for the five piping base cases. 

LOCA frequencies for each of the LOCA sensitive piping systems identified for PWRs and BWRs were 

then estimated. This was done by picking the base case which is most representative of the specific 

LOCA sensitive system, considering plant type, material of construction, operating conditions and 

relevant degradation mechanisms, and then scaling the base case frequencies for each LOCA category 

based on judgment of any substantive differences between the base case and the system under evaluation. 

One of the main factors accounted for in this scaling process was differences in the size of the systems, in 

terms of number of welds of various pipe sizes (based on a system-by-system weld census that was 

provided to the panel). Scaling was also used to account for system specific factors, such as whether

repairs or mitigation have been applied to address degradation mechanisms considered in the base cases, 

and the timing of such actions. An estimate of the probability of breaks in small diameter socket welded 

piping (instrument, vent and drain lines) due to vibration fatigue was made, which was not included in

any of the base cases. This estimate was based on prior experience with this relatively common failure 

mechanism. 

It was felt that there is a relatively large uncertainty band in all of the above probability estimates; plus or 

minus an order of magnitude. Included in this uncertainty band is the potential development of new, as 

yet unseen degradation mechanisms in the future, which obviously weren’t considered in the base cases.

A set of base cases for non-piping LOCAs was developed, the methodology for which is documented in 

Appendix I to this NUREG report. These base cases included potential breaks due to small vessel 

penetrations such as CRD nozzles, medium size breaks due to larger diameter nozzles (excluding safe-end 

ruptures which are included in the piping base cases), and very large breaks due to pressure vessel 

ruptures (specifically addressing irradiation embrittlement of the RPV). The resulting base cases were 

then used to estimate contributions to LOCA frequency from non-piping LOCAs. 

The detailed rationale used in developing the elicitation response for each system was documented in a

report, to permit the reconstruction of the logic in the future if it becomes necessary. 
K-12

HELMUT SCHULZ 

The general approach and philosophy used follows the approached taken by GRS to estimate frequency of 

LOCA initiating events at passive systems for German PSAs. 

The major steps and assumptions of this approach are as follows: 

- In principle, wall penetration of pipes which would result into a leak follows in their geometries either 

• a slit type penetration originating from cracks caused by fatigue or corrosion or 

• a bulging type penetration caused by wall thinning.

Beyond critical dimensions wall penetrating stable defects turn into a full break. This means in practice 

that for each pipe size there are two or more leak sizes which are of a distinct different probability of 

occurrence governed by the likelihood of the respective active failure mechanism and the reliability to 

detect leaks and to take actions to avoid aggravation of the situation e. g. isolation of the leak, stop 

operation. 

The maximum leak size related to a wall penetrating stable defect (undercritical crack, bulge, pit) depends 

on the actual load specifically the relationship between membrane and bending stresses. The majority of 

systems being considered in the safety analysis of NPP's fall either into the category of high pressure or 

low pressure systems. For reasons of simplicity UB values can be taken to describe maximum leak sizes

connected to wall penetrating stable defects for each pipe size. Based on experimental evidence as well as 

fracture mechanics calculations the maximum leak size resulting from an undercritical crack is rather 

limited, expressed in terms of fractions of the pipe cross section it is only a few percent. This approach 

uses 2% of the cross section as a rule of the thumb for high pressure systems. Through wall corrosion pits 

are generally very small. Bulge-type wall penetrations caused by wall thinning have a potential for stable 

leaks of a considerable size. 

- The frequency of leaks is estimated based on the operating experience of the national population of 

nuclear power plants and in addition the worldwide experience is considered as far as applicable and 

available. In general, the operating experience give indications of leaks in a sense of precursors for most 

classes of piping or give indications of zero failures statistics only. 

- To estimate the probability of a break (which is connected by the diameter of the piping to the flow

rate) a relationship is used to describe the frequency of breaks in relation to the frequency of leaks as the 

function of the diameter of piping systems being designed to the same design parameters. For the small

bore piping (less than 2 inch) the relationship between leak and break is arrived from operating 

experience. For the largest pipe (main primary pipe) the relationship between leak and break is based on a 

number of technical arguments and PFM analyses. For the pipe sizes in between a linear relationship is

used between the UB and LB as described before. 

- For reasons of simplicity and in accordance with technical experience it is assumed that within the 

piping systems only so called leak relevant elements are contributing to the frequencies. These leak 

relevant elements are essentially welds which are adjacent to changes in geometry (nozzles, branches, 

reducers etc.) which in itself would introduce enhanced loads and to some extend represent more difficult 

areas for manufacturing and inspection. 

- The whole population of pipes, nozzles and penetrations connected to the main components are 

divided into subpopulations taking pipe diameters as orientation values, using e. g. 5 or 6 subpopulations 

to represent the difference pipe sizes, materials and operating conditions. For each subpopulation the

frequency of leaks is based according to the procedure described before (operating experience, zero 
K-13

failure statistics), the frequency of leaks is adjusted to the size of the relevant population each time and 

the frequency of breaks within the subpopulation is estimated using the described relationship. 

- The frequency of the different subpopulation which could contribute in a different way (leak or break) 

to the specified LOCA classes is then summed up. In view of the limitation regarding the verification of 

very low values of estimated frequencies a cut-off value is used. 
K-14

FRED SIMONEN 

Operating experience was applied as the best method for estimating frequencies for more common failure

events such as small detectable leakage and of ruptures for small diameter piping. Operating experience 

has the advantage of reflecting contributions from all degradation mechanisms and is not limited to a 

particular mechanism that can be addressed by a PFM model. For lower frequency events, for which 

there are little or no data from operating experience, the data were therefore supplemented by trends from 

PFM models. The fracture mechanics models were taken to provide relative frequencies such as for (for a 

given pipe size) the ratios of frequencies of for different categories of failures (in terms of leak-rates). 

Similarly, models can indicate ratios for one failure category of leak for differing pipe sizes. Reports of 

small detectable leakage (from data bases) were taken to be precursor events that can be used to calibrate 

estimates of frequencies for categories of larger leakage events. Another consideration was that only the 

larger pipe sizes contribute to the frequencies for larger leak rates. It was implicitly assumed that

contributions from smaller pipe sizes dominate for the smaller categories of leak rate frequencies. 

Non-nuclear experience was also used to support estimates of failure frequencies for nuclear components. 

Component designs, materials, construction codes, operating conditions etc. are much the same for 

nuclear applications and as for many non-nuclear applications. Non-nuclear experience however provides 

a much larger number of years of plant operation (by orders of magnitudes) than available from nuclear

experience. Non-nuclear experience therefore provides additional justification for very low failure 

frequencies for components such as pump bodies, tube sheets, manways, etc. that imply large 

extrapolations from the limited years of nuclear plant operation. 
K-15

GERY WILKOWSKI 

The piping non-seismic LOCA evaluations were conducted for PWR and BWR piping separately from a 

bottom-up approach using reference cases for certain pipe systems in each type of plant. The reference 

cases were determined from a combination of the base case results supplied to the elicitation panel. The 

base cases supplied consisted of two independent PFM analyses, and two independent operatingexperience evaluations for certain pipe systems. The probabilistic pipe fracture mechanics analyses 

(PRAISE or PRODIGAL) base cases were not chosen since Dr. Wilkowski did not believe that those 

computer codes properly determined the probability of a long surface crack occurring, which is the actual 

way that a LOCA would occur, i.e., a through-wall leaking crack will be readily discovered by leakage 

before failing at normal operating conditions. Consequently, the two historical base cases were averaged, 

but only up to 25 years of operation (current time period). Dr. Wilkowski did not believe that the 

historical based approaches would be that good for extending the LOCA reference cases to 40 or 60 years. 

Consequently, his reference cases were only for 25-year time period (present), and the 40- and 60-year

evaluations were adjusted depending if he thought the particular pipe system would be susceptible to 

some near term or long-term degradation mechanism (e.g., PWSCC), and if that mechanism could 

produce a large surface crack. These evaluations were done for 12 different PWR pipe systems and 13 

different BWR pipe systems, with six different LOCA flow-rate categories. The uncertainties (5 and 95

percent bounds) in the predictions generally increased as the amount of time increased, i.e., the 

uncertainty for 25 years was less than 40 years, and the uncertainty in the 40-year predictions was less 

than for the 60-year predictions. 

For non-piping, Dr Wilkowski felt he did not know enough about failure modes of all the different 

categories of non-piping components (with the exception of a few categories like CRDM nozzle ejection). 

He therefore chose the steam generator tube historical failure frequencies for small LOCA as a controlling 

PWR case, but all the other cases were governed by the piping failure probabilities. 

APPENDIX L 

DETAILED RESULTS 

L-1

APPENDIX L 

DETAILED RESULTS 

The detailed results from the elicitation procedure are presented in this appendix. Each panelist’s 

quantitative elicitation responses can be found through the “Electronic Reading Room” link on the NRC’s 

public website (http://www.nrc.gov/) using ADAMS. The documents are found in ADAMS using the 

following accession number: ML080560005. Both the quantitative (i.e., numerical LOCA frequency 

estimates) and qualitative results (i.e., rationale) are presented in this appendix. The quantitative results 

are often presented in the form of box and whisker plots. Box and whisker plots (often referred to simply 

as box plots) are a graphical representation of a data set. A box plot is typically based on five data points: 

the minimum and maximum of the data set plus the median, lower quartile (LQ), and upper quartile (UQ) 

associated with the data set. The LQ and the UQ are the 25th and 75th percentiles, respectively, of the data 

set. 

An example box plot is shown in Figure L.1. The region defined by the LQ and UQ is drawn as a shaded 

box with a vertical line through the median. In this appendix, the 10th and 90th percentiles are also 

indicated in the box plots with short vertical lines. Note, for the plotting program used to generate these 

box plots, the vertical lines (representative of the 10th and 90th percentiles) and the associated horizontal 

connecting them only appear when there was at minimum of nine points in the data set being plotted. 

Finally, the box plots are also overlaid with a horizontal scatter plot of the data set, with the left-most 

point being the minimum value and the right-most point being the maximum value in the data set. Note 

that the 10th and 90th percentiles are not necessarily points in the data set. The range in the data set 

encompassed by the shaded box (i.e., the range between the 25th and 75th percentiles) is referred to as the 

interquartile range (IQR) of the box plot. In the box plots in this appendix, letter designators are often 

included for the minimum and maximum values (e.g., the letters G and D in the example shown in Figure 

L.1). The letters designate the code for the panelist whose estimated value was either the lowest or the 

highest value of all the panelists who provided responses for the quantities in the data set. 
L-2

Data Array Values

0 10 20 30 40 50 60

G D

Figure L.1. Example “Box and Whisker” Plot 

Generally, the source of the qualitative responses (i.e., the rationale) came from the individual elicitations 

although there were some opinions expressed during the various plenary panel meetings that were also 

included. For each of the individual elicitations, minutes were taken. Minutes were also taken at each of 

the plenary panel meetings (see Appendix B). In addition, the participants often provided a handout to 

lead the discussion at their individual elicitations. After each elicitation, most of the participants also 

provided formal written responses to the elicitation questions. It was from these minutes, handouts, and 

written responses that the rationale provided below was extracted. Finally, each of the elicitations was 

audio taped and each meeting was video taped to provide a permanent record of the proceedings. 

Most of the panelists believe that precursor events (e.g., cracks and leaks) are a good barometer of LOCA 

susceptibility. This is reflected in the fact that almost all the panelists anchored their responses against 

some form of the available operational experience data. A distinct advantage of the operational 

experience data is its inclusion of all degradation mechanisms which have emerged to date, while the 

PFM approaches only address selected degradation mechanisms. The advantage of the PFM approaches 

is that they are best suited for addressing LOCA size and operating time effects. A number of panelists 

used the PFM results as a basis for adjusting the operational experience data in this manner. 

A major assumption made in the elicitation procedure is that all components, both piping and non-piping, 

were fabricated in accordance with applicable code standards, e.g., there were no counterfeit bolts used. 

L.1 Safety Culture

Figures L.2 and L.3 show the effect of the industry and regulatory safety culture, respectively, on the 

LOCA Ratio (i.e., the ratio of the LOCA frequency in the future to the LOCA frequency at 25 years) for

Category 1 LOCAs. Figures L.4 and L.5 show the effect of industry and regulatory safety culture on the 
L-3

LOCA Ratio for Category 4 LOCAs. Ratios less than 1.0 are indicative of a perceived reduction in the 

LOCA frequency as a result of improvements in the safety culture mindset. As can be seen in these 

figures, the panel members overwhelming expected the safety culture to either improve or remain 

constant over the next ten to fifteen years and beyond. The panel felt that the industry as whole was

acting in a consistent manner. However, a few plants with a less diligent safety culture mindset would 

provide the greatest challenge from a LOCA perspective. It was thought that these outlier plants may not 

affect the mean trends, but could strongly influence the bounds. The Davis-Besse experience was 

frequently cited as an example of this effect. The panel also expressed the opinion that the industry and 

regulatory safety culture are highly positive correlated. Therefore, regulatory and industry changes are 

expected to occur virtually simultaneous. 

LOCA Ratio: Future/Current

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5

C I

C G

60/25 year mid-value ratios

40/25 year mid-value ratios

Median values 1.0 for both 40/25 and 60/25 year ratios

Figure L.2 Effect of Utility Safety Culture on Category 1 LOCAs 
L-4

LOCA Ratio: Future/Current

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5

60/25 year ratios

40/25 year ratios

C

H

I

K

Figure L.3 Effect of Regulatory Safety Culture on Category 1 LOCAs 

LOCA Ratio: Future/Current

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5

60/25 year ratios

40/25 year ratios

Median values 1.0 for both 40/25 and 60/25 year ratios

D E

D E,G

Figure L.4 Effect of Utility Safety Culture on Category 4 LOCAs 
L-5

LOCA Ratios: Future/Current

0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5

60/25 years

40/25 years

D

E

H E

Figure L.5 Effect of Regulatory Safety Culture on Category 4 LOCAs 

As can be seen in comparing Figures L.2 with L.4 and L.3 with L.5, the panel members felt that any 

improvements in safety culture would be more beneficial for the smaller LOCA categories than their 

larger counterparts because the smaller LOCA categories constitute the bulk of the experience base. The 

frequency of the larger LOCA categories due to safety culture effects is expected to remain relatively

constant over time. 

The bottom line from this discussion is that because the panel members felt that the effect of safety 

culture was relatively minor, the LOCA frequencies developed during this exercise were not 

modified to account for this effect. The main caveat to this general conclusion is the previously 

mentioned concern that the LOCA frequencies developed through the elicitation process could be 

significantly degraded by a safety-deficient plant operating philosophy. The other concern frequently

expressed was that the industry safety culture mindset may deteriorate near the end of a plant’s license as 

management tries to “squeeze out” the final few years of operations without investing in the necessary

maintenance activities. Also, near the end of the plant’s license there was a concern expressed that the 

morale of the plant’s operating staff may begin to erode as they foresee a potential loss of employment. 

These concerns are manifested in the higher LOCA Ratios for the 60/25 year results when compared with 

the 40/25 year results in Figures L.2 through L.5. 

L.2 BWR Piping

The participants generally thought that the important degradation mechanisms for BWR piping were 

thermal fatigue, FAC, and IGSCC. It was argued that BWR plants are more prone to thermal fatigue 

problems than the PWRs because they experience a greater temperature fluctuation during the normal 

operating cycle. In BWRs, thermal fatigue is a concern for the feedwater lines, the main steam lines, and 

the RHR system. From a LOCA perspective, thermal fatigue is an important aging mechanism because it 

does not manifest itself as a single crack, but as a family of cracks over a wide area. As such, it can lead 
L-6

to a large LOCA. Thermal fatigue cracks also tend to propagate rapidly, and since it is not material 

sensitive (i.e., it can attack a number of materials), it is difficult to prioritize critical areas for inspections. 

Only the feedwater piping system is highly susceptible to FAC. The main steam line is the other major

carbon-steel piping system which experiences constant fluid flow. However, it is not as susceptible to 

FAC because the erosion rates associated with two-phase flow are less severe. While FAC caused a 

serious accident in the secondary side piping at Surry 15 years ago, the panel members generally thought 

that the industry had inspection programs in place today to prevent the reoccurrence of such an event,

especially for the primary side piping systems. However, a number of panel members expressed the 

concern that the water chemistry improvements which mitigate IGSCC could lead to unexpected FAC 

problems. 

The panel consensus is that the susceptibility to IGSCC is greatly reduced compared to the past. 

Measures such as improved HWC, weld overlay repairs, and pipe replacement with more crack resistant 

materials had essentially reduced the likelihood of IGSCC. However, there is still residual concern about 

the failure likelihood of the large recirculation piping material that has not been replaced. Furthermore, 

even for the pipe which has been replaced, the question was raised as to whether the new replaced pipe

was immune to this type of degradation, or is the problem simply been move out into the future. The 

German experience with Type 347 stainless steel was raised in this regard. There was also concern 

expressed about the effects of increased sulfate levels in the future due to efforts focused at extending the 

life of some of the filters in the plants. 

Another aging mechanism of concern is mechanical fatigue. This is primarily a problem in smaller 

diameter piping, especially those with socket welds, and is caused by an adjacent vibration source. From 

a LOCA perspective, it was noted that locations susceptible to mechanical fatigue damage were not 

always obvious. It is impossible to eliminate all plant vibrations, and furthermore, changing the 

configuration of the plant can result in newly susceptible areas. 

As part of this elicitation exercise a total of 14 LOCA-susceptible piping systems were considered for the 

BWR plants. Of these, however, most of the participants focused on a few common systems as being the 

important LOCA contributors. Figure L.6 shows the Category 1 LOCA frequencies for each of these 

piping systems at 25 years of plant operation (present day). Note, the results for the HPCS and LPCS 

systems are combined as a single entry in Figure L.6 (HPCS/LPCS). For these smaller category LOCAs, 

the main concern is with the smaller diameter lines, such as the instrument and drain lines. Most of the 

participants believe that it is more likely to have a complete break of a smaller diameter line than a

comparable size opening in a larger diameter pipe. One reason for this is that for a given crack size, the 

crack is a larger percentage of the pipe circumference in the smaller diameter pipes, and it was thought 

that a small diameter pipe was just as likely to have a crack of a certain length as a larger diameter pipe. 

Furthermore, smaller diameter lines are often fabricated from socket welded pipe which has a history of 

mechanical fatigue damage from plant vibrations. These lines may also be susceptible to external failure 

mechanisms arising from human error (e.g., damaging with equipment, such as fork trucks). Finally, 

these smaller diameter lines are often subject to fabrication flaws and they are typically more difficult to 

inspect, if they are inspected at all. In-service inspection is not routinely performed on these lines. 

Conversely, the larger diameter lines are inspected more rigorously and routinely. 

Besides the instrument and drain lines, the recirculation and, to a slightly lesser extent, the CRD and RHR 

lines are also of concern, primarily as a result of SCC susceptibility. 
L-7

LOCA Frequency (CY-1)

1e-14 1e-13 1e-12 1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2

Recirc

Feedwater

Main Steam

HPCS/LPCS

RHR

RWCU

CRD

SLC

INST

Drain

Head Spray

SRV

RCIC

N = 8

Figure L.6 Category 1 LOCA Frequencies for BWR Piping Systems at 25 Years of Plant Operation

 

For larger Category 3 LOCAs, the recirculation system was the largest contributor to the overall LOCA 

frequencies, see Figure L.7. (Note in this figure that the instrument and drain lines, as well as the CRD 

lines, are no longer shown in that these smaller diameter lines cannot support a Category 3 LOCA.) The 

fact that the recirculation system is the largest contributor is a slight departure from the PWR estimates 

where the smallest diameter piping system that can support a particular LOCA category consistently had

the highest LOCA frequencies. The main concern with the recirculation system piping continues to be 

SCC, even when considering the effective mitigation programs in place today. Of secondary importance 

were the feedwater, RHR, RWCU, core spray, and SRV systems. There was wide variability expressed 

for the feedwater system. Several participants thought that its susceptibility was similar to that of the 

recirculation system while others thought that it would make an inconsequential contribution. This latter 

group generally thought that the mitigation programs in place for the feedwater system were overall 

effective. The RHR system was deemed important by some panel members due to the relatively larger 

number of precursor events reported and the relatively high number of welds. A number of the 

participants used the weld census data provided to differentiate the relative contributions between systems 

for those systems that have similar operating experience. The SRV lines were judged to be potentially

problematic by four of the eight respondents who addressed the question of BWR piping. They pointed 

out that the SRV lines are subject to high dynamic loads during the relatively common SRV discharge 

events, however, only a short section of these lines are actually susceptible to a LOCA event. Overall, in 

comparing Figure L.6 with Figure L.7, one can see approximately a one order of magnitude reduction in 

the LOCA frequency between the Category 1 and 3 LOCAs for most of the BWR piping systems 

considered. 
L-8

LOCA Frequency (CY-1)

1e-14 1e-13 1e-12 1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2

Recirc

Feedwater

Main Steam

HPCS/LPCS

RHR

RWCU

SLC

Head Spray

SRV

RCIC

N = 8

Figure L.7 Category 3 LOCA Frequencies for BWR Piping Systems at 25 Years of Plant Operation 

For the largest category BWR piping LOCAs (Category 5), the recirculation system remains the main 

contributor to the overall LOCA frequencies, see Figure L.8. The RWCU system had about the same 

median value, however, there was a question expressed as to whether the RWCU system could actually 

sustain such a high flow rate LOCA. One of the participants thought that the maximum diameter for this 

system was only 6-inches, not 24-inches as specified in the development of the elicitation questions. 

Besides the recirculation, and RWCU systems, the next two largest contributors to the BWR Category 5 

LOCA frequencies were the feedwater and RHR systems. As for the Category 3 LOCAs, the RHR 

system was deemed important due to the large number of precursor events reported and the large number 

of potentially susceptible welds. Several of the participants indicated that these lines are susceptible to 

SCC. 
L-9

LOCA Frequency (CY-1)

1e-12 1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3

Recirc

Feedwater

Main Steam

RHR

RWCU

B F

E C

F

G F

E C

G

N = 8

Figure L.8 Category 5 LOCA Frequencies for BWR Piping Systems at 25 Years of Plant Operation 

Figure L.9 is a plot of the cumulative BWR piping LOCA frequencies (including contributions from all of 

the piping systems) for Category 1 through 5 LOCAs. The BWR piping LOCA frequency decrease with 

LOCA size is relatively shallow, i.e., approximately 1⁄2 order of magnitude per LOCA category. The 

results tend to be governed by the results from the recirculation system. It was noted that for the 

recirculation system that the mitigation programs in place for controlling IGSCC promote a more uniform 

residual stress field which can in turn promote longer cracks which are more likely to cause a LOCA. 

This effect will potentially offset the overall reduction in crack growth due to the mitigation program. It 

is also of note from Figure L.9 that the variability in the results as expressed by the interquartile range and 

the difference between the minimum and maximum values does not vary much with LOCA size. It is 

also of note that the expert ranking is relatively consistent with LOCA size, i.e., Participant C always 

predicted the highest LOCA frequencies and Participants E and G consistently predicted the lowest 

LOCA frequencies. 
L-10

N = 8

BWR LOCA Frequencies at 25 Years (CY-1)

1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2

Cat 1 LOCAs

Cat 2 LOCAs

Cat 3 LOCAs

Cat 4 LOCAs

Cat 5 LOCAs

E C

C

C

C

C,E

E,G

E,G

E

G

Figure L.9 Cumulative BWR LOCA Frequencies at 25 Years of Plant Operations 

Figure L.10 shows the effect of operating time on the cumulative Category 1 LOCA frequencies for BWR 

piping systems. As can be seen in Figure L.10, there is not much of an effect of operating time on the 

cumulative Category 1 frequency. Similar findings were evident for the larger Category 3 and 5 LOCAs. 

Obviously, any unabated degradation mechanism would cause an increase in the overall LOCA 

frequencies. However, it was generally assumed by the panel members that any new degradation 

mechanism that came on the scene would be aggressively met by the industry and NRC, just like the 

IGSCC problem in BWRs was met in the past and the PWSCC problem in PWRs is being met today. The 

minimal changes in LOCA frequencies with time evident in Figure L.10 were the result of a number of 

compensating factors considered by the panel members. From the perspective of potential decreases in 

the LOCA frequencies, the recirculation lines should see a decrease in the LOCA frequencies with respect 

to the current-day estimates that are based on an analysis of operational experience data due to improved 

mitigation strategies that have been put in place. The panelists generally felt that the IGSCC issue for 

BWRs had been effectively mitigated for the foreseeable future. In addition, the core spray systems may 

see a decrease in the LOCA frequencies with time as the segments of stainless steel piping potentially

susceptible to IGSCC are replaced with carbon steel piping. Finally, future inspection and mitigation

programs are expected to lead to additional decreases in the predicted LOCA frequencies. In this regard, 

having the industry focus its inspection resources on the more important systems through risk-informed

ISI should help reduce the propensity for LOCAs. Counteracting these potential decreases are potential 

increases due to bigger thermal fatigue and FAC concerns in the future. Concern was expressed about the 

high usage factors that will exist near the end-of-plant license. Also, there is the concern with new, 

previously unknown degradation mechanisms that may arise in the future. In this regard, the inspection 

methods of today may not be reliable for these new mechanisms. Furthermore, these new mechanisms 

may not manifest themselves in the same locations of concern today. Finally, while timely and proper 

maintenance programs are always beneficial, there are instances in which they may prove 

counterproductive. The frequent opening and closing of systems for inspections increases the likelihood 
L-11

for human error such as having tools and other debris left behind or bolts not being torqued properly. 

Also, improper service of active components (e.g., valves) can lead to passive system failures. 

N = 8

BWR Cat 1 LOCA Frequencies (CY-1)

1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2

25 Years

40 Years

60 Years

A,G

G

G

H

H

C,E

Figure L.10 Effect of Operating Time on the Cumulative Category 1 LOCA Frequencies for BWR 

Piping Systems 

Figures L.11 and L.12 show the cumulative MV estimates, along with the 5% and 95% bound values for 

the various participants for the Category 1 and 3 LOCAs, respectively. The uncertainty range (difference 

between 5% LB and 95% UB values) for the Category 3 LOCAs are comparable (or slightly greater than) 

for the Category 1 LOCAs. Only participants A, E, and F expressed considerably more uncertainty for 

the Category 3 LOCAs than they did for the Category 1 LOCAs. Similar findings were found when 

comparing the Category 5 results with the Category 3 results. Overall, the panelists appeared more 

confident about their BWR estimates than they did for the corresponding PWR estimates. They had less 

uncertainty about future and bigger size LOCA frequencies compared with their PWR predictions. There 

was also less uncertainty among the panelists about the magnitude of the dominant contributing factors. 

In addition, the panel members used more consistent approaches and more consistent base case estimates

for the BWR estimates than they did when making their estimates for PWRs. 
L-12

1.00E-11 1.00E-09 1.00E-07 1.00E-05 1.00E-03 1.00E-01

BWR Category 1 LOCA Frequencies (CY-1)

A

B

C

E

F

G

H

Solid Diamonds are 5% Lower Bound, Mid-Value, 95% Upper Bound

D

Figure L.11 BWR Category 1 LOCA Frequencies Showing MVs, 5% LB, and 95% UB Values for 

All Participants Who Responded to the BWR Piping Questions 

1.00E-11 1.00E-09 1.00E-07 1.00E-05 1.00E-03 1.00E-01

BWR Category 3 LOCA Frequencies (CY-1)

A

B

C

E

F

G

H

Solid Diamonds are 5% Lower Bound, Mid-Value, 95% Upper Bound

D

Figure L.12 BWR Category 3 LOCA Frequencies Showing MVs, 5% LB, and 95% UB Values for 

All Participants Who Responded to the BWR Piping Questions 
L-13

L.3 PWR Piping

The primary aging mechanisms identified by the participants for PWR plants are thermal fatigue, 

PWSCC, and mechanical fatigue. The concerns associated with thermal and mechanical fatigue in PWR 

plants are similar to those in BWR plants. In PWRs, the surge line is thought to be susceptible to thermal 

fatigue due to cyclic thermal stratification stresses. Also, the DVI and CVCS lines were thought to be 

susceptible due to periodic testing which imposes additional thermal cycles. In addition, the DVI and

Accumulator SIS lines have experienced some thermal fatigue cracking issues due to leakage of cold 

water past the check valves in these systems. 

PWSCC is a relatively new mechanism that has manifested itself in this country over the last 5 years. It 

has many similar characteristics to the IGSCC problem experienced in BWR reactors in the past. It is a 

temperature dependent mechanism that attacks Alloy 600 type materials such as bimetallic Inconel 

82/182 welds. Many panel members believe that PWSCC problems will be resolved (i.e., mitigated) over 

the next 15 years. Therefore, its contribution to the overall LOCA frequencies may peak between the 25 

and 40 year time period, but then decrease in the future. Today, instances of PWSCC have been observed 

in surge lines at the surge line to pressurizer weld in the United States at Three Mile Island, in Belgium 

and Japan, and in hot legs at the hot leg to RPV weld in the United States at V.C. Summer and Sweden. 

Other lines in which PWSCC may become an issue in the future are the cold leg and the pressurizer spray 

lines. However, since the cold leg operates at lower temperatures than the hot legs and surge lines, any 

problems that may materialize in the cold leg will be delayed until later in the operating life. 

As part of this elicitation a total of 12 LOCA sensitive piping systems were considered for the PWR 

plants. However, as was the case for BWR piping, most of the participants focused on a few common 

systems as being important LOCA contributors. Figure L.13 shows the Category 1 LOCA frequencies at 

25 years for the 12 PWR LOCA sensitive piping systems. As was the case for the BWR piping systems, 

the Category 1 LOCA frequencies are dominated by the small diameter instrument and drain lines. The 

estimated Category 1 LOCA frequencies for these lines are two orders of magnitude higher than the hot 

leg and surge lines. This again reflects the belief that a complete break of a smaller pipe is more likely 

than a partial break of comparable size of a larger pipe. The reasoning provided by the panel members for 

why these small diameter lines are so susceptible to these smaller category LOCAs are the same as 

provided for the case of BWRs. These 1- and 2-inch socket welded lines are susceptible to vibration 

fatigue concerns. Also, they are more difficult to inspect, if they are inspected at all, than their larger 

diameter counterparts. Generally speaking, the benefits of leak detection and ISI decrease with 

decreasing pipe size. Finally, they are more susceptible to other, non-aging related failure mechanisms, 

e.g., an externally applied overload. 
L-14

LOCA Frequency (CY-1)

1e-13 1e-12 1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2

Hot Leg

Cold Leg

Surge

SIS: ACCUM

SIS: DVI

DRAIN

CVCS

RHR

SRV

PSL

RH

INST

N = 9

Figure L.13 Category 1 LOCA Frequencies for PWR Piping Systems at 25 Years of Plant 

Operation

Figure L.14 shows the breakdown of PWR Category 3 LOCA frequencies by piping system at 25 years of 

plant operations (present day). The small diameter instrument and drain lines, as well as the RH lines, do 

not appear on this figure in that they are of such size that they could not sustain a Category 3 LOCA. 

Again, as was the case for the PWR Category 1 LOCAs, the smallest diameter lines that can sustain this

size (i.e., category) of LOCA are the dominant contributors. These include the CVCS, SIS-DVI, RHR, 

surge, and PSL. This is different than what was observed for the BWR Category 3 LOCAs where the 

larger recirculation system was the dominant contributor, primarily due to its susceptibility to IGSCC. 

The two most listed systems as being major contributors to this category of LOCA for PWR piping were 

the CVCS and SIS-DVI lines. For both, the primary concern was fatigue. One participant commented 

that the CVCS line was one of the most fatigue sensitive locations in the entire plant. Another 

commented that they were concerned with environmentally-assisted fatigue for this system. With regard

to the SIS-DVI (and the SIS-Accumulator lines for that matter), several participants indicated that both 

lines had experienced thermal fatigue cracking in the past due to cold water leaking past the check valves. 

Another line that a number of participants thought would be a major contributor to this category of LOCA 

was the RHR lines. The concern with these lines was with environmental attack due to the stagnant 

nature of the flow in these lines. The pressurizer spray lines were of a concern due to the chance for 

PWSCC at one of the bimetal welds. 
L-15

n = 9

LOCA Frequency (yr-1)

1e-13 1e-12 1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2

Hot Leg

Cold Leg

Surge

SIS: ACCUM

CVCS

RHR

SRV

PSL

SIS: DVI

Figure L.14 Category 3 LOCA Frequencies for PWR Piping Systems at 25 Years of Plant 

Operation

For the largest categories of PWR piping LOCAs (Categories 5 and 6), the hot leg, cold leg, surge line, 

and RHR lines all contribute to the overall LOCA frequencies, see Figure L.15 for the Category 5 

LOCAs. Of these, the median value of the LOCA frequency for the cold leg is about a half order of 

magnitude less than the median values for the other three piping systems. This slight reduction is 

primarily due to the fact that the cold leg is less susceptible to PWSCC than either the hot leg or surge 

line at this time (25 years of plant operations) due to the fact that it operates at a slightly lower 

temperature. Somewhat surprisingly in examining Figure L.15, a number of the participants felt that the 

hot leg would have a greater propensity for a Category 5 LOCA than the surge line. Both lines are 

susceptible to PWSCC due to the presence of bimetallic welds and the high operating temperatures, but 

the surge line was also judged to be susceptible to thermal fatigue due to thermal stratification and 

thermal striping stresses. Also, the surge line is smaller diameter, which based on the thought that smaller 

diameter lines are more prone to LOCAs than their larger counterparts, would imply that the Category 5

LOCA frequencies for the surge line should be higher. Finally, at least one participant argued that the 

surge line to pressurizer bimetallic weld was one of their biggest concerns in the entire plant due to its 

susceptibility to PWSCC and the fact that it is a very difficult weld to inspect. Counteracting these

arguments, however, is the fact raised by a number of the participants that there are more hot leg to RPV 

bimetal welds (2 to 4 depending on the number of loops) in a plant than there are surge line to pressurizer 

bimetal welds (one).
L-16

n = 9

LOCA Frequency (yr-1)

10-15 10-14 10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 10-5

Hot Leg

Cold Leg

Surge Line 

RHR Line

A

L

L

L

J

J

J

J

Figure L.15 Category 5 LOCA Frequencies for PWR Piping Systems at 25 Years of Plant 

Operation 

Figure L.16 is a plot of the cumulative PWR LOCA frequencies at 25 years of plant operation. 

Cumulative frequencies are shown for Category 1, 3, and 6 LOCAs. Based on a review of Figure L.16 

there appears to be approximately a one order of magnitude reduction in LOCA frequency between each 

successive LOCA category. 
L-17

N = 9

PWR LOCA Frequencies at 25 Years (CY-1)

1e-15 1e-14 1e-13 1e-12 1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2

Cat 1

Cat 3

B Cat 6

I

L I

G,H,J

J

Figure L.16 Cumulative PWR LOCA Frequencies at 25 Years of Plant Operations 

Figure L.17 shows the effect of operating time on the cumulative Category 1 LOCA frequencies for PWR 

piping systems. Several participants felt that the operational experience is sufficient to expect the

frequencies to remain relatively constant out to 60 years of life. Degradation and aging will naturally 

continue to occur. However, the inspection and mitigation strategies will effectively identify and temper 

the frequency increases caused by this aging. Some panelists expected a short term frequency increase

due to PWSCC before effective mitigation is developed. This trend is consistent with the historical 

response to evidence of emerging degradation by the industry. Also, at least one participant expressed a 

concern about the high usage factors that will exist at 60 years at many locations. All of these concerns 

are reflected in the results showing the effects of operating time and aging in Figures L.17 and L.18 for 

Category 1 and 3 LOCAs, respectively. As can be seen in Figures L.17 and L.18, there is a slight 

increase in the cumulative Category 1 and 3 LOCA frequencies between 25 and 40 years, but not much of 

an effect between 40 and 60 years. The median LOCA frequencies for the Category 1 and 3 LOCAs at 40 

years are an order of magnitude higher than the median LOCA frequencies for the Category 1 and 3 

LOCAs at 25 years. Similar findings were evident for the larger Category 6 LOCAs. The rationale 

behind this is that this size of LOCA (and associated pipe size) is most affected by aging. These pipes are 

not as easily inspected, or as leak sensitive, as their larger counterparts and these pipes have not 

experienced the infant mortality as their smaller counterparts. 
L-18

N = 9

PWR Category 1 LOCA Frequencies(CY-1)

1e-7 1e-6 1e-5 1e-4 1e-3 1e-2 1e-1 1e+0 1e+1

25 years

40 years

60 years

A

A

L

J

I

I

Figure L.17 Effect of Operating Time on the Cumulative Category 1 LOCA Frequencies for PWR 

Piping Systems 

N = 9

PWR Category 3 LOCA Frequencies(CY-1)

1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2

25 Yrs

40 Yrs

60 Yrs

G,H,J

G

J

B

I

G

Figure L.18 Effect of Operating Time on the Cumulative Category 3 LOCA Frequencies for PWR 

Piping Systems 
L-19

Figures L.19 and L.20 show the cumulative MV estimates, along with the 5% and 95% bound values for 

the various participants for the Category 1 and 3 LOCAs, respectively. The uncertainty range (difference 

between 5% LB and 95% UB values) for the Category 3 LOCAs are typically greater than for the 

Category 1 LOCAs for most of the participants. In a similar vein, the level of uncertainty for the 

Category 6 estimates were much greater than for the Category 1 or 3 estimates, see Figure L.21. All of 

the panelists had at least two orders of magnitude difference between the LB and UB values for their 

Category 6 estimates, and some of the panelists (C, E, and J) had greater than four orders of magnitude 

difference. 

1.00E-17 1.00E-15 1.00E-13 1.00E-11 1.00E-09 1.00E-07 1.00E-05 1.00E-03 1.00E-01

PWR Category 1 LOCA Frequencies (CY-1)

Solid diamonds are 5% Lower Bound, Mid-Value, and 95% Upper Bound Values

J

I

H

G

E

C

B

A

L

Figure L.19 PWR Category 1 LOCA Frequencies Showing MVs, 5% LB, and 95% UB Values for 

All Participants Who Responded to the PWR Piping Questions 
L-20

1.00E-17 1.00E-15 1.00E-13 1.00E-11 1.00E-09 1.00E-07 1.00E-05 1.00E-03 1.00E-01

PWR Category 3 LOCA Frequencies (CY-1)

A

B

C

E

G

H

I

J

L

Solid diamonds are 5% Lower Bound, Mid-Value, and 95% Upper Bound Values

Figure L.20 PWR Category 3 LOCA Frequencies Showing MVs, 5% LB, and 95% UB Values for 

All Participants Who Responded to the PWR Piping Questions 

1.00E-17 1.00E-15 1.00E-13 1.00E-11 1.00E-09 1.00E-07 1.00E-05 1.00E-03 1.00E-01

PWR Category 6 LOCA Frequencies (CY-1)

A

B

C

E

G

H

I

J

L

Solid diamonds are 5% Lower Bound, Mid-Value, and 95% Upper Bound Values

Figure L.21 PWR Category 6 LOCA Frequencies Showing MVs, 5% LB, and 95% UB Values for 

All Participants Who Responded to the PWR Piping Questions 
L-21

In general, the results for PWR piping appear consistent. The quantitative results and the qualitative 

opinions and rationale were for the most part consistent. The variability between participants stems from 

the different approaches used and the basis for their estimates. Several different approaches with different 

anchoring points were used by the different panelists. The variability between the participants seems

reasonable given the frequency magnitudes being computed. 

L.4 BWR Non-Piping

Generally speaking making estimates of LOCA frequencies for non-piping components is more 

challenging than making estimates for piping systems. There are multiple components to consider, each

having different operating requirements, design margins, materials, and inspectability. There are also 

widely varying failure modes and scales to consider. For PWRs for the smaller category LOCAs, one 

must consider SGTRs and small penetration failures. For the larger category LOCAs, common cause 

bolting failures and component shell failures need to be considered. For the larger components, the 

bigger design margins (compared to those for piping) are somewhat offset by the decreased inspection 

quantity and quality. Compounding all of this is the fact that there is generally not as much precursor 

information available for the non-piping components as there is for piping. 

For the BWR plants, the three major non-piping components that were considered were the RPV, the 

pumps, and the valves. In general, many of the same degradation mechanisms that are of concern for 

BWR piping are also a concern for BWR non-piping components. Stress corrosion cracking (specifically 

PWSCC) is a concern for many of the smaller Alloy 600 components, such as the CRDMs and other 

penetrations. As with piping, multiple cracks and fast propagation rates could lead to LOCAs. While the 

mechanism (PWSCC) is more severe at higher temperatures associated with PWRs, this mechanism could 

become a more significant issue later in the life of the BWRs. Thermal fatigue is another degradation

mechanism associated with BWR non-piping components that is common with BWR piping. Thermal 

fatigue is especially of concern at inlet nozzles and other locations that experience thermal stratification, 

especially at the feedwater nozzles. For the same reasons as highlighted above for BWR piping, thermal 

fatigue can possible lead to larger leaks or LOCAs. 

Other mechanisms for non-piping components that were not of concern for BWR piping are radiation 

embrittlement, common cause bolting failures, and thermal aging of cast stainless steel components, such 

as pump and valve casings. Radiation embrittlement reduces the base metal toughness of the RPV. 

Fortunately, for BWRs, it is not as of much concern as it is for PWRs due to the increased shielding 

available with the BWRs. Common cause bolting failures are important for manways and bolted valves. 

The common cause mechanisms may possibly include: improper installation or maintenance of the bolts, 

i.e., improper torque, external corrosion of multiple bolts, and steam cutting of bolts. One participant 

thought that these common cause failures will cause the greatest risk. Thermal aging of cast stainless 

steels can cause a significant reduction in the fracture toughness of these materials, however, fortunately 

to date no cracking mechanisms have emerged for these materials. 

Figure L.22 shows the Category 1 LOCA frequencies for the RPV, pumps, and valves at 25 years. The 

RPV shows the biggest expected Category 1 LOCA frequencies. The Category 1 RPV LOCA 

frequencies are driven by the CRDM penetration failures. However, the severity of the CRDM failures 

for BWRs was reduced by about one order of magnitude with respect to PWR CRDM failures due to the 

BWR heads operating at a lower temperature. Other than these head penetrations, nozzle and body 

cracking were mentioned as possible sources of failures. A number of precursor cracking incidents have 

been seen in service. The valves and pumps contribute to a lesser extent. Most of the panelists generally 

treated these components the same. At least one panelist (F) had some experience with manufacturing 
L-22

defects in valve bodies which led to some increased concern with valves. Other issues with valves, and 

pumps, included the potential for bolt failures for the reasons outlined above, and the fact that the material 

they are made of (cast stainless steel) is notoriously difficult to inspect and is subject to toughness 

degradation due to thermal aging. 

N = 8

LOCA Frequency (CY-1)

10-12 10-11 10-10 10-9 10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1

RPV

Pumps

Valves

F C

G E

G C

Figure L.22 BWR Category 1 Non-Piping LOCA Frequencies by Major Component at 25 Years of 

Plant Operations 

Figure L.23 shows the Category 3 non-piping LOCA frequencies at 25 years of plant operations. The 

most noticeable difference between the Category 3 and Category 1 LOCA frequencies is the three orders 

of magnitude reduction in the median value of the estimated LOCA frequencies for the RPV as the 

CRDM concerns disappear. A single CRDM failure cannot support a Category 3 LOCA. Only about half 

of the participants were concerned about RPV nozzle failures, but those that were assigned comparatively 

high frequencies to them. For the pumps and valves, the corresponding decrease in LOCA frequency is 

only about one order of magnitude. Consequently, the pumps, valves, and vessel now contribute about 

equally to the overall LOCA frequency. 
L-23

N = 8

LOCA Frequency (CY-1)

10-12 10-11 10-10 10-9 10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1

RPV

Pumps

Valves

B,F A

G E

G C

Figure L.23 BWR Category 3 Non-Piping LOCA Frequencies by Major Component at 25 Years of 

Plant Operations 

Figure L.24 shows the Category 5 non-piping LOCA frequencies at 25 years for the BWR non-piping 

components. As was the case for the Category 3 LOCAs, the pumps, valves, and vessel are all now of 

about equal importance. For these large LOCAs, the panelists felt that the valve, pump, and vessel bodies 

were the most likely subcomponents to fail. For the vessel body, the concern was with LTOP while for

the valve and pump bodies, the concern was with fatigue and SCC. 
L-24

N = 8

LOCA Frequency (yr-1)

10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 10-5

RPV

Pumps

Valves

B

C

G A

G F

Figure L.24 BWR Category 5 Non-Piping LOCA Frequencies by Major Component at 25 Years of 

Plant Operations 

Figure L.25 shows the cumulative LOCA frequencies for the BWR non-piping components at 25 years of 

plant operations. On average there is about a one order of magnitude shift in the cumulative LOCA 

frequency between each successive LOCA category. The median value for the estimate of the Category 1 

LOCA frequency is approximately 10-4 while the median value for the Category 6 LOCA frequency is 

about 10-9. 
L-25

LOCA Frequency (yr-1)

10-14 10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 10-5 10-4 10-3 10-2

N = 8

Cat 1 F

Cat 3

Cat 4

Cat 5

Cat 6

Cat 2

H

C,H

H

A

C

D,F,G

A

H

H

H A

Figure L.25 Cumulative BWR Non-Piping LOCA Frequencies at 25 Years of Plant Operations 

Figures L.26 and L.27 show the effect of time on the Category 1 and 3 cumulative LOCA frequencies, 

respectively, for the BWR non-piping components. For all intents and purposes there is almost no effect 

of time on the predicted LOCA frequencies. The median values do not change nor does the variability, 

i.e., the interquartile ranges remain the same. Non-piping components are affected by similar partially 

compensating factors as the piping components. In addition, a number of participants expressed the belief 

that the maintenance and mitigation issues raised for piping also apply for non-piping components. The 

only thing that changes is the minimum value predicted by Participant H for LOCA Category 3. 

Participant H foresees the non-piping LOCA frequencies increasing at both the 40 and 60 year time 

interval. Figure L.28 shows the effect of time on the Category 5 frequencies. In this case the median 

values do not vary with time, nor does the maximum values, however, a number of participants started to 

see the LOCA frequencies increasing near the end-of-plant license renewal (60 years) such that the lower 

end of the IQR (i.e., the 25th percentile) increased an order of magnitude at 60 years over what it was at 40 

and 25 years. This increase in the Category 5 predictions was driven by aging concerns of a few of the 

panelists at 60 years. 
L-26

LOCA Frequency (yr-1)

10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1

N = 8

60 Years

40 Years

25 Years C,H

H

H

F

F

F

Figure L.26 Effect of Operating Time on the Cumulative Category 1 LOCA Frequencies for BWR 

Non-Piping Components 

LOCA Frequency (yr-1)

10-12 10-11 10-10 10-9 10-8 10-7 10-6 10-5 10-4

N = 8

60 Years

40 Years

25 Years

A

B,H

A

A

H

H

Figure L.27 Effect of Operating Time on the Cumulative Category 3 LOCA Frequencies for BWR 

Non-Piping Components 
L-27

LOCA Frequency (yr-1)

10-14 10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6

N = 8

60 Years

40 Years

25 Years

A

H

A

A

H

H

Figure L.28 Effect of Operating Time on the Cumulative Category 5 LOCA Frequencies for BWR 

Non-Piping Components 

Figures L.29 and L.30 show the cumulative MV estimates, along with the 5% and 95% bound values for 

the various participants for the BWR Category 1 and 3 non-piping LOCA frequency estimates, 

respectively, at 25 years of plant operating time. Of note from these figures is the fact that a number of 

the participants (A, E, F, and H) predicted greater uncertainty for the Category 3 LOCAs than they did for 

the Category 1 LOCAs. This is not unusual in that one would expect the uncertainty to increase for lower 

frequency events, such as larger LOCAs. It is probably somewhat more surprising that the other four 

participants predicted comparable uncertainty for the Category 1 and 3 LOCAs. Overall the predictions

for BWR non-piping were more consistent than for PWR non-piping discussed next. For the BWRs, 

there are less components and failure modes to consider and the approaches used to estimate the 

frequencies were more closely related. 
L-28

LOCA Frequency (yr-1)

10-11 10-10 10-9 10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1

A

B

C

E

G

F

H

D

Figure L.29 BWR Non-Piping Category 1 LOCA Frequencies Showing MVs, 5% LB, and 95% UB 

Values for All Participants Who Responded to the BWR Non-Piping Questions 

LOCA Frequency (yr-1)

10-11 10-10 10-9 10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1

A

B

C

E

G

F

H

D

Figure L.30 BWR Non-Piping Category 3 LOCA Frequencies Showing MVs, 5% LB, and 95% UB 

Values for All Participants Who Responded to the BWR Non-Piping Questions 
L-29

L.5 PWR Non-piping

The same three major non-piping components (RPV, valves, and pumps) as considered for BWRs are 

considered for PWRs, plus the steam generator and pressurizer are added. One of the operational 

experience databases showed an order of magnitude higher incident rate for PWR non-piping than BWR 

non-piping. This was partially attributed to the fact that there are more PWRs than BWRs. However, this 

comparison is also biased by the large number of steam generator tube failures reported in the databases. 

Steam generator tubes are subjected to a host of degradation mechanisms: fatigue, denting, external SCC, 

PWSCC, and overload failures. It was almost universally accepted that SGTRs would be the dominant 

contributor to the PWR Category 1 non-piping LOCA frequency. In fact the PWR steam generator tube 

failure frequency is the dominant contributor to the overall PWR small-break LOCA (Category 1 LOCAs) 

frequency when considering both the piping and non-piping contributions. Even so, it is the expectation 

of a number of the panel members that the steam generator tube contribution to the small-break LOCA 

frequency will decrease with time due to steam generator tube replacement programs and improvements 

made to the secondary side water chemistry. 

In general, many of the same degradation mechanisms that are important for PWR piping are important 

for the non-piping components as well. PWSCC is an important degradation mechanism for many of the 

smaller Alloy 600 components such as the CRDMs, heater sleeves, steam generator tubes, and other 

penetrations. As was the case for piping, the likelihood of multiple cracks forming, and possibly 

coalescing, and the relatively fast propagation rates associated with this type of cracking makes this

mechanism a major concern from a LOCA perspective. Also, since this mechanism is more severe at the 

higher temperatures associated with PWRs, it is considered to be a bigger threat for the PWRs than the

BWRs, at least in the short term. As was the case for PWR piping, thermal fatigue is also a concern for 

PWR non-piping components. It is especially of concern at nozzle inlets and other locations where 

thermal stratification may exist. Furthermore, for all the reasons highlighted for piping, thermal fatigue is 

a mechanism that can lead to large leaks, i.e., fast propagation rates, attacks a wide area, and difficult to 

prioritize inspection protocols due to the fact that it can attack a variety of materials. Mechanical fatigue 

is another common degradation mechanism to both PWR piping and non-piping components. 

Mechanical fatigue is most important for smaller components, such as heater sleeves and small 

penetrations that are subjected to vibratory stresses due to equipment operation. 

Another mechanism of special concern to non-piping components is common cause bolting failures. This 

is especially relevant to manways and bolted valves. The common cause mechanism could be improper 

installation or maintenance of bolts, e.g., improper torque, external corrosion of multiple bolts, or 

possibly steam cutting of multiple bolts. 

Also, boric acid corrosion of carbon steel components such as RPV and steam generators can be 

aggressive under certain conditions. 

Figure L.31 shows the Category 1 LOCA frequencies for the major PWR non-piping components at 25 

years. As expected, the expected failure frequencies are highest for the steam generator. The higher

LOCA frequency for the steam generator is driven by the SGTR data. 
L-30

N = 9

LOCA Frequency (CY--1)

1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2

RPV

Pumps

Valves

PRESS

SG

G H

C

C

I

G

G B

G C

Figure L.31 PWR Category 1 Non-Piping LOCA Frequencies by Major Component at 25 Years of 

Plant Operations 

The other major contributors to the Category 1 LOCA frequencies for PWR non-piping were the RPV and 

the pressurizer. The main subcomponent contributing to the RPV frequency is the CRDMs while the 

main subcomponent contribution to the pressurizer frequency is the heater sleeves. For Category 2 

LOCAs, a single SGTR cannot sustain such a leak. Thus, for the Category 2 LOCAs, the CRDM and 

pressurizer heater sleeves became the main contributors. 

For Category 3 and 4 LOCAs there was no consistent agreement among the panelists as to the major 

contributors. As one can see in Figure L.32, all five major components contribute fairly equally to the 

Category 4 LOCA frequencies. As such, there is tremendous variability about the frequency associated 

with each component. This variability was also apparent for the Category 6 LOCAs. This variability 

reflects the inconsistent opinions and approaches followed by the panelists, as well as the difficulty of this 

type of assessment. As is to be expected, there was a wide array of possible failure modes for dissimilar 

components to be considered, and the panelists tended to gravitate towards a few of the failures that they 

personally thought were most credible. Given all of this, the level of variability was thought to be 

reasonable given the event frequencies. This was one reason for adopting the elicitation approach in the 

first place. The highest LOCA frequencies were for the pressurizer nozzle. In addition, many of the 

panelists considered manway or shell failures important, irrespective of the component type. Thus, they 

anticipated similar distributions for both the steam generator and pressurizer. There were also major

differences of opinion among the panelists as to the most important failure modes. 
L-31

N = 9

LOCA Frequency (CY-1)

10-11 10-10 10-9 10-8 10-7 10-6 10-5

RPV

Pumps

Valves

PRESS

SG

H I

C

C

G

G

G I

C A

Figure L.32 PWR Category 4 Non-Piping LOCA Frequencies by Major Component at 25 Years of 

Plant Operations 

Figure L.33 shows the cumulative LOCA frequencies for the PWR non-piping components at 25 years of 

plant operations. The Category 1 LOCA frequencies for PWR non-piping are the highest frequencies 

estimated by the elicitation panel for piping or non-piping, BWR or PWR. The median frequency is 

almost 5x10-3. The variability among the panelists was very small. The difference between the minimum 

and maximum predictions was less than an order of magnitude. These high frequencies and low 

variability were driven by the SGTR data for which ample data exist in the operational experience 

databases; thus explaining both the high frequencies and excellent agreement between participants. For 

the Category 2 LOCAs, the agreement, at least on a minimum and maximum basis, is not nearly as good. 

However, the agreement on the basis of the spread in the IQR is nearly as good as it is for the Category 1 

LOCAs. Again, for the Category 2 LOCAs, the major contributors are the CRDMs and the pressurizer 

heater sleeves. The much wider variability for the Category 3 through 6 LOCAs reflects differences in

opinion among the panelists as to the important failure modes and their associated frequencies. 
L-32

LOCA Frequency (CY-1)

1e-14 1e-13 1e-12 1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2 1e-1

Cat 1

Cat 2

Cat 3

Cat 4

Cat 5

Cat 6 J A

J

H

J

G B,H

H

I

I

A

H

N = 9

Figure L.33 Cumulative PWR Non-Piping LOCA Frequencies at 25 Years of Plant Operations 

Figure L.34 shows the effect of time on the PWR non-piping Category 1 LOCA frequencies. There is a 

very slight decrease in the frequency between 25 years (present day) and 40 years (end of plant license) 

due mostly to steam generator replacement programs and improved inspection and mitigation programs, 

e.g., improved eddy current inspection programs and improved secondary side water chemistry. There 

was also an expected decrease in the LOCA frequencies associated with CRDMs due to on-going head 

replacement programs and improved CRDM inspection programs that may go into effect over the next 

few years. However, there was some concern expressed that the maintenance and inspection programs 

for the larger component bodies (pressurizer, steam generator, RPV) may not be as rigorous as for the 

piping systems. Figure L.35 shows the effect of operating time on the PWR non-piping Category 6 

LOCA frequencies. As can be seen in Figure L.35, the median values remain constant with time and the 

variability among the participants (at least on the basis of the IQR) also remains fairly constant. This 

tends to indicate that the participants did not foresee any significant aging effects to occur. 
L-33

LOCA Frequency (CY-1)

1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2 1e-1 1e+0

N = 9

60 Years

40 Years

25 Years

B,H

H

J

G

L

L

Figure L.34 Effect of Operating Time on the Cumulative Category 1 LOCA Frequencies for PWR 

Non-Piping Components 

LOCA Frequency (CY-1)

1e-14 1e-13 1e-12 1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5

N = 9

60 Years

40 Years

25 Years

A

A

A

J

J

L

Figure L.35 Effect of Operating Time on the Cumulative Category 6 LOCA Frequencies for PWR 

Non-Piping Components 
L-34

Contrary to what was observed for the Category 1 and 6 LOCA frequencies, the median value of the 

Category 4 LOCA frequencies increases an order of magnitude between 25 and 40 years and then remains 

constant after that, see Figure L.36. As was the case for PWR piping, aging was thought to have the 

largest impact on LOCA Categories 3 and 4. It was thought by some that aging could accelerate near the 

end of the plant license faster than the effects of mitigation and inspection could become effective, 

especially if the plant operators do not see a return on their investment for such inspection and mitigation 

programs near the end of the plant’s license. 

LOCA Frequency (CY-1)

1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2 1e-1

N = 9

60 Years

40 Years

25 Years

I

J

H

J

H

H

Figure L.36 Effect of Operating Time on the Cumulative Category 4 LOCA Frequencies for PWR 

Non-Piping Components 

Figures L.37 and L.38 show the cumulative MV estimates, along with the 5% and 95% bound values for 

the various participants for the PWR Category 1 and 4 non-piping LOCA frequency estimates, 

respectively, at 25 years of plant operating time. Of note from these figures is higher uncertainty among 

almost all of the participants for the Category 4 LOCAs when compared with the Category 1 LOCAs. A 

number of the panelists showed 3 to 4 orders of magnitude of uncertainty for the Category 4 LOCAs 

while all of the panelists had less than approximately 2 orders of magnitude of uncertainty in their 

Category 1 results. In addition, the variability in the panelist’s MV estimates was within 1 order of

magnitude for their Category 1 results while the variability among their results spread over a range of 

almost five orders of magnitude for their Category 4 results. The fact that the agreement among the 

panelists was so good for the Category 1 predictions plus the low uncertainty of their individual 

predictions is a reflection that there was near consensus agreement that the single overwhelming dominant 

contributor to this class of LOCAs was SGTRs, for which ample field experience is available in the 

operational experience databases. 
L-35

LOCA Frequency (CY-1)

10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1 100

A

B

C

E

G

I

J

L

H

Figure L.37 PWR Non-Piping Category 1 LOCA Frequencies Showing MVs, 5% LB, and 95% UB 

Values for All Participants Who Responded to the PWR Non-Piping Questions 

LOCA Frequency (CY-1)

10-13 10-12 10-11 10-10 10-9 10-8 10-7 10-6 10-5 10-4 10-3 10-2 10-1 100

A

B

C

E

G

I

J

L

H

Figure L.38 PWR Non-Piping Category 4 LOCA Frequencies Showing MVs, 5% LB, and 95% UB 

Values for All Participants Who Responded to the PWR Non-Piping Questions 

APPENDIX M 

PUBLIC COMMENT RESPONSES 

M-1

APPENDIX M – PUBLIC COMMENT RESPONSES 

General Comments

Comment Number: GC1 

Submitted by: Bill Galyean – Idaho National Laboratory 

Comment: [Note: The footnote indications below do not appear in the submitted comment; they 

were added as reference points for the response.] Aside from the fact that I was a contributing panel

member in the elicitation process, I want to express my compliments on the effort made by the NRC 

management and staff to produce realistic and useful results. I believe with the significant research 

performed in recent years coupled with the accumulated operating experience, reasonable estimates of 

LOCA frequencies can be made. NRC has recognized these facts and acted accordingly and 

appropriately. That said, I also wish to express my opinion on the some of the details of the elicitations 

process and portions of the subsequent analyses about which I disagree, but with the acknowledgement 

that had they been done differently, the results would not change significantly (i.e., the reported results 

would probably be reduced by less than an order of magnitude). The first issue relates to the 

interpretation of the LOCA frequencies and associated uncertainties. The instructions given to the panel 

members stated that we were to make a best-estimate1

 of the “single ‘true’ value”2 for the industry-wide 

(or more accurately BWR-wide and PWR-wide) LOCA frequency. This is an issue because I do not 

believe a “single true value” exists for the LOCA frequency. Specifically, I believe each plant has 

differences in design, construction, operations, age, and maintenance such that in reality the plant-to-plant 

variation in LOCA frequencies will be quite large. These two interpretations can be made consistent 

however, if the “single true value” is viewed as an average or mean value of the population of 

frequencies. While on the surface this might seem to be a question of semantics and appears to have little 

significance, the implication of the interpretation on the uncertainty characterization is significant. If the 

“single true value” interpretation is employed, the question becomes, what does the uncertainty associated 

with this value represent. Since the implicit assumption is that plant-to-plant variability does not exist 

(otherwise how can there be a single frequency appropriate for the entire industry?), then there is no

stochastic or aleatory uncertainty associated with the estimate. That is, the presence of outliers (or event 

plant-to-plant variability) in the population of nuclear power plants (NPP) is ignored. The uncertainty 

therefore represents the level of confidence of each panel member’s estimate of this single true 

frequency.3

 Given that the uncertainty represents an individual’s confidence in their own estimate, what is 

the basis for automatically assuming the probability distribution associated with this confidence 

uncertainty is not symmetrical?4

 Or to ask more specifically, with this interpretation of the point estimate 

value and associated interpretation of the uncertainty, why do the authors assume that the uncertainty

surrounding each panel member’s estimate should be a lognormal distribution that is weighted toward 

higher (conservative) values? While I agree that the uncertainty associated with LOCA frequencies 

should be asymmetrically weighted toward higher values; this is based on the observation that not all 

plants are identical, and that if a LOCA occurs at a plant, that plant will likely be shown to be a poor 

performer, not representative of the fleet as a whole.5 This is not the same as assuming the confidence 

uncertainty surrounding the estimates provided by the panel members, of a “single true value” should be 

represented with a lognormal (which was done in the elicitation).6 [Note: the use of the lognormal 

distribution became entrenched in PRA with its use in WASH-1400 (1975). However, the model 

employed then was motivated by the sparse data available from the commercial nuclear power industry at

that time. Data used in that study were collected from many different industries and sources. These data 

were used to develop a probability distribution of the population of possible values, effectively capturing 

the random component of the uncertainty. Specifically, the authors of WASH-1400 did not know which 

value in the population of values collected, was the most appropriate value to use. Therefore the entire 

population of values was characterized in the probability distribution. The data were not combined, or
M-2

averaged to estimate a single value, and the uncertainty modeled with the lognormal distribution was not 

meant to describe a statistical confidence on a single true value, but used to describe the variability in 

possible values. Quoting directly from WASH-1400: “Because of the large spread, the failure rate data 

were treated as random variables, incorporating both the physical variability and the uncertainty 

associated with the rates. Moreover, since the study’s results were to apply to a population of 

approximately 100 nuclear plants, it was important to show the possible variability and uncertainty in this 

population.” The consequences of employing the lognormal distribution to characterize each individual

panel member’s uncertainty, manifests itself in how the estimates were interpreted and processed (by the 

authors) and in the effects of the overconfidence adjustment made to the base case results. Specifically, 

the best-estimates solicited by the authors of this “single true value” were interpreted as median values of 

the lognormal distribution, allowing for the derivation (by the authors) of a higher mean value. [Note that 

for any probability distribution skewed toward higher values, the lognormal being one example, the mean 

will always be greater than the median.] This calculated mean value was then used to represent the panel 

member’s input to the aggregation process used for generating the LOCA frequency results. Additionally, 

this assumption of a lognormal for the uncertainty on each panel member’s estimate, has the additional

impact of calculating an even higher (compared to the non-adjusted calculated mean) mean value after the 

widening of the uncertainty in the overconfidence adjustment. My concerns are twofold. First, the 

opinions of the panel members were solicited, and then modified (increased) by the authors.7

 Irrespective 

of the instructions and discussions during the elicitation process, this point was viewed with dismay by 

more than one panel member. Second, the processes and analyses employed have introduced a 

conservative bias into the final base case results, with additional conservative bias8 inserted in the various 

sensitivity studies.9 This “creeping conservatism” is not necessarily undesirable given the significant 

uncertainties and the uses to which the results will be employed; however, it should be explicitly 

acknowledged and clearly stated rather than obscured in the details of the analyses.10 

Response: The authors have identified ten separate issues in this comment, as indicated by the inserted 

footnotes. The responses to these issues are provided below. 

Issue 1: The instructions given to the panel members stated that we were to make a best-estimate1

 of the 

“single ‘true’ value”2

 for the industry-wide (or more accurately BWR-wide and PWR-wide) LOCA 

frequency. 

1. Issue 1 Response: The panel members were not asked to provide a ”best-estimate“ of any 

quantity. Rather, they were asked to provide a MV and LB and UB values for each question. 

The MV was defined such that, in the panel member’s opinion, the unknown true value for that 

particular question has a 50% chance of falling above or below the MV, with similar definitions 

for the LB and UB values (Section 3.8.5). 

Issue 2: The instructions given to the panel members stated that we were to make a best-estimate1 of the 

“single ‘true’ value”2 for the industry-wide (or more accurately BWR-wide and PWR-wide) LOCA 

frequency. 

2. Issue 2 Response: Additionally, the panel members were not asked to estimate a ”single ’true‘ 

value“ of any quantity. Rather, the elicitation focused on estimating generic, or average, LOCA 

frequencies for the commercial fleet by combining the contributions from individual component 

failures. As stated in Section 2 of NUREG-1829, the generic BWR and PWR estimates were 

determined by first estimating the separate LOCA frequency contributions associated with 

specific BWR piping, BWR non-piping, PWR piping, and PWR non-piping failures for each 

panelist. These individual piping and non-piping component failure frequencies were then 

combined to estimate parameters of the total passive system LOCA frequency distributions for 

BWR and PWR plants at each distinct LOCA category and time period. Panelists were 

specifically instructed to consider broad plant differences in estimating these component failure 
M-3

frequencies and their uncertainties (Section 3). More information related to the instructions 

given to the panel can be found in Sections 2 and 3 of NUREG-1829. 

Issue 3: If the “single true value” interpretation is employed, the question becomes, what does the 

uncertainty associated with this value represent. Since the implicit assumption is that plant-to-plant

variability does not exist (otherwise how can there be a single frequency appropriate for the entire 

industry?), then there is no stochastic or aleatory uncertainty associated with the estimate. That is, the 

presence of outliers (or event plant-to-plant variability) in the population of nuclear power plants (NPP) is 

ignored. The uncertainty therefore represents the level of confidence of each panel member’s estimate of 

this single true frequency.3

3. Issue 3 Response: More precisely, as stated in Section 3.8.5, the elicited quantities for each 

question (MV, LB and UB) are percentiles of each panel member’s subjective distribution. As 

stated above, this subjective distribution should consider broad plant difference related to plant 

design that can affect the LOCA frequency. Also, except for the base case frequencies used to 

anchor the results, panel members were never asked about absolute frequencies, only about 

relative frequencies of specific components, subsystems or systems. Their responses were then 

combined to form estimates of LOCA frequencies. 

Issue 4: Given that the uncertainty represents an individual’s confidence in their own estimate, what is 

the basis for automatically assuming the probability distribution associated with this confidence 

uncertainty is not symmetrical?4

4. Issue 4 Response: The distribution associated with the response to any question was assumed 

to be asymmetrical only if the stated LBs and UBs were not symmetrical about the MV. 

Issue 5: While I agree that the uncertainty associated with LOCA frequencies should be asymmetrically

weighted toward higher values; this is based on the observation that not all plants are identical, and that if 

a LOCA occurs at a plant, that plant will likely be shown to be a poor performer, not representative of the 

fleet as a whole.5

5. Issue 5 Response: As noted above and in NUREG-1829, the estimated LOCA frequencies are 

industry-wide or “generic, or average, estimates for the commercial fleet” (Section 2). 

Consequently, the uncertainties associated with the estimates pertain to these generic estimates 

and not to any individual plants whose LOCA frequencies may differ from the generic 

estimates. As stated in Section 2 of the NUREG, the panelists were instructed to account for 

broad plant-specific factors which influence the generic LOCA frequencies in providing 

uncertainty bounds, but not consider factors specific to any individual plants. Thus, the 

uncertainty bounds should include both contributions related to the uncertainty of the generic 

estimates as well as uncertainty due to broad plant-specific fleet differences. 

Issue 6: This is not the same as assuming the confidence uncertainty surrounding the estimates provided 

by the panel members, of a “single true value” should be represented with a lognormal (which was done 

in the elicitation).6

6. Issue 6 Response: The lognormal, or split lognormal when the responses are asymmetrical, is a 

reasonable distribution for representing the responses to the various questions so that the 

responses can be combined to estimate LOCA frequencies. Based on the sensitivity analyses 

conducted in Section 7, the authors expect that the log-normal distribution assumption has an 

inconsequential impact on the bottom line parameter estimates, especially in light of the large 

uncertainties observed in the final results. Note that the report does not estimate the 

distributions of LOCA frequencies, but only the four bottom-line parameters (mean, median, 

5th and 95th percentiles) of these distributions. 
M-4

Issue 7: First, the opinions of the panel members were solicited, and then modified (increased) by the 

authors.7

7. Issue 7 Response: It is misleading to state that “...the opinions of the panel members [were] 

modified (increased) by the authors”. Rather, an overconfidence adjustment was applied to 

increase only the uncertainties for those panelists whose responses indicated more confidence 

than the group average. The panelists’ median estimates were not modified. The justification 

for the overconfidence adjustment is provided in Section 5.6.2. Responses were not modified 

in any other manner. 

Issue 8: Second, the processes and analyses employed have introduced a conservative bias into the final 

base case results, with additional conservative bias

8

 inserted in the various sensitivity studies.9

8. Issue 8 Response: The only “conservative bias” that may have been introduced in the summary 

estimates was through the use of the overconfidence adjustment. However, as discussed in 

Section 5.6.2 and by reviewing the wide range of uncertainty estimates provided by the 

panelists, there is good reason to believe that at least some of the panelists may have been 

overconfident. Only those panelists that were more confident than the group median were 

adjusted. Furthermore, sensitivity studies indicated that the error factor adjustment used was a 

reasonable adjustment scheme. This method was not the most conservative adjustment scheme 

which could have been used. 

Issue 9: Second, the processes and analyses employed have introduced a conservative bias into the final 

base case results, with additional conservative bias

8

 inserted in the various sensitivity studies.9

9. Issue 9 Response: Sensitivity analyses were conducted to examine the impact of the 

assumptions used to process the results. The assumptions and methods used to calculate the 

“baseline” LOCA estimates are clearly stated in Sections 5 and 7. Additionally, the methods 

used to conduct each sensitivity analysis are clearly explained in Sections 5 and 7 and 

comparisons are made to the baseline results so that the effects on the LOCA frequency 

estimates are readily apparent. Those sensitivity analyses resulting in the largest differences 

from the baseline estimates are clearly discussed in the report. 

Issue 10: This “creeping conservatism” is not necessarily undesirable given the significant uncertainties 

and the uses to which the results will be employed; however, it should be explicitly acknowledged and 

clearly stated rather than obscured in the details of the analyses.10 

10. Issue 10 Response: The NUREG report systematically documents the assumptions and analysis 

used to calculate the LOCA frequency estimates. The NUREG also discusses the effects of 

alternative assumptions and analysis methods using sensitivity studies. The analysis 

procedures utilized in the elicitation process were fully discussed with the panelists at several 

times throughout the process: during the base case review, prior to conducting the individual 

elicitations, during the presentation of the preliminary results, and during the preparation of the 

draft NUREG. Each panelist was also provided numerous opportunities to modify their results, 

including changes if they believed the initial instructions were unclear, or if the processing of 

their results did not actually reflect their solicited opinion. None of the panelists elected to 

provide wholesale changes to their estimates based on these issues. General feedback from the 

panelists about the elicitation process in general, and the processing of the results in particular, 

was favorable. 

Comment Number: GC2 

Submitted by Joseph Conen of the BWR Owners Group 

Comment: Another issue is the inclusion of thermal fatigue as a degradation mechanism for the BWR 

feedwater line. We are not aware of any thermal fatigue issue other than the feedwater nozzles; that issue 

was resolved in the early 1980s through several mitigation measures including the installation of GE-
M-5

designed triple thermal sleeve. A rigorous inspection program per NUREG-0619 is currently in place. Not 

a single one of hundreds of these inspections have turned up any evidence of cracking. Thus, thermal 

fatigue is not an issue in the NSSS portion of the BWR feedwater line. In addition, please refer to the 

letter from T. Essig (U.S.NRC) to T. J. Rausch (BWROG), dated June 5, 1998, subject: BWROG-Safety 

Evaluation of Proposed Alternative to BWR Feedwater Nozzle Inspections (TAC M94090). This letter 

provides evidence that the thermal fatigue of the FW nozzles is effectively managed at BWRs and 

provides the basis for revising the examination frequencies. 

Response: The NUREG report acknowledges the improvements that have been implemented in BWRs 

as a result of the NUREG-0619 inspection requirements. As stated in Section 6.3.2, "There was a rash of 

feedwater nozzle cracks reported in the 1970 to early 1980 time period in BWRs. Plant and system 

modifications were implemented after a detailed study of the problem and augmented inspections are 

being conducted based on NUREG-0619 requirements. These mitigation measures have proven effective 

as no new thermal fatigue cracks have been discovered in these BWR feedwater nozzles over the last 20 

years." However, as stated in Section L.2 when discussing the relative contributions for various BWR 

piping systems, "There was wide variability expressed for the feedwater system. Several participants 

thought that its susceptibility was similar to that of the recirculation system while others thought that it 

would make an inconsequential contribution.” This latter group generally thought that the mitigation 

programs in place for the feedwater system were generally effective and additional thermal fatigue 

locations within the feedwater system were not significant LOCA contributors. 

However, rationale provided by some of the panelists who believe that thermal fatigue is a significant

contributor to the LOCA frequency estimates is summarized in Section L.2 and Section 6.3.2. As stated

in Section 6.3.2, “The BWR plants are expected to be more prone to thermal fatigue problems compared 

with the primary side of PWR plants because they experience greater temperature fluctuations during the 

normal operating cycle. In BWR plants, thermal fatigue remains an important contributor for the 

feedwater lines and the RHR system.” Additionally, “...thermal fatigue is an aging mechanism that could 

lead to a large LOCA because it does not manifest itself as a single crack, but as a family of cracks over a 

wide area. Thermal fatigue cracks also tend to propagate rapidly, and since it is not material sensitive 

(i.e., it can attack a number of materials), it is difficult to prioritize critical areas for inspections.” These 

reasons explain why thermal fatigue is still regarded as an important LOCA contributor by many 

panelists. 

The general variability in opinion expressed by the elicitation panelists for feedwater, main steam, and 

RHR systems is summarized in the box and whisker plots in Figures L.6 through L.8. Some of the 

differences associated with the illustrated variability in these figures results from the diversity in opinion 

about the significance for thermal fatigue in BWR plants. 

Comment Number: GC3 

Submitted by Nuclear Energy Institute (NEI) 

Comment: As the NUREG-1829 report may be considered to be the “most recent applicable data” upon 

finalization, it is important that the final report provide an alternative to continue using operational 

experience data for the determination of small break LOCA frequencies. Most PRAs currently reference 

NUREG-5750, which used such a basis (at the time there were 1,250 reactor years of operating 

experience) to estimate small break LOCA frequencies. Since issuance of NUREG-5750, over one 

thousand additional reactor years of operational experience have confirmed the conclusions of NUREG5750 relative to small break LOCAs. Draft NUREG-1829 notes that, when steam generator tube rupture 

data are excluded, there is general correlation on small break frequencies with NUREG-5750. However, 

our review of the report indicated that draft NUREG-1829 estimates these frequencies over one order of

magnitude higher than the estimate of NUREG-5750. Using the NUREG-1829 small break LOCA 

frequency estimation, the US reactor fleet should be experiencing one small break LOCA on average 
M-6

every 4 years. However, no such LOCAs have occurred in the operating history of the US plants. 

Obviously, the incorporation of this frequency estimate into existing PRAs would lead to unwarranted 

impacts that are out of context with reality. 

Response: Comparing results from NUREG-1829 and NUREG/CR-5750 must be done with some care 

due to differences in how the LOCA frequencies were calculated. The SB LOCA frequencies typically 

reported in NUREG-1829 for PWRs include contributions from SGTRs while the SB LOCA frequencies 

reported in NUREG/CR-5750 exclude SGTRs. Furthermore, the LOCA frequencies in NUREG-1829 are 

based on threshold leak rates (Category 1 LOCAs include all leaks greater than 100 gpm [380 lpm], 

Category 2 LOCAs include all leaks greater than 1,500 gpm [5,700 lpm], etc.). Conversely, the SB 

LOCA results in NUREG/CR-5750 include only those events with leak rates between 100 and 1,500 gpm 

(380 and 5,700 lpm). 

These distinctions were considered in making the comparisons provided in Section 7.9 that are 

summarized in Table 7.20. The reported SB mean LOCA frequency estimates for NUREG/CR-5750 are 

4.0E-04 per calendar year and 7.4E-03 per calendar year for BWR and PWR plants, respectively. Note 

that these PWR SB LOCA estimates include the steam generator rupture frequencies calculated in 

NUREG/CR-5750. The revised NUREG-1829 SB LOCA estimates are 5.2E-04 and 6.6E-03 for BWR 

and PWR plants, respectively (geometric mean with overconfidence adjustment). The ratio of the 

NUREG/CR-5750 to the NUREG-1829 results is 0.76 for BWRs and 1.12 for PWRs. These SB LOCA 

estimates are therefore similar. 

It is also interesting to compare the PWR SB LOCA frequency estimates after excluding SGTR 

frequencies from the Category 1 LOCA estimates (Table 7.19). As reported in Section 7.8 of NUREG 

1829, the SGTR rupture frequencies predicted by the elicitation, NUREG/CR-5750, and operational 

experience are consistent. The cumulative SB LOCA Category 1 and 2 estimates without SGTR 

contributions are 1.9E-03 and 4.2E-04 per calendar year (Table 7.19). Therefore, the interval value that 

corresponds to the historical SB LOCA definition (i.e., breaks between 100 and 1,500 gpm [380 and 

5,700 lpm]) is 1.48E-03 per calendar year. This value is approximately 3.7 times higher than the mean

SB LOCA frequency from NUREG/CR-5750 of 4E-04 per calendar year. 

The increase in the elicitation result reflects the panelists’ opinion that current PWR SB LOCA 

frequencies for components other than steam generator tubes are higher than historical averages. This

increase primarily stems from current PWSCC concerns (Section 6.3.2). The practical implications from 

these differences however are not striking. The NUREG/CR-5750 estimate translates to an expected 

PWR SB LOCA every 36 years for the current fleet of 69 operating PWRs. The elicitation frequency 

corresponds to an expected PWR SB LOCA every 10 years. Additional comparisons between the 

NUREG-1829 and NUREG/CR-5750 estimates are contained in Section 7.9. 

Because of this and similar comments, Section 7.10 has been added to NUREG-1829 to compare the 

elicitation LOCA frequency estimates with estimates derived from operating experience. This 

comparison shows that the differences between the elicitation and operating experience-based estimates

of the non-SGTR, PWR Category 1 LOCA frequencies are not statistically significant. However, even 

though these differences are not statistically significant, this does not imply that the two approaches are 

estimating the same frequency. Because an operating experience-based estimate is an historical average 

based on many years of operation, a difference will exist if the panelists believe that the current failure 

frequency differs from the historical average. In fact, the increased elicitation estimate is supported by 

the panelists’ qualitative and quantitative responses. As noted above, the panelists indicated that medium 

and, to a lesser extent, small LOCAs in PWRs are most dramatically impacted by PWSCC in relatively 

small diameter passive system component (e.g., CRDMs, instrument nozzles, etc.) (Section 6.3.2). 
M-7

Additional details on this comparison are found in Section 7.10. Related information is also found in the 

responses to GC4, GC5, GC6, GC7, and 7-8. 

Comment Number: GC4 

Submitted by Nuclear Energy Institute (NEI) 

Comment: The Nuclear Energy Institute offers the following comments on the subject Federal Register 

notice, which solicited public comments on draft NUREG-1829. This NUREG was intended to provide 

technical support for the proposed rulemaking to 10 CFR 50.46 which would establish the option to revise 

the design basis LOCA break size. Thus, the emphasis of the expert elicitation was on estimating 

frequencies for large break LOCAs. Our comments are limited to the report’s estimation of small break 

LOCA frequencies, which, unlike large break LOCAs, are important contributors to PRA risk profiles. 

PRA standards have been developed by consensus bodies and endorsed by NRC, with the expectation that 

plants will be expected to conform to these standards to support regulatory applications. ASME PRA 

standard RA-S-2002 (endorsed through NRG Regulatory Guide 1.200) contains the following 

requirement relative to initiating event frequency estimation: 

IE-C1: Calculate the initiating event frequency from plant-specific data, if sufficient data are 

available. Otherwise, use generic data. Use the most recent applicable data to quantify the 

initiating event frequencies..... 

Response: Contrary to the commenter’s assertion, the objective of the expert elicitation (Section 2) was 

to calculate LOCA frequencies for all break sizes, not just large break LOCAs. Also, in general, the 

objectives and results of the expert elicitation as summarized in NUREG-1829 are consistent with the 

above statement in ASME PRA standard RA-S-2002. The NUREG-1829 small break LOCA estimates 

represent generic values. Calculations based on plant-specific data, specifically for PRA, are acceptable 

for use provided that a sufficient technical basis supporting these alternative estimates has been 

established. There is also an implication by this comment (and subsequent Comment GC5) that the 

NUREG-1829 estimates do not compare favorably with operator experience. As documented in the 

newly-added Section 7.10, the differences between the NUREG-1829 estimates and operating experience 

are not statistically significant. Those differences that do exist are also supported by the panelists’ 

qualitative and quantitative responses. More details on the comparison between operating experience and 

the NUREG-1829 LOCA frequency estimates are provided in Section 7.10 which was added to address 

this and similar comments. See also the responses to GC3, GC5, GC6, GC7, and 7-8 for similar 

discussions. 

Comment Number: GC5 

Submitted by Nuclear Energy Institute (NEI) 

Comment: Draft NUREG-1829 used plant experiences to estimate the steam generator tube rupture 

(SGTR) frequency which amounts to greater-than 50% of the total small LOCA frequency. Estimate of 

the remaining 50% of Category 1 LOCA was entirely based on expert elicitation. The resulting Category 

1 frequency estimates from the panel showed a significant divergence of opinions. It is recommend that

Category 1 LOCA frequency estimate should continue to be related to the large number of years of plant

experiences similar to the method used in NUREG-5750. The current lengths of those experiences 

amount to thousands of reactor-years, and are statistically significant for use in estimating the annual 

frequency of events at the 1E-2, 1E-3, and 1E-4 levels. Similar estimates are used in PRA models for 

numerous other important PRA parameters (such as SGTR). 

Response: The commenter incorrectly states that plant experiences were used to estimate the SGTR 

frequency while the remaining contributions were based on expert elicitation. All contributions to the 

Category 1 LOCA frequency estimates were determined by the expert elicitation process. For the reasons 
M-8

documented in Section 1 of NUREG-1829, the authors believe that the expert elicitation results are more 

accurate than results calculated simply from operational experience. 

However, the SGTR contributions provided by the panelists are nearly identical to estimates determined

simply from operational experience. This consistency, however, simply reflects the panelists’ combined 

opinion that the operational experience data is relevant for calculating the current day SGTR 

contributions. The contributions of non-SGTR failures to the PWR LOCA Category 1 frequencies and 

the BWR LOCA Category 1 estimates are also not inconsistent with estimates based solely on operational

experience (See Section 7.10). However, the panelists do expect some increase in the PWR LOCA 

Category 1 frequencies compared to operational experience-based estimates as a result of current PWSCC

issues (Sections 6.3.2 and 7). See Section 7.10, which was added to address this and similar comments, 

for additional details on the comparison between the elicitation estimates and operational experience data. 

Additional relevant information is also provided in the responses to GC3, GC4, GC6, GC7, and 7-8. 

Comment Number: GC6 

Submitted by Nuclear Energy Institute (NEI) 

Comment: The report should provide a discussion on statistical validation of small LOCA frequency. By 

using the method of Jeffrey’s non-informative prior (over the past 2,500 reactor years with zero events 

excluding steam generator tube ruptures), the expected small LOCA frequency is at or below the 1E-04 

level. This frequency is over one order of magnitude lower than the frequency reported in the draft 

NUREG. Plant operational experience of over 2,500 reactor-years should be considered as a valid 

predictor of small LOCAs. That consideration is further strengthened by improved methods and increased

requirements for in-service-inspections and leak detections. 

Response: The use of Jeffrey’s non-informative prior to estimate a frequency if no events have been 

observed leads to a significant underestimate of the mean frequency. Using this method, NEI states that 

the expected frequency is no more than 1E-04, based on a denominator of 2,500 reactor years. With zero 

events, a reasonable estimate of the mean is based on using a 50% confidence bound or a value of 0.7 in 

the numerator. This yields 0.7/2,500 = 2.8E-04 which is a factor of 3 larger than the estimate using 

Jeffrey’s non-informative prior. Furthermore, the use of 2,500 reactor years as the denominator clearly 

combines both PWR and BWR operating experience. More rigorous evaluation of operational experience 

data (See Section 7.10) has demonstrated that the operational experience-based and elicitation-based SB 

LOCA estimates are not inconsistent. Furthermore, differences are supported by the panelists’ responses 

and rationale. Section 7.10, which was added to address this and similar comments, provides additional 

details on the comparison between the elicitation estimates and operational experience data. Also, see the 

responses to GC3, GC4, GC5, GC7, and 7-8 for related information.

Comment Number: GC7 

Submitted by Nuclear Energy Institute (NEI) 

Comment: The report should provide a discussion on probabilistic validation of the small LOCA 

frequency. Using a Poisson distribution with failure rate of 2.9E-03 (NUREG-1829 Category 1 LOCA 

frequency, excluding steam generator tube rupture events), and considering the approximately 2,500 

reactor-years of operation experience, the probability of no small LOCA events (actual industry 

performance) is around 1 percent. This result shows an excessive conservatism in the category 1 LOCA 

frequency estimation of NUREG-1829. 

Response: A new section (Section 7.10) has been added to the revised NUREG which compares the 

small break LOCA frequency estimates from the elicitation with operating experience. As discussed in 

this section, the elicitation estimates for BWR and PWR small break LOCA frequencies are generally 
M-9

consistent with the historical data. See the response to Comment 7-8. Related insights are also provided 

in the responses to Comments GC3, GC4, GC5, GC6, and 7-8. 

Two additional points should be noted. First, the 2,500 reactor-years of operating experience referred to 

in the comment clearly combines BWR and PWR plants. The NUREG provides separate estimates for 

BWRs and PWRs and treats their operating experience separately. The number of reactor-years used in 

Section 7.10 is 1,023 for BWRs and 1,986 for PWRs, all as of December 2006. Second, as discussed in 

Section 7.10, there has been one pipe break in a PWR plant that exceeded the LOCA Category 1 flow rate

threshold defined in this study. Additionally, a second PWR pipe break resulted in flows that were close 

to the Category 1 threshold. The comment claims that there have been no non-SGTR Category 1 events. 

Comment Number: GC8 

Submitted by Professor Larry Hochreiter of Penn State University 

Comment: We have revised the data for pipe breaks and leaks following the suggestions from the NRC 

staff and have made comparisons to the NRC expert solicitation panel's break spectrum failure frequency 

plots that were developed in support of the revisions to 10CFR50.46. We had discussion with C. Gary 

Hammer and other members of the NRC staff on our original comparisons and they suggested that we 

limit our comparisons to just the class 1 piping for the reactor since that was the class of piping systems 

that the NRC break spectrum frequency study were originally developed for. Therefore, we eliminated all 

piping failure, leaks, and cracking data from the mixed data base that we had originally developed such 

that only the class 1 piping information remained. A CD with the revised data is included with this letter. 

We also reviewed the class 1 systems with the NRC to make sure that we isolated only those systems 

which should be compared to the estimated break frequency curves. When comparing the remaining data 

to the NRC break frequency curves, we used approximately the same break size bins that were used in the 

NRC study. We also normalized the individual bins of data using the approach recommended by the NRC 

in which the number of breaks or breaks and leaks were normalized on the numbers of effective full 

power days. 

We have separated out "Breaks" from the total data base as separate plots for both PWRs and BWRs as 

shown in Figures 1 and 2. As these figures indicate, there are no large breaks in the class 1 piping. 

However, for the smaller breaks, the data clearly lies above the estimated break frequencies estimated in 

the NRC solicitation study. The data is above the 95% percentile as compared to the NRC break 

frequency plots indicating that even smaller breaks are more frequent as compared to the estimated 

frequencies from the solicitation study. 

We also have combined "Breaks" and "Leaks" together on separate plots for PWRs and BWRs as seen in 

Figures 3 and 4. The rational for this is that leaks are really breaks since the pipe has failed and is leaking. 

We used the size of the pipe for when combining the leaks with the breaks. This may not be the best 

method of comparing the data to the break frequency distribution and it may appear to bias the results

since there are only leaks for the larger pipes and not breaks. However, it is not clear that, given a set of 

conditions, a leak in a large pipe could grow to a full pipe size break. Say for example, under a safe

shutdown earthquake condition. Also, this grouping could be viewed as being conservative since the 

original premise is that the pipes should not leak in the first place. 

All the data, breaks and leaks, when combined in this form indicate that there is a significant difference 

between the existing data and the break spectrum failure frequencies from the NRC expert solicitation 

study. This indicates to me that we should not be revising 10CFR50.46 by introducing a "transitional 

break size" and reducing the mitigation capabilities of the plant's ECC systems and defense in depth for 

the larger break sizes. I believe that we need to maintain the full functional capability of the robust ECC 

systems that we have in the current operating plants for all possible break sizes up to and including the 

full double-ended break of the reactor coolant piping. The break and leak data for the class 1 piping 
M-10

indicates to me that we need the full functional ECC systems and their supporting systems and they 

should not be compromised or sacrificed by revising 10CFR50.46. 
M-11
M-12

Response: There are several issues with the analysis performed by the commenter with respect to both 

the pipe “break” frequencies plotted in Figures 1 and 2 and the pipe “break and leak” frequencies in 

Figures 3 and 4. Figures 1 and 2 above were created by identifying “breaks” from a database of piping

failures to construct separate PWR and BWR break frequencies. As depicted in Figures 1 and 2 above, 

the resulting plot clearly lies above the estimated break frequencies estimated in the expert elicitation 

documented in NUREG-1829. 

The biggest issue pertaining to the commenter’s results is with the integrity of the database used in the 

analysis which was supplied as part of the commenter’s public comment. The database appears to be 

similar or identical to the database developed in Bush, et. al, “Piping Failures in US Nuclear Plants: 1961-

1995,” SKI 96.20. This can be surmised from the fact that there are only two events as recent as early 

1996 in the database, and the database contains the similar erroneous information to the Bush et. al. report 

for selected audited events. The Bush et. al. database was never validated and it was also the foundation 

of EPRI TR-111880. An independent review of this database1

 found that a large percentage of the 

database records involved non-piping failures. The EPRI report was subsequently withdrawn due to 

requests by EPRI members. 

The authors reviewed the larger pipe events that could be considered “breaks” and used in constructing

Figures 1 and 2. The authors examined events classified as a breakage, rupture, failure, or severance in 

the database for 2 inch diameter or greater pipes in both BWR and PWR plants. This approach was 

necessary because the commenter neither definitively identified the data used to construct Figures 1 and 

2, nor provided details supporting the calculations. Events were checked against the OPDE Database 

Rev. 0.e, dated 24 March 2004 which has been verified. In several cases, the original failure references 

were obtained for verification. 

 

1

Lydell, B., “Independent Review of SKI 96-20 Database,” Technical Note 1996-01, SKI Ref. No. 14.2-940477, February 1996.
M-13

Nineteen events were reviewed in all and this data verified many of the reported problems with the Bush 

et. al. database. Almost all of the database records contain some error or inconsistency. Many of the 

listed events cannot be referenced to a known event. Common problems include incorrect event dates, 

references, pipe sizes, or break sizes. The failure classification type (i.e., leak, rupture, severance, etc.) 

was also found to be both inconsistent and inaccurate in several cases. In spite of these problems, the 

authors searched the OPDE Database for events that likely correspond to the break events identified in the 

commenter’s database. Approximately 15 events were identified which either could be definitively 

matched to an event in the commenter’s database or at least represented by a pipe break at the listed plant 

in a similar system. None of break events occurred in an unisolable segment of the reactor coolant 

pressure boundary piping. This is the fundamental definition of a LOCA, and the objective of the 

elicitation was to estimate LOCA frequencies. The only basis for comparison with the elicitation results 

is an estimation of LOCA frequencies. No other basis, including a comparison of piping failure 

frequencies, is consistent with the elicitation results. Hence, the Figure 1 and 2 comparisons with the 

elicitation results are invalid without further consideration. 

However, there are other inconsistencies and errors in the commenter’s analysis. The largest “failed”

PWR piping in the database is for a 4 inch diameter pipe. Leak data is reported in the commenter’s 

database for PWR piping up to 13 inch diameter. However, Figure 1 above depicts breaks up to 12 to 15-

inch diameter. The basis for reporting PWR break data greater than 4 inch diameter is not supported by 

the supplied database. There are other, less significant, but still consequential problems with the 

commenter’s analysis. Contrary to the commenter’s statement that all the breaks and leaks represent 

“Class 1 piping” systems, most of the events occurred in lower grade piping. This is not a trivial 

distinction because Class 1 piping is subject to more rigorous design, testing, and inspection requirements 

than other piping systems within the plants. In addition, the commenter’s database contains nearly twice 

the number of carbon steel failures in PWR plants than in BWR plants. The finding is heavily biased by 

the consideration of non-ASME Code, FAC-susceptible piping in the commenter’s database. Also, 

several of the rupture sizes appear to overestimate either the actual pipe size or the rupture size which 

occurred. The distinction is important because the size of the rupture, not the piping size, determines the 

rate of loss of system fluid. A partial failure of a large pipe can lead to substantially lower fluid loss rates 

than expected if the pipe completely burst. 

The leak data comparisons presented in Figures 3 and 4 are similarly misleading. The most substantive

issue is the commenter’s contention that “leaks are really breaks since the pipe has failed and is leaking.” 

This is simply untrue. Pipe breaks occur once a flaw reaches a critical size such that it grows unstably at 

the applied load level. The piping can then not support required operational loading, a readily-apparent 

hole or breach forms, and internal fluid is released at a rate beyond the reactor water make-up system of 

the plant. Conversely, piping having leaking joints or through-wall cracks typically remains intact and 

continues to support internal and external stress without imminent failure. These cracks may continue to 

grow until imminent failure occurs, but the difference between the leaking crack size and the rupture 

crack size increases with pipe size. Hence, larger pipes provide more margin against failure after a leak 

appears. The NRC’s LBB approach for large pressure boundary piping is based on the premise that a 

leaking through-wall flaw can be detected at normal operating loads via the plant’s normal leakage 

detection systems before the flaw reaches a size that could grow unstably at emergency or faulted load

conditions. The commenter’s database records support this premise as the larger diameter, Class 1 piping 

events are leaks that were found and repaired prior to a catastrophic failure. The LBB approach has used 

to ensure compliance of GDC-4 requirements for over 20 years. 

Additionally, the commenter’s analysis of the leak data using the supplied database suffers from similar 

issues as the break data. A majority of these leak events actually occurred in secondary-side plant 

systems and the pipe leak did not result in loss of pressure boundary coolant. Therefore, longer term
M-14

failures of these pipes (presuming that they remained undetected and the associated flaw continued to 

grow to failure) would not have resulted in a primary LOCA event. 

In contrast to the commenter’s analysis, a review of some of the more well-established and wellrecognized piping failure databases (PIPEex, OPDE, and SLAP) has found 1 event that could be 

characterized as a SB LOCA in PWR plants and no events that could be characterized as SB LOCAs in 

BWR plants (See Section 7.10 of NUREG-1829). No MB or LB LOCA events have occurred. As 

previously discussed (See responses to GC3, GC4, GC5, GC6, and GC7) the LOCA elicitation results for 

the present day estimates are consistent with this operating experience. The comparison of the elicitation 

results and operating experience is also provided in Section 7.10 of the revised NUREG report. 

Comments Related to Executive Summary of NUREG-1829

Comment Number: ES1 

Submitted by Westinghouse Owners Group 

Comment: Method of Aggregating Individual Frequency Responses into a Group Response. For this 

study the geometric mean aggregation method was used instead of the arithmetic mean method or mixture 

distribution methods, which would give higher mean values of LOCA frequency. The reviewers concur 

that the geometric mean is most representative of the consensus of the group (expert panel). As an 

example, consider possible individual responses the different group means for the distribution of factors 

on a baseline frequency in the following table (see WOG response for table): 

Number or Responses Value of Factor 

1 0.01 

2 0.1 

3 1.0 

2 10 

1 100 

For this example, the arithmetic mean value for the 9 responses is 13.69 while the geometric mean value 

(average of the logarithms) is 1.0, which seems to be much more representative of the group's opinions. 

Part of the reason for this is that the probabilities and frequencies of failures of structural components, 

such as piping, are normally expressed as orders of magnitudes much less than one. Uncertainties on these 

values are also expressed as factors instead of differences because the physical contributions to structural 

failures (leaks and breaks), such as flaw sizes and crack growth rates, are also known to be log-normally 

distributed. Use of logarithmic distributions and geometric means is also consistent with NRC Guidance

on Risk-Informed In-service Inspection for Piping (Draft Report NUREG-1661, January 1991). Figures 

3.3 and 3.4 of this guidance show the range of frequency estimates from expert elicitation, plotted on a 

logarithmic scale, for failure of auxiliary feedwater system components and the reactor pressure vessel, 

respectively. Figure 4.6 of this same report shows the uncertainty in the best estimate (median value) of 

piping failure probability, calculated using probabilistic fracture mechanics methods, to also be 

logarithmically distributed. Scanned copies of these figures are provided in Appendix A2 of the WOG’s 

public comment document. 

Response: The authors agree that the geometric mean aggregation supports the elicitation objective of 

developing a consensus group estimate by reducing the effect of either wide differences in the individual 

 

2

Letter from Frederick P. Schiffley, II, Chairman to USNRC, “Westinghouse Owners Group Comments on Draft NUREG-1829, 

‘Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Expert Elicitation Process’ (MUHP-3062)”, WOG-05-

517, dated November 28, 2005, ADAMS Accession # ML0503340274. 
M-15

estimates, or the effect of a single estimate which is significantly higher than the others. As documented 

in Section 7, geometric-mean aggregation, in this study, produces group estimates which approximate the 

median of the individual estimates. In the example cited by this commenter, the arithmetic mean estimate 

of 13.7 is larger than all but one of the responses and does not approximate the median of the individual 

estimates. 

However, as indicated in the Executive Summary, because the alternative aggregation methods can lead 

to significantly different results, a particular set of LOCA frequency estimates is not generically 

recommended for all risk-informed applications. The purposes and context of the application must be 

considered when determining the appropriateness of any set of elicitation results. This position is 

consistent with the recommendation of the NRC’s Advisory Committee on Reactor Safeguards (Letter 

from W.J. Shack to D.E. Klein dated December, 20, 2007, Subject: Draft Final NUREG-1829, 

“Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process,” and Draft 

NUREG-XXXX, “Seismic Considerations for the Transition Break Size”, ADAMS Accession Number 

ML073440143). 

Comment Number: ES2 

Submitted by Joseph Conen of the BWR Owners Group 

Comment: It is apparent that the panel has not given appropriate credit to the IGSCC mitigation 

measures for the NSSS stainless steel piping that the BWR owners have implemented since the early 

1980s. For example, the second paragraph from bottom on page xvii states, in part: “...the biggest 

frequency contributors for each LOCA size tend to be systems having the smallest pipes, or component, 

which can lead to that size LOCA. The exception to this general rule is the BWR recirculation system, 

which is important at all LOCA sizes due to lingering IGSCC concerns.” Since the largest pipe size in

the recirculation piping system can be up to 28-inches, the preceding statement essentially implies that LB 

LOCA redefinition is not applicable to BWRs. The panel did not seem to give adequate credit for several 

effective mitigation measures in terms of better material (e.g., use of nuclear grade stainless steel in 

replacement lines), stress improvement (e.g., induction heating stress improvement [IHSI], last pass heat 

sink welding [LPHSW], and mechanical stress improvement process [MSIP]) and water environment 

(e.g., hydrogen water chemistry [HWC]) and repair measures such as the weld overlays and elimination 

of creviced geometries. The panel apparently did not consider the report GE-NEA41-00110-00-1, Rev. 0, 

A Review of NUREG/CR-5750 IGSCC Improvement Factor and Probability of Rupture Given a ThroughWall Crack, provided to NRC by the BWROG on April 25, 2002 (ADAMS Accession NO. 

ML021210417). In addition, the panel did not recognize the contribution of BWRVIP-75, which provides 

evidence that IGSCC is effectively managed at BWRs and provides the basis for revising examination 

frequencies. We consider these significant oversights, given their relevance to the panel’s conclusions. On 

the other hand, the panel did accept the future effectiveness of mitigating measures for PWSCC issue for 

the PWR small diameter piping (p. 6-5) in reducing failure rates for this piping. The NUREG should 

provide similar credit for the BWR IGSCC mitigation measures noted above with regard to break 

frequencies. 

Response: The authors disagree with the contention that the panel has not given appropriate credit to the 

IGSCC mitigation measures for the NSSS stainless steel piping that the BWR plants have implemented 

since the early 1980s. The report referenced above (GE-NEA41-00110-00-1) was provided as 

background documentation to the peer review panel. Additionally, several of the panelists had extensive 

experience with the assessment of the IGSCC issue and the development of appropriate mitigation 

strategies. One of the BWR base cases specifically investigated the failure probability of primary 

recirculation piping due to IGSCC. The base case model plant was assumed to follow the Generic Letter

88-01 inspection technique, used weld overlay to reinforce the flawed piping, and utilized normal water 

chemistry. 
M-16

This base case definition was chosen for convenience to improve the expected accuracy of the PFM 

analysis to be conducted by only considering one easily modeled mitigation strategy, i.e., weld overlay. It 

was well recognized and stressed during meetings that this base case was generally not representative of 

BWR plants. It was noted that most BWR plants employ hydrogenated water chemistry and may employ 

mitigation strategies other than weld overlays. Additional sensitivity analyses of the base case results 

were conducted to evaluate the effectiveness of BWR mitigation strategies for IGSCC. These sensitivity 

studies are summarized in Sections 4.3.3 and 4.3.4 of NUREG-1829, while more detail is provided in 

Appendices D – F. These results were presented to the panelists during the base case review meeting and 

copies were available to support the preparation of their individual elicitations. This information, the 

background information provided, and the experience of individual panelists was sufficient to ensure that 

the panelists were sufficiently informed about the effectiveness of IGSCC mitigation so that it was 

properly credited during the elicitation. 

However, although the mitigation has been effective in reducing the associated LOCA frequencies, there

is still risk associated with failure of BWR systems containing pre-existing flaws. As summarized in 

Section 6.3.2 of NUREG-1829, “The panel consensus is that the susceptibility of BWR piping systems to 

IGSCC is greatly reduced compared to what it was in the past. Measures such as improved HWC, weld 

overlay repairs, stress relief, and pipe replacement with more crack resistant materials have led to this 

reduction. Inspection quality has also improved such that the probability of crack detection is much better 

than in the past. However, as indicated earlier, there remains concern about the failure likelihood of the 

large recirculation piping and the RHR lines that have not been replaced. The original piping materials 

are much more susceptible to IGSCC and many lines retain preexisting cracks that initiated and grew 

before HWC was adopted.” 

Additionally, at least one panelist was also concerned that the more IGSCC-resistant replacement piping 

materials may still crack under service conditions. This panelist cited the German plant experience with 

cracking in Type 347 stainless steel. Another panelist raised the possibility that cold work (e.g. due to 

grinding) could increase the IGSCC susceptibility of the low carbon (L grade) stainless steel that has been 

used as a replacement material in many U.S. plants. However, the U.S. BWR experience with L grade 

stainless steel piping was widely recognized by the panel as being very good thus far. For these reasons, 

many panelists believe that continued vigilance is required through the augmented inspection 

requirements in Generic Letter 88-01 and NUREG-0313. 

Key elements of this response have been used to modify the Executive Summary and Sections 3.5, 4.3, 

and 6.3 of the revised NUREG. 

Comment Number: ES3 

Submitted by Joseph Conen of the BWR Owners Group 

Comment: Table 1 on page xxi and Table 3.8 mention effective break size of 42 inches for BWRs. It is 

likely to be an artifact of an assumed LOCA size of 500,000 gpm and not representative of the BWR 

NSSS geometries. 

Response: The LOCA categories were defined in terms of "threshold" flow rates, see Table 3.2. A 

Category 6 LOCA was defined as a LOCA which resulted in a flow rate greater than 500,000 gpm 

(1,900,000 lpm). For BWR thermal/hydraulic conditions for a liquid (not steam), this corresponded to an 

effective break size of 41 inches (1040 mm) assuming a circular opening (see Table 3.8). Similarly a 

Category 5 LOCA (>100,000 gpm [380,000 lpm]) would result in a 18 1/2 inch (470 mm) circular 

opening for BWR liquid conditions. The panel members recognized the fact that when considering BWR 

piping systems, the largest piping system in the plant (~28 inch diameter recirculation system) would not 

support a Category 6 (>500,000 gpm [1,900,000 lpm]) LOCA. However, a significant RPV rupture 
M-17

(classified as a non-piping failure) could result in a Category 6 LOCA. Table 1 combines both piping and 

non-piping LOCA contributions which is why this size LOCA is possible for a BWR plant. 

The sources of BWR Category 6 LOCA contributions have been clarified in Sections 3.4.1 and 3.7 of the 

revised NUREG. 

Comments Related to Section 1 of NUREG-1829

 

None 

Comments Related to Section 2 of NUREG-1829

 

None 

Comments Related to Section 3 of NUREG-1829

Comment Number: 3-1 

Submitted by Joseph Conen of the BWR Owners Group 

Comment: Tables 3.7 and B.1.9 depict the assumed conditions for the base cases. For the BWR 

recirculation line, the assumed plant water chemistry condition is NWC. This is not representative of the 

current US BWR fleet where most of the plants are operating on HWC. It is not clear to this reviewer if 

the panel has factored in the improvements in reactor water conductivity, irrespective of whether the plant 

is on NWC or HWC, which most BWR plant owners have put in place in the last decade. 

Response: The authors agree that most BWRs no longer operate with normal water chemistry and that 

there have been improvements in reactor water conductivity. However, the purpose of the base case 

analyses was to provide anchoring points as a basis for panel members’ elicitation responses. The panel 

members accounted for improvements in water chemistry through their adjustment ratios which were 

used to modify these base case frequencies. 

Additionally, the operating-experience base case estimates did directly consider the effect of IGSCC 

mitigation on the cracking rates. Bill Galyean’s estimates were developed by only considering post-1985 

operating experience so that IGSCC cracking frequencies were not overestimated. The year of 1985 was 

used in the base case analyses to demark the period where wholesale mitigation methods were adopted by

the industry. See Appendix E for more details. Bengt Lydell’s estimates also considered the effects of all 

mitigation techniques (e.g., HWC, weld overlays, etc.) on post-1985 through wall IGSCC cracking 

frequencies by using Bayesian analysis to determine the effect of mitigation techniques. See Appendix D 

for more details. As a result of their similar consideration of operating experience, these BWR 

recirculation system LOCA Category 1 base case results are within approximately a half order of 

magnitude of each other. See Section 4.3.4 for a more detailed discussion on IGSCC mitigation for these 

base case analyses. 

Additional base case sensitivity analyses also examined the effect of a weld overlay (another common 

IGSCC mitigation technique) on one of the PFM base case analyses. This effect is documented in 

Section 4.3.3. Additionally, the general effectiveness of all IGSCC mitigation measures was provided to 

the panelists by analyzing the operating experience both prior to and after 1985. The pre and post-1985 

cracking frequencies are discussed in Section 4.3.4. 

Key elements of this response have been incorporated into Sections 3.5.1 and 4.3.4 of the revised 

NUREG. 
M-18

Comment Number: 3-2 

Submitted by Joseph Conen of the BWR Owners Group 

Comment: For the BWR feedwater line base case, Tables 3.7 and B.1.9 mention assumed water 

chemistry condition as NWC and include flow-assisted corrosion (FAC) as an aging mechanism. This 

reviewer believes that the FAC is a potential issue in BWR feedwater lines only when the oxygen level is 

very low (e.g., few ppb) – a possibility with HWC. During the NWC condition assumed in the base case, 

the oxygen level is high enough that FAC is not likely to be an issue. Also, most BWR plants with HWC 

have implemented controls to maintain a certain minimum oxygen level in the incoming feedwater to 

mitigate likelihood of FAC. 

Response: Several elicitation panelists largely echoed the sentiment in this comment. As summarized in 

Section 6.3.2 of NUREG-1829, most panel members believe that the industry has inspection programs in 

place today to prevent FAC, especially in the primary side piping systems. However, one panel member 

expressed the concern that the water chemistry improvements (HWC) that mitigate IGSCC could lead to 

FAC in unanticipated locations that are not monitored as part of these inspection programs. The 

elicitation responses generally reflect the group expectation that FAC is not as significant a contributor to 

the LOCA frequency contribution as is IGSCC and other mechanisms. Typically panel members 

estimated that the FAC LOCA frequency contribution is between 1⁄2 to 1 order of magnitude less than the 

contribution from the recirculation piping. Accordingly, the contribution of FAC to the feedwater system 

does not significantly contribute to the BWR LOCA frequency estimates. 

Comment Number: 3-3 

Submitted by Nuclear Energy Institute (NEI) 

Comment: The equivalent break diameters used in various PRA models that form boundaries between 

various LOCA categories do not necessarily match those used in the draft NUREG. For example, a small 

LOCA range may extend up to 0.03 square feet equivalent break area (derived through existing capability 

of high pressure safety injection). The draft NUREG should give a clear guideline on interpolations 

between the various LOCA frequency values, including advice on arithmetic or geometric preference for 

interpolation. 

Response: Because LOCA frequencies were estimated only for the six specified LOCA categories, 

interpolation between them may be problematic. In the figures, the plotted points were connected with

straight lines for visual clarity, but this should not be taken to mean that linear interpolation is 

recommended. Interpolation of frequencies between category sizes can be done at the user’s discretion

depending on the conservatism required by the application. Some common interpolation schemes are 

linear, multi-point nonlinear and cubic spline. A step-wise or stair-step interpolation between two 

categories where the frequency for the lower category size is used for all flow rates or corresponding

break sizes between the two categories provides the most conservative interpolation scheme. Note that

any interpolation scheme does not reflect the uncertainty in the interpolated frequencies.

These guidelines were added to the Executive Summary and to the introduction to Section 7 of the revised 

NUREG. 

Comments Related to Section 4 of NUREG-1829

Comment Number: 4-1 

Submitted by Joseph Conen of the BWR Owners Group 

Comment: Table 4.1 shows a six-order of magnitude difference between the PFM and the field history 

estimates of through-wall cracking frequencies for the BWR-2 base case. Although the report suggests 
M-19

that service history data could be analyzed to resolve this difference, it is not clear if this was actually 

done. This also points out to the need for a rigorous examination of the statistical methods used to 

translate field leak data or service experience data into pipe break frequencies. 

Response: The complexity of translating field leak data or operating-experience data into pipe break 

frequencies was part of the rationale for conducting the elicitation (Section 1). It was also a major

consideration for performing multiple base case analyses so that results from different approaches could 

be compared. The small break LOCAs and through-wall cracking frequencies represent the categories 

where operating experience and PFM comparisons are most meaningful because the least extrapolation is 

required. For the small break LOCAs (Category 1), the results are typically within 2 orders of magnitude 

for most of the base case team members. However, the total spread in both the through-wall cracking 

leak frequencies (Figure 4.1) and SB LOCA frequencies (Figure 4.2) can be greater than three orders of

magnitude. As the commenter indicates, the BWR-2 Category 0 frequencies (Figure 4.1) vary 

significantly (4-5 orders of magnitude in final NUREG) 

Furthermore, while the PFM and the operating-experience predictions of the base case analyses did not 

always readily agree, this simply reflects the current uncertainty in calculating these estimates. 

Differences among these estimates reflect the different assumptions and approaches used in the various

analyses. Specifically, the BWR-2 operating-experience frequency estimate was based on 20 reported 

incidents in BWR feedwater systems. However, none of these events resulted in through-wall leakage. 

Therefore, the operating-experience estimate is undoubtedly conservative. 

The elicitation reported in NUREG-1829 did not attempt to resolve these discrepancies; rather, the 

elicitation’s goal was to reflect the scientific uncertainty in the technical community. Therefore, the 

purpose of the base case estimates was not to obtain convergence among the predictions. Rather, it was to 

understand how the assumptions and approaches associated with particular analyses contributed to the 

disparity in the results. These analyses and their differences were clearly presented to the elicitation 

panelists so that they could judge their effect as part of their elicitation. As a result, most panel members 

chose to anchor their current-day LOCA Category 1 responses using operating experience and used PFM, 

if at all, to inform their assessments for LOCA sizes greater than Category 1 or for future LOCA 

frequency estimates. These are the domains where relevant operational experience data does not exist.

Key elements of this response were incorporated into Section 4.2 of the revised NUREG. 

Comments Related to Section 5 of NUREG-1829

 

None 

Comments Related to Section 6 of NUREG-1829

None 

Comments Related to Section 7 of NUREG-1829

Comment Number: 7-1 

Submitted by Zouhair Elawar – Palo Verde Nuclear Generating Station 

Comment: Draft NUREG-1829 used plant experiences to estimate the steam generator tube rupture 

(SGTR) frequency which amounts to greater than 50% of the total small LOCA frequency. Estimate of 

the remaining 50% of Category 1 LOCA was entirely based on expert elicitation. The resulting Category 

1 frequency estimates from the panel showed a significant divergence of opinions. I strongly recommend
M-20

that Category 1 LOCA frequency estimates should continue to be related to the large number of years of

plant experiences similar to the method used in NUREG 5750. The current lengths of those experiences 

amount to thousands of reactor-years. They are statistically significant to be used to estimate the annual 

frequency of events at the 1E-2, 1E-3, and 1E-4 levels. Similar estimates are used in PRA models for 

numerous other important PRA parameters (such as SGTR). 

Response: Operating experience becomes more relevant as the LOCA size decreases, especially for 

events such as SGTR, because the database is populated with actual events. However, there is still scant 

data on SB LOCAs due to limited passive piping or non-piping component failures. The panelists were 

made aware of this operating experience data, and SGTR was specifically used as a base case estimate, 

along with frequencies for other piping and non-piping precursor events (e.g., cracking). However, one of 

the cautions with using operating experience solely to calculate Category 1 LOCA frequency estimates is 

that past experience is not necessarily indicative of current or future performance. Common cause material 

degradation can result in systematic increases in generic frequencies as a function of time compared to 

frequencies based on operating experience. Conversely, wholesale mitigation programs (as with IGSCC) 

can result in systematic decreases in LOCA frequencies over time. 

These common cause considerations are one reason why elicitation is preferable to simple operating 

experience estimates for even small break LOCAs. In fact, the elicitation results are most justified for this 

LOCA category because of operating experience data. The operating experience data was fully considered 

by the panelists to estimate past LOCA frequencies, and they then used their best judgment to modify these 

estimates, as appropriate, based on their current knowledge. It is preferable to use current knowledge in 

this way rather than wait for several years for operating experience to adequately reflect the new 

conditions. It is worth noting, however, that the elicitation results and operating experience data are 

generally consistent for LOCA Category 1. The only differences, which are not statistically different, are 

for the frequencies associated with non-SGTR, Category 1 PWR failures. The panelists justified the 

higher NUREG-1829 estimates as a result of the effects of PWSCC on small diameter piping failures. 

Key elements of this response have been incorporated into Section 7.10 of the revised NUREG. This 

comment is also very similar to GC5. Please see the response to this comment for additional information. 

Comment Number: 7-2 

Submitted by Zouhair Elawar – Palo Verde Nuclear Generating Station 

Comment: The draft NUREG combined a variety of LOCA sources into each LOCA category. Piping 

LOCAs and several non-piping LOCAs were pooled together to form each of the LOCA categories. It 

would be useful for each of the 6 LOCA categories to add a table of LOCA sources and frequency 

contributions. This breakdown is particularly important for the small and medium LOCA categories. 

Some contributors to the small and medium LOCAs are modeled separately in most PRA models (SGTR, 

RCP seals, inter-system LOCAs and others). If the end user does not subtract the separately-modeled 

LOCA contributors, then the contribution to CDF (core damage frequency) from those contributors would 

be conservatively and redundantly modeled. 

Response: First, it is important to recall that this study was only concerned with estimating passive 

system failure frequencies of structural system components (SSCs) within the primary system (Section 2). 

Therefore, LOCAs associated with RCP seals, interfacing system LOCAs, and active system LOCAs are 

not included in the NUREG-1829 estimates. There are previous estimates for these types of events (e.g., 

NUREG/CR-5750). The revised NUREG (Section 7.8) includes a table of SGTR frequencies as 

requested by this comment because all the panelists provided sufficient information to determine 

individual LOCA frequency estimates for this system. Section 7.8 of the revised NUREG also contains 

estimates for all non-SGTR passive-system SSC failures. 
M-21

In general, however, it was not possible to develop information for individual system failures from the 

elicitation responses. For each LOCA category there was a potential contribution from either 12 or 13

different piping systems (12 for the PWRs and 13 for the BWRs) plus a contribution from either 3 or 5 

non-piping components (3 for the BWRs and 5 for the PWRs). In addition, for each of the non-piping 

components, there were up to 8 (for the RPVs) subcomponents (e.g., vessel body, CDRMs, nozzle, head 

bolts, etc.) that were potential contributors. In order to make the responses tractable, the panelists were 

requested to only provide information for systems that they expected to make up at least 80 percent of the 

failure frequency contribution. Thus, not every panelist provided responses for every piping system or 

non-piping component or subcomponent. Furthermore, it was not possible to impute this data accurately

because the relative frequency contributions to each panelist’s estimates are small. Consequently, it was 

not possible to calculate group estimates of individual system failure frequencies because this would have 

required each panelist to provide estimates for every piping system and non-piping component.. 

Key elements of this response have been incorporated into Section 2 of the revised NUREG. Additionally, 

Section 7.8 has been added to the revised NUREG to provide the SGTR and non-SGTR estimates for PWR 

plants. 

Comment Number: 7-3 

Submitted by Zouhair Elawar – Palo Verde Nuclear Generating Station 

Comment: The steam generator tube rupture frequency (merged with LOCA category 1) was reported as 

mean value = 3.5E-03 based on number of industry events averaged over years of reactor operations. This 

mean value frequency should be separated from the main small LOCA category and estimated as a 

“range” consisting of upper and lower bounds. Plants with aging steam generators are encouraged to use

the upper bound of the range. And, plants with new steam generators may use the lower bound of the 

range. Of course, numerous plants would use the overall mean value of 3.5E-03. 

Response: Separate SGTR frequencies have been calculated from the elicitation responses and are 

provided in the revised NUREG (Section 7.8). The current-day median STGR frequency is 2.6E-3 and the 

mean frequency is 3.7E-3 (Table 7.18). These values compare favorably with the operating-experience 

Category 1 frequency of 3.5E-3 for SGTRs as reported in Section 4.4.1. The 5th and 95th percentiles of the 

group estimates are 5.0E-4 and 1.0E-2, respectively. Additional discussion concerning the SGTR 

estimates is also contained in Section 7.10 of the revised NUREG. 

Comment Number: 7-4 

Submitted by Zouhair Elawar – Palo Verde Nuclear Generating Station 

Comment: The various LOCA frequencies are reported in the several tables as cumulative values. In 

order to isolate the frequency of each LOCA category, one has to subtract the frequency of the next 

higher ranking category. This reporting format may lead to human errors. Some users may not become 

aware of the cumulative table format since that description is briefly stated at the later sections of a very 

large report. Please add a footnote under each LOCA frequency table explain how to obtain the frequency 

of each LOCA category. 

Response: The six LOCA categories were defined in terms of cumulative thresholds because the 

panelists felt more comfortable with providing their responses as cumulative thresholds rather than to

intervals defined by consecutive thresholds. (Appendix B) Because the differences between the results 

for consecutive LOCA categories are typically much smaller than their uncertainties (see Table 1, Figure 

1, Figure 7.36 and Figure 7.37), there is no statistical difference between using the cumulative frequencies 

or interval frequencies determined by subtraction. If interval-defined LOCA frequencies are required,

simply use the NUREG cumulative threshold estimates in Section 7.7. However, Section 7.9 does use 

interval-defined frequencies for consistency in comparing the NUREG-1829 estimates with other prior 
M-22

study estimates. This point has been clarified in the revised NUREG in the Executive Summary, Section

3.4.1, and Section 7.9. 

Comment Number: 7-5 

Submitted by Zouhair Elawar – Palo Verde Nuclear Generating Station 

Comment: The equivalent break diameters used in various PRA models that form boundaries between 

various LOCA categories are not necessarily matching those used in the draft NUREG. For example, a 

small LOCA range may extend up to 0.03 square feet equivalent break area (derived through existing 

capability of high pressure safety injection). The draft NUREG should give a clear guideline on 

interpolations between the various LOCA frequency values, including advice on arithmetic or geometric 

preference for interpolation. 

Response: This comment duplicates Comment 3-3. See the response to Comment 3-3 for guidance on 

interpolation schemes. 

Comment Number: 7-6 

Submitted by Zouhair Elawar – Palo Verde Nuclear Generating Station 

Comment: The various LOCA frequency tables provided results for the past 25-year operating time 

period and for the 40-year average plant life. The more suitable value to select for use in PRA models

may well be the expected LOCA frequencies applicable to the next 15 years. That particular selection 

allows for frequency penalties at aging plants as well as credits at plants with new steam generators 

and/or improved methods of inspections & leak detections. In order to extract the next 15-year LOCA 

frequencies from the existing tables, the user has to perform simple arithmetic calculations which may

lead to human errors and inconsistent applications across the industry. Please provide frequency estimates 

for the next 15-year time period. 

Response: As stated in the Executive Summary, the frequency estimates are not expected to change 

dramatically over the next fifteen years for any size LOCA, or even the next thirty-five years for LOCA 

Category 4 and smaller. Because of the predicted stability in these estimates over the near-term, it is 

recommended that the 25 year (i.e., current-day) results be used to estimate the average LOCA 

frequencies over the next 15 years of fleet operation. This last point was incorporated in both the 

Executive Summary and Section 7.4 of the revised NUREG. 

Comment Number: 7-7 

Submitted by Zouhair Elawar – Palo Verde Nuclear Generating Station 

Comment: There is insufficient description of small LOCA frequency’ “comparison results” relating to 

those in NUREG 5750 (which is most typically used in current PRA models). If one excludes the 

contribution of steam generator tube rupture frequency, the draft NUREG-1829 small LOCA value is 

ONE order of magnitude higher. Please add justifications for that large difference. 

Response: This comment is very similar to Comment GC3. See the response to Comment GC3 for the 

comparison of NUREG-1829 with both NUREG/CR-5750 and operating experience estimates. More 

detailed information on these comparisons is available in Sections 7.9 and 7.10.

Comment Number: 7-8 

Submitted by Zouhair Elawar – Palo Verde Nuclear Generating Station 

Comment: Provide a section on statistical validation of small LOCA frequency. By using the method of 

Jeffrey’s non-informative prior (over the past 1,500 reactor years with ZERO events excluding steam 

generator tube ruptures), the expected small LOCA frequency is at the 1E-04 level. This frequency is one 

order of magnitude lower than the frequency reported in the draft NUREG. Plant experiences of >1,500 

reactor-years of operating history should be considered as valid predictor of small LOCAs. That 
M-23

consideration is further strengthened by improved methods and increased requirements for in-serviceinspections & leak detections.

Response: A new section has been added to the NUREG (Section 7.10) to compare SB LOCA frequency 

estimates from the elicitation with operating experience. As discussed in this section, the elicitation 

estimates for BWR and PWR small break LOCA frequencies are generally consistent with operational 

experience estimates. The BWR and total PWR elicitation mean frequency estimates are only 20 percent 

and 100 percent higher, respectively, than operational experience estimates. Further the PWR SGTR 

frequencies are virtually identical to operating experience predictions. The biggest difference between the 

elicitation results and operating experience occurs for SB LOCA estimates that are determined without 

SGTR contributions. The elicitation mean frequency estimate is approximately 5 times higher than the 

operating experience estimate which accounts for nearly all of the 100% increase in the total PWR SB 

LOCA frequencies indicated above. Although the five-fold increase in the elicitation non-SGTR SB 

LOCA frequencies is not inconsistent with operating experience (Section 7.10), this difference is 

physically supported by the panelists’ qualitative and quantitative responses. The panelists indicated that 

medium and, to a lesser extent, small LOCAs in PWRs are most dramatically impacted by current 

PWSCC concerns (Section 6.3.2). This increase reflects this concern. 

More detail is found in Section 7.10 of the revised NUREG. Related insights are also provided in the 

responses to Comments GC3, GC4, GC5, GC6, and GC7. 

Comment Number: 7-9 

Submitted by Zouhair Elawar – Palo Verde Nuclear Generating Station 

Comment: Provide a section on probabilistic validation of the small LOCA frequency. Using a Poisson 

distribution with failure rate of 2.9E-03 (NUREG-1829 category 1 LOCA frequency excluding steam 

generator tube rupture events). And, approximately 1500 reactor-years, the probability of ZERO small 

LOCA events (actual industry performance) is ONLY 0.013 (i.e.1.3% chance). This result shows an 

excessive conservatism in the category 1 LOCA frequency. Are we to expect an average of ONE small 

LOCA event (with 74% chance) per 345 reactor-years (equivalent to ~4 calendar years)? I think not. 

Response: This comment is similar to both GC7 and 7-8. See the responses to these comments for 

comparisons between the NUREG-1829 Category 1 LOCA estimates and operating experience. Section 

7.10 has also been added to provide an in-depth comparison with operating experience. 

Comment Number: 7-10 

Submitted by Westinghouse Owners Group 

Comment: It would be useful to the PRA community, and help facilitate plant to plant consistency if, for 

the smaller break sizes, more information or guidance were provided to help separate out the frequencies 

of SGTR from small break LOCA, and CRDM nozzle breaks from medium LOCAs, as well as any other 

contributors other than primary system piping. Although there is no current intention to use the results of 

the expert elicitation to update the various LOCA frequencies assumed in individual plant PRAs, as plants 

go forward with peer reviews of PRAs, it is likely that LOCA frequencies for small, medium and large 

LOCAs will be compared with NUREG-1829 results. The NUREG-1829 frequencies for the smaller 

LOCA sizes are several times higher than the values presented in NUREG-5750. The NUREG-1829 

values listed for the small break frequency (> 100 gpm) are so high that one would expect to have seen 

an event in the US every three or four years, whereas to date we have seen none. It would be helpful if the 

conservatism in the estimates for the smaller beak sizes was discussed and some caveat provided so that 

plant PRAs don't end up with excessive conservatism in their small LOCA risk estimates. 

Response: This comment is similar to Comment 7-2. See this response to this comment for a discussion 

about separating failure of individual components like steam generator tubes and CRDMs from the total 
M-24

LOCA frequency estimates. Section 7.8 in the revised NUEREG provides separate SGTR estimates. 

Also, the responses to Comments GC3, GC4, GC5, GC7, 7-1, 7-3, 7-8, and 7-11 summarize comparisons 

between NUREG-1829 and either NUREG/CR-5750 or operating experience estimates. See also 

Sections 7.9 and 7.10 of the revised NUREG for more detailed comparisons. 

Comment Number: 7-11 

Submitted by Nuclear Energy Institute 

Comment: The draft NUREG combined a variety of LOCA sources into each LOCA category. Piping 

LOCAs and several non-piping LOCAs were pooled together to form each of the LOCA categories. It 

would be useful for each of the 6 LOCA categories to add a table of LOCA sources and frequency 

contributions. This breakdown is particularly important for the small and medium LOCA categories. 

Some contributors to the small and medium LOCAs are modeled separately in most PRA models (SGTR, 

RCP seals, inter-system LOCAs and others). If the end user does not subtract the separately-modeled 

LOCA contributors, then the contribution to CDF (core damage frequency) from those contributors would 

be conservatively and redundantly modeled. 

Response: This comment is similar to Comment 7-2. See the response to this comment for a discussion 

about separating failure of individual components like steam generator tubes and CRDMs from the total 

LOCA frequency estimates. Section 7.8 of the revised NUREG provides separate SGTR estimates. 

Comment Number: 7-12 

Submitted by Nuclear Energy Institute 

Comment: The various LOCA frequencies are reported in the several tables as cumulative values. In 

order to isolate the frequency of each LOCA category, one has to subtract the frequency of the next 

higher ranking category. This reporting format may lead to human errors. Some users may not become 

aware of the cumulative table format since that description is briefly stated at the later sections of a very 

large report. Please add a footnote under each LOCA frequency table explain how to obtain the frequency 

of each LOCA category. 

Response: This comment is identical to Comment 7-4. See the response to this comment for guidance 

on obtaining interval values from the cumulative results reported in NUREG-1829. This point has been 

clarified in the revised NUREG in the Executive Summary, Section 3.4.1, and Section 7.9. 

Comments Related to Section 8 of NUREG-1829

Comment Number: 8-1 

Submitted by Westinghouse Owners Group 

Comment: In Chapter 8, Ongoing Work, it is noted that the LOCA elicitation results were for normal 

operating conditions only. The effects of Service Level D transients, of which seismic was found by NRC 

to be the most prominent, were not considered in the elicitation efforts. The reason seismic loading was 

not explicitly considered was that most of the expert panel did not believe that it would significantly 

change the LOCA frequencies for normal operation. Experience from probabilistic fracture mechanics 

calculations indicates that severe seismic loading, such as that from a design-basis safe shutdown 

earthquake, could increase the conditional probability of failure in flawed piping by one to two orders of 

magnitude. However, the probability of having the severe seismic loading during the worst time in life, 

such as the 40th or 60th year of operation, would be a maximum of 0.001 and would likely be much less.

Thus, the maximum effect of this severe seismic loading would be to increase the LOCA frequency 

during normal operating conditions by 10 percent. This increase was deemed to be insignificant relative to 

the other uncertainties that were considered by the expert panel in the elicitation process for LOCA 

frequencies. 
M-25

Response: The results from a separately-sponsored NRC-led seismic LOCA study (Reference: Chokshi, 

N.C., Shaukat, S.K., Hiser A.L., DeGrassi, G., Wilkowski, G., Olson, R., and Johnson, J.J., "Seismic 

Considerations For the Transition Break Size," NUREG-1903, U.S. Nuclear Regulatory Commission, 

February 2008) tend to support this comment. In this study, both unflawed and flawed piping analyses 

were conducted in order to ascertain the magnitude of any potential adjustments to the baseline TBS for 

the proposed rule change to 50.46a due to failures associated with seismic loading.

The principal findings from this study are that the critical flaws associated with the stresses induced by 

seismic events of 10-5 and 10-6 annual probability of exceedance are large. When considering the effects 

of mitigation strategies to preclude large flaws in service, the probabilities of pipe breaks larger than the 

TBS are likely to be less than 10-5 per year. Similarly, for the cases studied, the probabilities of indirect 

failures of large RCS piping systems are less than 10-5 per year. 

These findings tend to support the contention of the commenter that seismic loading would not 

significantly change the LOCA frequencies under normal operation. As a result of this and related 

comments, the NRC report on Seismic Considerations for the Transition Break Size is now referenced in 

Section 2 of NUREG-1829. In addition, a summary of the seismic LOCA analysis and results is provided 

in the Executive Summary and in Section 7.2 of the report. Additionally, Section 2 clearly identifies that 

the elicitation LOCA frequency estimates do not consider rare event loading from seismic, severe water

hammer, and other similar sources. 

Comments Related to Section 9 of NUREG-1829

None 

Comments Related to Appendix A of NUREG-1829

None 

Comments Related to Appendix B of NUREG-1829

None 

Comments Related to Appendix C of NUREG-1829

None 

Comments Related to Appendix D of NUREG-1829

Comment Number: D-1 

Submitted by Joseph Conen of the BWR Owners Group 

Comment: Figure D.7 in Appendix D shows two through-wall IGSCC cases for 22 inch and 28 inch 

stainless steel pipe field history data. This reviewer is not aware of any through-wall 

IGSCC cracks in large diameter (>20-inch) BWR stainless steel pipes. A primary reason for this is the 

presence of mid-wall compressive weld residual stresses in such pipe that tend to retard deep cracks. 

Response: Figure D.7 in the draft NUREG only shows selected IGSCC data points (only weld flaws for 

which detailed sizing data are available). In actuality there are have been other leaks in 22-inch and 28-
M-26

inch diameter recirculation lines in BWRs. According to the expanded OPDE database used as the basis 

of this query (currently 1,215 records on IGSCC), there have been 10 instances of circumferential 

through-wall cracking in large diameter (D=22 inch to 28 inch recirculation system piping, 8 of which

were leaks. Three of the leaks were in 22-inch diameter piping: a cap-to-manifold leak at Hatch-1 (LER 

82-089, November 1982) and two welds at Monticello (LER 82-013, October 1982). The other five leaks 

were associated with the 28-inch diameter recirculation line at Brunswick-1 (LER 85-026, July 1985): 

Weld 1B32-RR-28-A-4, Weld 1B32-RR-28-A-14, Weld 1B32-RR-28-B-4, Weld 1B32-RR-28-B-8, and 

Weld 1B32-RR-28-A-15. 

Comments Related to Appendix E of NUREG-1829

None 

Comments Related to Appendix F of NUREG-1829

None 

Comments Related to Appendix G of NUREG-1829

None 

Comments Related to Appendix H of NUREG-1829

None 

Comments Related to Appendix I of NUREG-1829

None 

Comments Related to Appendix J of NUREG-1829

None 

Comments Related to Appendix K of NUREG-1829

None 

Comments Related to Appendix L of NUREG-1829

None 
BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse)

NRC FORM 335

(9-2004)

NRCMD 3.7

U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

 (Assigned by NRC, Add Vol., Supp., Rev.,

 and Addendum Numbers, if any.)

NUREG-1829, Vol. 2

3. DATE REPORT PUBLISHED

MONTH

April

YEAR

2008

4. FIN OR GRANT NUMBER

N6360

2. TITLE AND SUBTITLE

Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process

Appendices A through M

5. AUTHOR(S)

Robert Tregoning and Lee Abramson, US NRC

Paul Scott, Battelle

6. TYPE OF REPORT

Technical

7. PERIOD COVERED (Inclusive Dates)

N/A

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor,

Division of Engineering Battelle-Columbus

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Same as above"; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, 

provide name and mailing address.)

and mailing address.)

10. SUPPLEMENTARY NOTES

A. Csontos, NRC Project Manager

11. ABSTRACT (200 words or less)

The NRC is developing a risk-informed revision of the design-basis pipe break size requirements in 10 CFR 50.46, Appendix K

to Part 50, and GDC 35 which requires estimates of loss-of-coolant-accident (LOCA) frequencies as a function of break size. 

Separate BWR and PWR piping and non-piping passive system LOCA frequency estimates were developed as a function of 

effective break size and operating time through the end of license extension. The estimates were based on an expert elicitation

process which consolidated service history data and insights from probabilistic fracture mechanics studies with knowledge of 

plant design, operation, and material performance.

The elicitation required each member of an expert panel to qualitatively and quantitatively assess important LOCA contributing

factors and quantify their uncertainty. The quantitative responses were combined to develop BWR and PWR total LOCA

frequency estimates for each contributing panelist. The individual estimates were then aggregated to obtain group estimates,

along with measures of panel diversity. Sensitivity studies were conducted to examine the effects of distribution shape,

correlation structure, panelist overconfidence, measures of panel diversity, and aggregation method. The group estimates are

most sensitive to the method used to aggregate the individual estimates.

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.)

piping

risk-informed

emergency core cooling system (ECCS)

loss-of-coolant accident (LOCA)

break frequencies

design-basis break size

LOCA frequency estimates

expert elicitation

aging

14. SECURITY CLASSIFICATION

13. AVAILABILITY STATEMENT

unlimited

(This Page)

unclassified

(This Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (9-2004) PRINTED ON RECYCLED PAPER

Office of Regulatory Research 505 King Avenue

U.S. Nuclear Regulatory Commission Columbus, OH 43201

Washington, DC 20555-0001

Same as above