NUREG/CR-6936 

PNNL-16186 

Probabilities of Failure and 

Uncertainty Estimate 

Information for Passive 

Components – A Literature 

Review 

Pacific Northwest National Laboratory 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Regulatory Research 

Washington, DC 20555-0001 
 NUREG/CR-6936 

 PNNL-16186 

Probabilities of Failure and 

Uncertainty Estimate 

Information for Passive 

Components – A Literature 

Review 

Manuscript Completed: March 2007 

Date Published: May 2007 

Prepared by 

S.R. Gosselin (1), F.A. Simonen (1), S.P. Pilli (1) 

B.O.Y. Lydell (2) 

(1) Pacific Northwest National Laboratory 

P.O. Box 999 

Richland, WA 99352 

(2) Sigma-Phase Inc. 

16917 Orchid Flower Trail 

Vail, AZ 85641 

S.N. Malik, NRC Project Manager 

Prepared for 

Division of Fuel, Engineering and Radiological Research 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

NRC Job Code N6019 
iii 

Abstract 

This report describes a study performed for the U.S. Nuclear Regulatory Commission by Pacific 

Northwest National Laboratory with subcontractor support from Sigma-Phase, Inc. The effort involved 

reviewing the open literature to collect estimates of failure probability and their associated uncertainties 

for passive reactor components. The review focused primarily on probabilistic structural mechanics 

evaluations of plant-specific components for domestic nuclear power plants with pressurized water 

reactors or boiling water reactors as well as for international plants of similar design. 

A computerized search of several databases identified more than 7500 documents, of which only a small 

fraction were related to probabilistic structural mechanics calculations. Probabilistic treatments and 

characterizations of fatigue and stress corrosion cracking were found to be well represented in the 

literature, but only a limited number of the publications described plant-specific and component-specific 

probabilistic evaluations based on actual design or operating stresses. The NRC will apply these results 

during the development of probabilistic fracture mechanics tools to generate failure probabilities for 

passive reactor components for use in regulatory decision making. 
v 

Foreword 

The present approach to effective materials degradation management in nuclear power plants involves 

selecting appropriate materials for the design of components and monitoring degradation during 

operations. The Code of Federal Regulations identifies the regulatory requirements for both component 

design and periodic inservice inspections to ensure that design safety margins are maintained throughout 

component life. Plant technical specifications also include requirements for leakage monitoring and 

reactor shutdown to provide defense in depth to ensure the integrity of the reactor coolant system 

boundary. Lastly, the U.S. Nuclear Regulatory Commission (NRC) issues generic letters, bulletins, and 

orders to address emergent issues. 

Notwithstanding this multifaceted regulatory framework, instances of unexpected materials degradation 

in nuclear power plants during recent years have led to a heightened interest by the nuclear power 

industry and the NRC in developing a proactive approach to materials degradation management. The 

establishment of a proactive program requires the identification of the components and materials that are 

expected to experience future degradation and the associated degradation mechanisms. This report 

presents the results of the NRC’s review of the open literature to collect estimates of failure probabilities 

and their associated uncertainties for passive reactor components. The review focused primarily on 

probabilistic structural mechanics evaluations of plant-specific components for domestic nuclear power 

plants with pressurized-water reactors or boiling-water reactors, as well as for international plants of 

similar design. 

The review showed that probabilistic treatments and characterizations of fatigue and stress-corrosion 

cracking are well represented in the literature, but only a limited number of the publications describe 

plant-specific and component-specific probabilistic evaluations based on actual design or operating 

stresses. The NRC will apply these results during the development of probabilistic fracture mechanics 

tools to generate failure probabilities for passive reactor components for use in regulatory decision 

making. 

 Brian W. Sheron, Director 

 Office of Nuclear Regulatory Research 

 U.S. Nuclear Regulatory Commission 
vii 

Contents 

Abstract.............................................................................................................................................. iii 

Foreword............................................................................................................................................ v 

Executive Summary ........................................................................................................................... xiii 

Acknowledgments.............................................................................................................................. xv 

Abbreviations and Acronyms ............................................................................................................ xvii 

1 Introduction................................................................................................................................ 1.1 

2 Summary of Past Failure Probability Studies ............................................................................ 2.1 

2.1 Rates of Initiating Events................................................................................................... 2.1 

2.2 LOCA Frequency Elicitation Process................................................................................ 2.2 

2.3 Risk-Informed Inservice Inspection Evaluations............................................................... 2.4 

2.4 Other Studies ..................................................................................................................... 2.4 

3 Review of Existing Probabilistic Fracture Mechanics Models for Reactor Components.......... 3.1 

3.1 Reactor Pressure Vessel Codes.......................................................................................... 3.1

3.2 Reactor Piping Codes ........................................................................................................ 3.1 

3.2.1 PRAISE.................................................................................................................. 3.1 

3.2.2 PRO-LOCA............................................................................................................ 3.2 

3.2.3 Other Piping Codes ................................................................................................ 3.3 

4 Open Literature Search .............................................................................................................. 4.1 

4.1 Methodology...................................................................................................................... 4.1 

4.1.1 Search Strategy for Compendex, Corrosion Abstracts, Metadex, and 

 NTIS Search ........................................................................................................... 4.2 

4.1.2 Search Strategy for ADAMS.................................................................................. 4.3 

4.1.3 Search Strategy for Science Research Connection................................................. 4.3 

4.2 Literature Search Results................................................................................................... 4.4 

5 Failure Probabilities for Reactor Components........................................................................... 5.1 

5.1 Stress Corrosion Cracking ................................................................................................. 5.1 

5.1.1 Intergranular Stress Corrosion Cracking................................................................ 5.1 

5.1.2 Transgranular Stress Corrosion Cracking .............................................................. 5.9 

5.1.3 Primary Water Stress Corrosion Cracking ............................................................. 5.11 

5.2 Fatigue ............................................................................................................................... 5.14 

5.2.1 Thermal Fatigue ..................................................................................................... 5.14

5.2.2 Vibration Fatigue ................................................................................................... 5.22
viii 

5.3 Flow-Assisted Degradation ............................................................................................... 5.24 

5.3.1 Flow-Accelerated Corrosion .................................................................................. 5.24 

5.3.2 Erosion-Corrosion .................................................................................................. 5.26 

5.3.3 Cavitation-Erosion ................................................................................................. 5.28 

5.4 Local Corrosion Mechanisms............................................................................................ 5.28

6 Uncertainty in Estimated Failure Probabilities .......................................................................... 6.1 

6.1 Uncertainties in Fatigue Calculations with PRAISE ......................................................... 6.1 

6.2 Uncertainties in LOCA Frequencies from Expert Elicitation............................................ 6.5 

7 Conclusions................................................................................................................................ 7.1 

8 References.................................................................................................................................. 8.1 
ix 

Figures 

5.1 Field Observations of Leak Probabilities and PRAISE Predictions at Various 

Residual Stress Adjustment Factor Values for 304 Stainless Steel Pipes Having 

Outside Diameters Greater Than 508 mm............................................................................... 5.4 

5.2 Sample Comparison of PRAISE Cumulative Leak Probabilities and Results 

Obtained by Khaleel et al. (1995) for Various Values of Residual Stress Adjustment 

Factors and Plant Loading/Unloading Frequencies................................................................. 5.4 

5.3 Mark I BWR Reactor Recirculation System Piping................................................................ 5.5 

5.4 NUREG/CR-4792 Relative Contributions of Various Weld Types to BWR 

Recirculation System Probabilities of (a) Leak and (b) DEGB............................................... 5.6 

5.5 NUREG/CR-4792 Cumulative Leak Probabilities for BWR Recirculation System 

Weld Types in the “Old” Configuration with TP304 Stainless Steel Piping........................... 5.6 

5.6 NUREG/CR-4792 Cumulative DEGB Probabilities Between the Old Configuration 

with TP304 Stainless Steel Piping and the New Configuration with TP316 

Stainless Steel Piping .............................................................................................................. 5.7 

5.7 Effect of Lowering Oxygen Concentration at Midlife in 102-mm Schedule 80 

304 Stainless Steel BWR Pipe................................................................................................. . 5.8 

5.8 Effect of Reducing Residual Stress at Midlife in 102-mm Schedule 80 

Type 304 Stainless Steel BWR Pipe ....................................................................................... 5.8 

5.9 Cumulative Probability of a Leak Exceeding 378.5 lpm Versus Time for a 304-mm 

Recirculation Line Weld With and Without a Weld Overlay at 20 Years ............................. 5.10 

5.10 Comparisons of SRDM Estimates with All Available Data on Alloy 600 Initiation 

Life for Various Tests and Conditions in High-Purity Water Environments ......................... 5.12 

5.11 PWSCC Crack Growth Rate versus Crack Stress Intensity for Alloy 600 

Laboratory Data, EPRI Material Reliability Program Crack Growth Curve, 

Modified Scott Curve, and CGR Data Points for Cook 2 Nozzle 75 ..................................... 5.13 

5.12 Comparisons of MRP-115 Curves for Alloys 182/132 and 82 with MRP 55 

Westinghouse CGR Data for Weld Material Removed from VC Summer Reactor 

Hot Leg Nozzle Dissimilar Metal Weld ................................................................................. 5.13 

5.13 Zion 1 Primary System Piping Weld Locations and NUREG/CR-2189 Estimated 

Leak Frequencies..................................................................................................................... 5.17 
x 

5.14 NUREG/CR-2189 Conditional LOCA Probabilities as a Function of Time for 

Two Zion I Representative Weld Locations Showing the Influence of Seismic Events ........ 5.18 

5.15 Socket Weld Failure as Percentage of All Failures of Safety Related Piping of 

Nominal Pipe Size ≤102-mm ................................................................................................. 5.23 

6.1 Histogram for Probability of Leak for Q = 100....................................................................... 6.4 

6.2 Comparisons of Probabilities from Uncertainty Analyses with Best-Estimate 

Calculations ............................................................................................................................. 6.4 

6.3 Maximum Failure (Leak and Break) Probabilities from Sample of 100 Cases Evaluated 

by Uncertainty Analyses ......................................................................................................... 6.5 

6.4 Total BWR LOCA Frequencies as a Function of the Threshold Break Diameter 

at 25 Years of Plant Operations............................................................................................... 6.6 

6.5 Total PWR LOCA Frequencies as a Function of the Threshold Break Diameter 

at 25 Years of Plant Operations............................................................................................... 6.6 

Tables 

2.1 Estimated LOCA Frequencies for PWR and BWR Plants ...................................................... 2.2 

2.2 Baseline LOCA frequency Estimates from Expert Elicitation ............................................... 2.4 

5.1 Summary of IGSCC Service Experience in BWR Nuclear Power Plants............................... 5.2 

5.2 Residual Stress Adjustment Factors, f .................................................................................... 5.3 

5.3 Summary of Leak and DEGB Probabilities as a Function of Time for 

Three Pipe Size Ranges .......................................................................................................... 5.3 

5.4 NUREG/CR-4792 Mark I BWR Recirculation System Configurations ................................. 5.6 

5.5 Cumulative Probability PRAISE Results for the 304-mm and 711-mm Recirculation 

Line Welds .............................................................................................................................. 5.10 

5.6 Summary of Service Experience Involving TGSCC............................................................... 5.11 

5.7 Summary of Service Experience in Piping Involving PWSCC............................................... 5.12 
xi 

5.8 Cumulative PRAISE Failure Probabilities for Hot Leg Pressure Vessel-to-Safe End 

Alloy 600 Weld ...................................................................................................................... 5.15 

5.9 Summary of Service Experience Involving Thermal Fatigue ................................................ 5.15 

5.10 Zion 1 Primary System Piping Design Transients and Number of Postulated 

Occurrences in 40-year Design Life........................................................................................ 5.16

5.11 Summary of Results for All Seven Plants - Water Environment ............................................ 5.19 

5.12 Summary of Results for All Seven Plants - Air Environment ................................................ 5.20 

5.13 NUREG-1829 Appendix F Cumulative Probabilities for PWR Surge Line Elbow, 

PWR HPI Nozzle Safe End and BWR Feedwater Line .......................................................... 5.22 

5.14 Service Experience Involving Butt Weld and Socket Weld Failure Due to 

Vibration Fatigue..................................................................................................................... 5.24 

5.15 Summary of Service Experience Involving Flow-Accelerated Corrosion .............................. 5.25 

5.16 Surry Unit 1 305-mm Carbon Steel Feedwater System Piping Segment Small 

Leak Failure Probability by the Westinghouse SRRA Wastage Model Output ..................... 5.27 

5.17 Summary of Service Experience Involving Erosion-Corrosion of 

Service Water Piping............................................................................................................... 5.27 

5.18 Summary of Service Experience Involving Cavitation-Erosion.............................................. 5.28 

5.19 Summary of Service Experience Involving MIC .................................................................... 5.29 
xiii 

Executive Summary 

Failure frequencies for passive pressure boundary components often have been estimated to support 

probabilistic risk assessments. These estimates have been based primarily on reported failures from past 

plant operation and have established failure frequencies on a system rather than a component level. As 

such, these estimates have limitations in terms of predicting future performance and in identifying 

priorities for managing the integrity of specific components. Being based on reported operational events, 

the estimates are relatively good for failure modes consisting of small leaks, which have little potential to 

contribute to core damage. Corresponding estimates for the frequencies of larger leaks and pipe ruptures 

are, however, subject to much larger uncertainties. 

This report describes a study performed for the U.S. Nuclear Regulatory Commission by Pacific 

Northwest National Laboratory with subcontractor support from Sigma-Phase Inc. This effort collected 

estimates of failure probability and their associated uncertainties from the open literature for passive 

reactor components. The review focused primarily on probabilistic structural mechanics evaluations of 

plant-specific components for domestic plants with pressurized water reactors and boiling water reactors 

as well as for international plants of similar design. 

Computerized searches of several databases identified more than 7500 documents, of which only a small 

fraction were related to probabilistic structural mechanics calculations. Approximately 60% of the 

documents were related to corrosion mechanisms (e.g., flow-accelerated corrosion, erosion-corrosion, 

cavitation erosion, boric acid corrosion, crevice corrosion, pitting). The remaining documents were 

divided approximately equally between fatigue and stress corrosion cracking mechanisms. The vast 

majority of the references addressed various mechanistic aspects of degradation mechanisms and applied 

deterministic rather than probabilistic models. The literature search did not identify many plant-specific 

studies completed by industry-sponsored owners’ groups or by research organizations since they are 

reported in documents not publicly available in the open literature. 

Although the probabilistic treatments and characterizations of fatigue and stress corrosion cracking are 

well documented in the literature, only a limited number of these publications document plant-specific 

and component-specific evaluations based on actual design or operating stresses. Examples from the 

more relevant of these studies are discussed in this report. 

The literature search did not identify any applications of probabilistic structural mechanics models for 

local corrosion such as pitting and microbiologically influenced corrosion. However, estimates of failure 

probabilities have been reported in plant-specific risk-informed inservice inspection (RI-ISI) evaluations. 

In some applications, the failure frequency estimates were based on statistical evaluations of service 

failure data. Failure probability calculations have been based also on assumed estimates of corrosion 

rates and on qualitative evaluations made by plant RI-ISI expert panels. 

The open literature search did not identify any published applications of probabilistic structural 

mechanics models in nuclear power plant systems subject to the various wall-thinning types of 

degradation mechanisms, such as flow-assisted corrosion. As with other corrosion mechanisms, failure 

probabilities for flow-assisted corrosion have been estimated in support of plant-specific RI-ISI 

evaluations. In these cases, failure frequency estimates have been based on estimated inputs for rates of 
xiv 

material wastage and wall thinning rates. In these calculations, corrosion-specific mechanistic models 

were not applied, and wall thinning was assumed to have the same impact on structural integrity as a 

circumferential crack that penetrated the wall to the same depth. 

Formal attempts to quantify uncertainties in estimated failure probabilities show that the estimated failure 

probabilities for specific components are subject to large uncertainties, whether the estimate is based on 

data from operating experience or based on an application of probabilistic structural mechanics models. 

The reported uncertainties are particularly large when the estimated probabilities are very small. Initial 

flaw size distributions were identified as an especially large source of uncertainty in calculated failure 

probabilities because of the unavoidable difficulty in estimating the very low probabilities for the large 

fabrication flaws, which (if present) have a major impact on piping integrity. Consequently there can be 

significant uncertainties in calculated probabilities that are related to both the probabilistic fracture 

mechanics models as implemented in the various computer codes and the judgments made in defining 

critical input parameters for plant-specific calculations. 
xv 

Acknowledgments 

The authors wish to thank the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory 

Research for supporting this work and, in particular, Dr. J. Muscara and NRC project manager, 

Dr. S. N. Malik. 
xvii 

Abbreviations and Acronyms 

ASME American Society of Mechanical Engineers 

B&W Babcock & Wilcox 

BWR boiling water reactor 

CC component cooling 

CGR crack growth rate 

Compendex Computerized Engineering Index 

CVC chemical and volume control 

DEGB double-ended guillotine break 

DET detection 

DHR decay heat removal 

EMC2 Engineering Mechanics Corporation of Columbus 

EPRI Electric Power Research Institute 

FAC flow-accelerated corrosion 

GE General Electric 

GPM gallons per minute 

GRA&I Government Report Announcements & Index 

IGSCC intergranular stress corrosion cracking 

INEEL Idaho National Engineering and Environmental Laboratory 

LAS low alloy steel 

LER Licensee Event Report 

LOCA loss-of-coolant accident 

L/min liters per minute

METADEX Metals Abstracts/Alloy Index 

MIC microbiologically influenced corrosion 

MRP materials reliability program 

MSIP mechanical stress improvement process 

NPP nuclear power plant 

NPS nominal pipe size 

NRC U.S. Nuclear Regulatory Commission 

NTIS National Technical Information Service 
xviii 

PFM probabilistic fracture mechanics 

PIRT Phenomena Identification and Ranking Technique 

PMDA Proactive Materials Degradation Assessment 

PNNL Pacific Northwest National Laboratory 

POD probability of detection 

PPM parts per million 

PRA probabilistic risk assessment 

PTS pressurized thermal shock 

PTWC probability of through-wall crack 

PWR pressurized water reactor 

PWSCC primary water stress corrosion cracking 

RI-ISI risk-informed inservice inspection 

RVH reactor vessel head 

SCC stress corrosion cracking 

SKI Swedish Nuclear Inspectorate 

SRDM simplified strain rate damage model 

SRRA structural reliability and risk assessment 

SW service water 

TGSCC transgranular stress corrosion cracking 

TWC through-wall crack 

UHS ultimate heat sink 

USNRC U.S. Nuclear Regulatory Commission 

UT ultrasonic testing 

W Westinghouse 
1.1 

1 Introduction 

The U.S. Nuclear Regulatory Commission (NRC) has supported the research program Proactive Materials 

Degradation Assessment (PMDA) (USNRC 2007). The objective of this program has been to assess the 

possible occurrence of materials degradation and component degradation of light water reactors from the 

view of long-term operation such as would be relevant to relicensed plants. A central intent of the PMDA 

program has been to predict future occurrences of degradation that may or may not have been observed in 

the field or in the laboratory. Another objective is to consider the possibility of unexpected increases in 

degradation rates with time. 

The PMDA program has included an assessment, conducted under contract with Brookhaven National 

Laboratory, of past and possible future materials degradation in light water reactors. The study used the 

Phenomena Identification and Ranking Technique (PIRT) involving eight experts from five countries who 

first met to discuss the technical issues and then to perform individual assessments. The analyses focused 

on materials degradation modes associated with operating environments for specific components, such as 

stress corrosion cracking, fatigue, flow-accelerated corrosion, boric acid corrosion, thermal aging 

embrittlement and radiation effects for existing plants. The work encompassed passive components for 

which failure of the pressure boundary could lead to a release of radioactivity to the environment or 

would degrade the ability of safety systems to perform their intended functions. The study did not 

address design issues such as mechanics and thermal hydraulics, the consequences of degradation, or the 

failure of active components such as values and pumps. 

A detailed review by the panel of experts addressed more than 2000 components in the primary, 

secondary, and tertiary systems of specific reactor designs. Each expert individually provided judgments 

to score individual components in terms of “degradation susceptibility” and the “extent of knowledge” 

needed to predict the potential for future failures. These inputs were compiled and were used to generate 

a summary that reflected the collective judgment of the experts as a whole. 

This report describes a study performed for the NRC by Pacific Northwest National Laboratory (PNNL), 

with subcontractor support from Sigma-Phase, Inc., to collect probability of failure estimates and their 

associated uncertainties as reported from previous studies in the open literature for passive reactor 

components. Although the literature contains examples of failure probability and frequency(a) estimates 

from statistical treatment of service failure data, this review focused primarily on examples of 

probabilistic structural mechanics evaluations of plant-specific components for domestic pressurized 

water reactor (PWR) and boiling water reactor (BWR) plants as well as international plants of similar 

design. 

Section 2 describes some previous studies that have generated estimates of failure frequencies for reactor 

components in general and piping systems in particular. 

 

(a) In this report, component failure, failure probability, and failure frequency generally refer to failure 

of the component pressure boundary associated with through-thickness cracks and small/large leaks. 

In some references, failure has also been used when referring to the presence of part throughthickness cracks.
1.2 

Section 3 provides background on past and current efforts to develop probabilistic fracture mechanics 

codes, both in the United States and overseas. These codes have addressed the structural integrity of both 

reactor pressure vessels and reactor piping systems. This section describes some of these codes with an 

emphasis on codes that have been developed and applied in NRC research programs. 

Section 4 describes the strategy, methodology, and results of a literature search for failure probability 

information that covered the reactor components and degradation mechanisms within the scope of the 

PMDA program. Computerized searches of several databases identified more than 7500 documents, of 

which only a small fraction was related to probabilistic fracture mechanics calculations. 

Section 5 includes summaries of pipe failure probability estimates obtained by applying probabilistic 

structural mechanics models for different degradation mechanisms and for different reactor components. 

The information was obtained from reviews of public domain literature and is organized by the 

degradation mechanism types, including stress corrosion cracking (e.g., intergranular stress corrosion 

cracking [IGSCC], transgranular stress corrosion cracking [TGSCC], and primary water stress corrosion 

cracking [PWSCC]), thermal and vibration fatigue, flow-assisted degradation (e.g., flow-accelerated 

corrosion [FAC], erosion-corrosion, and cavitation erosion), and local corrosion (e.g., crevice corrosion, 

pitting, and microbiologically influenced corrosion [MIC]). 

Section 6 presents published results of formal attempts to quantify the levels of uncertainty in estimated 

failure probabilities. The uncertainties associated with the individual inputs and modeling assumptions 

used in probabilistic fracture mechanics (PFM) calculations for fatigue failures are presented along with 

the effects of these uncertainties on the calculated failure probabilities. Uncertainties assigned by 

participants in an expert elicitation process used to estimate loss-of-coolant accident (LOCA) frequencies 

also are presented. 

Section 7 summarizes the results and conclusions from this literature review. 
2.1 

2 Summary of Past Failure Probability Studies 

The NRC research on the PMDA program (USNRC 2007) has the objective of assessing possible future 

degradation and failures of components from the viewpoint of long-term operation. One objective has 

been to predict future occurrences of new types of degradation that would not be expected based on the 

extrapolations of prior field experience. Another objective is to anticipate unexpected and significant 

increases in the failure rates for previously experienced degradation mechanisms. 

Failure frequencies for passive pressure boundary components have often been estimated as described in 

NUREG/CR-5750 (Poloski et al. 1999) for use in probabilistic risk assessments (PRAs). Such estimates 

have been based on reported failures (e.g., leaks, ruptures) from past plant operation and have established 

failure frequencies on a system rather than component level. As such, these estimates have limitations in 

terms of predicting future performance and in identifying priorities for managing the integrity of specific 

components. Being based on reported operational events, the estimates will be relatively good for failure 

modes consisting of small leaks that have little potential to contribute to core damage. Corresponding 

estimates for the frequencies of larger leaks and pipe ruptures are, therefore, subject to large uncertainties. 

The development of risk-informed inservice inspection (RI-ISI) programs also has required estimates of 

failure frequencies. In this work, the focus has been on the component level to a greater extent than other 

evaluations intended to support PRAs. However, the postulated degradation mechanisms have largely 

been limited to the mechanisms known to be active based on past experience. As such, other mechanisms 

that may have long incubation times have not been considered, and time-related increases in failure 

frequencies for known mechanisms have not been explicitly addressed. New mechanisms and increased 

failure rates for known mechanisms are addressed by periodic updating of the risk-informed inservice 

inspection programs. 

This section describes examples of some previous studies that have generated estimates of failure 

frequencies for reactor components in general and piping systems in particular. 

2.1 Rates of Initiating Events 

A study (Poloski et al. 1999) by Idaho National Engineering and Environmental Laboratory (INEEL) 

provided estimates of initiating events at U.S. nuclear power plants, based on operating experience as well 

as other engineering evaluations. Loss-of-coolant accidents including pipe breaks were a major 

consideration in the study. The objectives were to 1) provide revised, historical frequencies for the 

initiating events, 2) compare these estimates with prior estimates used in PRAs, and 3) review the plant 

data for trends related to specific plant types (PWR plants versus BWR plants). The data to support the 

INEEL work came largely from the NRC Licensee Event Reports (LERs). This review gave 

approximately 2000 events to support the evaluations. Because the events of interest were limited to 

those that resulted in reactor trips, only the more significant modes of structural failures were covered. 

As such, occurrences of small leaks and observations of material degradation just simply requiring the 

repairs would have been largely excluded from consideration. In the case of rare events such as LOCAs, 

the INEEL researchers supplemented their LER-based evaluations with trends from available evaluations 

of engineering aspects of LOCA events. In these cases, analysis of available operating data, fracture 

mechanics analyses, and expert elicitation processes were employed to support reasonable but 
2.2 

conservative adjustments to previous best-estimate LOCA frequencies in WASH-1400 (USNRC 1990b) 

and NUREG-1150 (USNRC 1975). 

The estimated frequencies for LOCA events were less than the frequencies for other important initiating 

events such as loss of offsite power and loss of feedwater flow. The frequency for the small pipe break 

LOCA was estimated to be 5 × 10-4 (per plant per year) for both PWR and BWR plants. For PWR plants, 

the estimated frequency for steam generator tube rupture was about a factor of 10 greater, at 7 × 10-3. 

Table 2.1 lists the frequencies for small, medium, and large LOCAs. For medium and large breaks, the 

frequencies were based in part on the operating data for small breaks and in part on conservative 

estimates from available fracture mechanics evaluations for the ratios between frequencies for the 

different LOCA categories. The frequencies reported in Table 2.1 (Poloski et al. 1999) were based on 

generic definitions of small, medium and large LOCAs and break sizes assumed in NUREG-1150 

(USNRC 1990b). 

Table 2.1 Estimated LOCA Frequencies (per plant per year) for PWR and BWR Plants 

(Poloski et al. 1999) 

Plant Type Small LOCA Medium LOCA Large LOCA 

PWR 5 × 10-4 yr-1 4 × 10-5 yr-1 5 × 10-6 yr-1

BWR 5 × 10-4 yr-1 4 × 10-5 yr-1 5 × 10-6 yr-1

The LOCA frequencies in Table 2.1 are of limited usefulness in meeting the needs of the PMDA program 

because they represent the total contribution for all of the piping of the primary coolant system ranging 

from small-bore to large-diameter piping. For PRA evaluations, there is no need to identify specific 

components, materials, and environments that contribute to the overall LOCA frequency. A simplified 

approach could apportion the failure frequency equally among all the welds in the system, but this 

approach would fail to identify the small fraction of welds that would be expected to make the dominant 

contributions to the system-level failure frequency. 

2.2 LOCA Frequency Elicitation Process 

A recent NRC study applied the expert elicitation process to consolidate service history data and insights 

from PFM studies with knowledge of plant design, operation, and material performance and to thereby 

estimate failure frequencies for pressure boundary components (Tregoning et al. 2005). The objective 

was similar to that of the INEEL study (Poloski et al. 1999) in that the final results were estimates of 

system-level LOCA frequencies. However, the study required that the experts identify the components 

and degradation mechanisms that were the dominant contributors to the various categories of system-level 

LOCA frequencies. As such, the results of the elicitation were more closely aligned with the needs of the 

PMDA program. 

Separate piping and non-piping passive system LOCA frequency estimates were developed as a function 

of effective break size and operating time for both BWR and PWR plants. The elicitation focused solely 
2.3 

on determining event frequencies that initiate by unisolable primary system side failures that can be 

exacerbated by material degradation with age. The expert elicitation process employed was an adaptation 

of the formal expert judgment process used in NUREG-1150 (USNRC 1990b). 

Most panelists agreed that a complete break of a smaller pipe is more likely than an equivalent size of 

opening in a larger pipe. Frequency estimates were not expected to change dramatically over the next 

15 years or even the next 35 years. Although aging will continue, the consensus was that mitigation 

procedures are in place or will be implemented in a timely manner to alleviate possible LOCA frequency 

increases. 

The quantitative responses were analyzed separately for each panel member to develop individual BWR 

and PWR total LOCA frequency estimates of the mean, median, 5th, and 95th percentiles. The LOCA 

frequencies for the individual panelists were then aggregated to obtain group LOCA frequency estimates, 

along with measures of panel diversity. While there was general qualitative agreement among the 

panelists about important technical issues and LOCA contributing factors, the individual quantitative 

estimates were much more variable. Additionally, as the LOCA size increased, the panel members 

generally expressed greater uncertainty in their predictions, and the variability among individual 

panelists’ estimates increased. The elicitation LOCA frequency estimates were generally much less than 

the previous WASH-1400 (USNRC 1975) estimates and more consistent with the NUREG/CR-5750 

estimates (Poloski et al. 1999). The elicitation medium-break LOCA estimates were higher than the 

NUREG/CR-5750 estimates by factors of 2.5 and 10 for BWR and PWR plant types, respectively. The 

baseline LOCA frequency estimates for the current day and end of license extension period are provided 

in Table 2.2 for both BWR and PWR plants. The aggregated group estimates for the median, mean, and 

5th and 95th percentiles are summarized. 

Selected PFM analyses were performed as part of the elicitation process to assist the participants in 

estimating failure frequencies. Results from some of these calculations are described in Section 5. The 

experts were also provided for their consideration a collection of data and evaluations coming from the 

PIPExp database (Lydell et al. 2004). The experts tended to use trends from the data on the reported 

operating events as the primary consideration in estimating failure frequencies and, in particular, for the 

smaller LOCA sizes. However, trends from fracture mechanics calculations were used for insights into 

the relative frequencies for the larger LOCA sizes compared to the frequencies for the small LOCA sizes 

(e.g., through-wall crack with perceptible leak). 

The pipe failure database PIPExp was used in the LOCA frequency study. This database in an extension 

of the OECD Piping Failure Data Exchange (OPDE) database (Nuclear Energy Agency 2006) and makes 

use of the Microsoft Access database software. Over the time period of its development beginning in 

1998, the PIPExp database has been expanded in terms of both the absolute number of event records and 

the depth of the database structure. The PIPExp database includes applicable worldwide BWR- and 

PWR-specific service experience with Code Class 1 piping. As of December 31, 2002, the database 

accounted for approximately 1992 and 3621 critical reactor-years of operating experience with 

commercial BWR and PWR plants, respectively. The database is actively maintained and periodically 

updated. The effort has involved populating the database while at the same time ensuring data quality. 

For an event to be considered for inclusion, it undergoes screening for eligibility. 
2.4 

Table 2.2 Baseline LOCA frequency Estimates from Expert Elicitation 

(Tregoning et al. 2005) 

Current Day Estimate (per cal. year) Next 15 Year Estimate (per cal. yr.) 

Plant (25-yr fleet average operation) (End of original license) 

Type 

LOCA 

Size 

(gpm) 

Break 

Size 

(in.) 5th Median Mean 95th 5th Median Mean 95th 

>100 1/2 3.1 × 10-5 2.3 × 10-4 5 × 10-4 1.6 × 10-3 2.5 × 10-5 2 × 10-4 4.5 × 10-4 1.4 × 10-3

>1,500 1-7/8 3.1 × 10-6 4.1 × 10-5 9.9 × 10-5 3.3 × 10-4 2.5 × 10-6 3.5 × 10-5 9 × 10-5 3 × 10-4

>5,000 3-1/4 4.7 × 10-7 6.8 × 10-6 1.9 × 10-5 6.7 × 10-5 4.1 × 10-7 6.8 × 10-6 2.1 × 10-5 7.3 × 10-5

>25K 7 7.3 × 10-8 1.5 × 10-6 4.9 × 10-6 1.7 × 10-5 6.4 × 10-8 1.6 × 10-6 6.1 × 10-6 2.1 × 10-5

>100K 18 7.4 × 10-9 1.3 × 10-7 9.8 × 10-7 2.6 × 10-6 7.7 × 10-9 1.4 × 10-7 1.3 × 10-6 3.1 × 10-6

BWR 

>500K 41 1.1 × 10-11 3.2 × 10-10 4.9 × 10-9 9.7 × 10-9 1.3 × 10-11 4.3 × 10-10 7.2 × 10-9 1.4 × 10-8

>100 1/2 5.6 × 10-4 3.1 × 10-3 6 × 10-3 1.8 × 10-2 2.7 × 10-4 1.9 × 10-3 4.4 × 10-3 1.3 × 10-2

>1,500 1-5/8 9.2 × 10-6 1.3 × 10-4 4.9 × 10-4 1.5 × 10-3 9.4 × 10-6 1.5 × 10-4 6.1 × 10-4 1.9 × 10-3

>5,000 3 1.8 × 10-7 3.6 × 10-6 1.5 × 10-5 5.1 × 10-5 3.5 × 10-7 8 × 10-6 3.4 × 10-5 1.1 × 10-4

>25K 7 1.1 × 10-8 2.6 × 10-7 1.5 × 10-6 4.5 × 10-6 2.6 × 10-8 5.4 × 10-7 3.4 × 10-6 1 × 10-5

>100K 14 3 × 10-100 1.2 × 10-8 1.5 × 10-7 3.1 × 10-7 1 × 10-9 2.9 × 10-8 3.8 × 10-7 7.9 × 10-7

PWR 

>500K 31 1.4 × 10-11 9.9 × 10-10 2.5 × 10-8 4.2 × 10-8 5 × 10-11 2.5 × 10-9 6.5 × 10-8 1.1 × 10-7

1 gpm = 3.7852 lpm 

1 in. = 25.4 mm 

2.3 Risk-Informed Inservice Inspection Evaluations 

Risk-informed inservice inspection evaluations have estimated failure (leak and rupture) frequencies. 

This work has been focused on a detailed component level to a greater extent than the more common 

system-level evaluations used for PRAs. Specific degradation mechanisms are identified for components 

and welds, usually on the basis of past operating experience. Future degradation mechanisms and 

increased failure rates for known mechanisms are addressed by periodic updating of the inspection 

programs. 

Most U.S. plants have adopted risk-informed inservice inspection programs for their piping systems. As 

such, damage mechanism assessments have been performed for these plants, and failure frequencies have 

been estimated. In some cases, the adopted methodology has been qualitative, with failure frequencies 

being classified in terms of categories (e.g., high, medium, and low). In other cases, the methodology 

(such as described by the Westinghouse Owners Group [1997]) has involved extensive probabilistic 

fracture mechanics calculations. Details of these plant-specific calculations have not been published but 

are available to the NRC through onsite inspection. Instead, the methodology and high-level results of the 

evaluations are submitted to the NRC as part of the review and approval process for the proposed riskinformed inservice inspection programs. 

2.4 Other Studies 

Other estimates of failure frequencies have been performed on a system- or plant-wide basis either for 

specific plants or on a more generic basis (such as PWR or BWR plants). Various methods have been 

applied, including expert elicitation (Vo et al. 1991). The very early studies, such as for WASH-1400, 

were based on nonnuclear data from other industries (USNRC 1975). 
2.5 

Other studies have used databases of reported pressure boundary failure events at operating plants. A 

Swedish evaluation of piping failure frequencies was performed for the Barsebäck-1 plant (Lydell 1999) 

in coordination with the Swedish Nuclear Inspectorate (SKI) using an international piping failure 

database that included Class 1, Class 2, and Class 3 piping. The SKI effort resulted in guidelines for 

using a database to evaluate piping reliability in terms of important influence and attribute factors. The 

guidelines include consideration of each failure mechanism separately (e.g., IGSCC, vibration fatigue, 

water hammer). All the relevant mechanisms are then combined to develop total LOCA frequencies. 

Only medium and large LOCA frequencies were determined in the SKI effort by considering events for 

these modes from failure mechanisms represented within the database. Maintenance of the database has 

nevertheless included reports of nonleaking flaws discovered during inspections to provide insights into 

the types of piping locations that have a potential for more significant modes of failure. 

The integrity of steam generator tubes has been a significant aging issue given the various degradation 

mechanisms that have been active at PWR plants. Methods have been developed to predict probabilities 

of burst for degraded tubes. These methods have been based on statistical correlations of experimental 

data from burst tests of degraded tubes. Correlations have related burst pressures to such parameters as 

crack dimensions (Keating et al. 1995) or voltage amplitudes as measured during eddy current 

inspections. Voltage-based correlations have been extensively used in connection with criteria for 

plugging degraded tubes, but the methodologies have been largely described within submittals from plant 

licensees rather than in the open literature. In all cases, the probabilistic evaluations for steam generator 

tubes have used statistical treatments of burst testing rather than probabilistic structural mechanics models 

as addressed by the current report. 
3.1 

3 Review of Existing Probabilistic Fracture Mechanics 

Models for Reactor Components 

Since the early 1980s, the NRC has supported the development and application of probabilistic fracture 

mechanics codes for predicting failure probabilities of reactor pressure boundary components. These 

codes have addressed the structural integrity of both reactor pressure vessels and reactor piping systems. 

This section describes some of these codes with an emphasis on those that have been developed and 

applied in NRC research programs. Various available codes are discussed also in Section 5 in connection 

with published failure probabilities as predicted for a range of degradation mechanisms. 

3.1 Reactor Pressure Vessel Codes 

Early work in the 1980s supported the development of the pressurized thermal shock (PTS) rule (10 CFR 

50.61) that defined acceptable limits for neutron embrittlement for materials of reactor pressure vessels. 

Two codes were developed during this time period: OCA-P by Oak Ridge National Laboratory 

(Cheverton and Ball 1984) and VISA-II by PNNL (Simonen et al. 1986). Later work beginning in the 

1990s and continuing to the present has focused on the FAVOR code (Dickson 1994; Dickson et al., 

NRC, unpublished(a)). These codes as well as similar codes developed both in the United States and 

overseas are used to predict brittle fracture in irradiated reactor pressure vessel beltline materials. The 

VISA-II code was used to support the current PTS rule, and FAVOR was used to support recently 

proposed PTS rule changes. 

Another probabilistic fracture mechanics code (VIPER) was developed to address issues specifically 

related to welds in boiling water reactor vessels (Tang et al. 1999). This code used a Monte Carlo 

approach and a fracture mechanics model similar to those used in the VISA-II and FAVOR codes. 

Special model features were incorporated to address weld residual stresses and crack initiation and growth 

by stress corrosion mechanisms. 

3.2 Reactor Piping Codes 

A number of codes for predicting failure probabilities of reactor piping have been developed. Each is 

described in the following sections. 

3.2.1 PRAISE 

The PRAISE code (Harris et al. 1981, 1986; Harris and Dedhia 1992) was developed as an NRC-funded 

effort starting in the early 1980s to predict failure probabilities for reactor piping systems. The original 

work was directed at fatigue failures of large-diameter reactor coolant piping due to the growth of weld 

fabrication flaws. The first version of PRAISE was developed for the NRC by Lawrence Livermore 

National Laboratory to support the seismic safety margins project and was applied in early work to 

address PWR piping; it was later extended to address IGSCC in BWR stainless steel piping welds. 

 

(a) Dickson TL, PT Williams, and S Yin. Unpublished. Fracture Analysis of Vessels – Oak Ridge, 

FAVOR, v04.1, Computer Code: User’s Guide. NUREG/CR-6855, ORNL/TM-2004/245, October 

2004, in review by the NRC. 
3.2 

PRAISE has seen continued applications over the years with ongoing updates to include capabilities 

needed to address specific piping integrity issues, most notably the initiation of fatigue cracks (Khaleel 

et al. 2000). 

The enhanced version of PRAISE has allowed for crack initiation at multiple sites around the 

circumference of girth welds and simulated the linking of adjacent cracks to form longer cracks more 

likely to cause large leaks and pipe ruptures. By coupling S-N crack initiation and fatigue crack growth 

models, simulations allow cracks to initiate and grow in a more realistic manner. 

In the early 1990s with the evolution of computer technology a version of PRAISE was developed to run 

on personal computers (Harris and Dedhia 1992). The mid 1990s saw the development of methods for 

risk-informed inservice inspection, and there were many new applications of PRAISE. During this 

period, a commercial version of PRAISE was made available by Dr. David Harris of Engineering 

Mechanics Technology that simplified the input to the code with an interactive front end (Harris and 

Dedhia 1998). PNNL also performed numerous calculations with PRAISE that applied probabilistic 

fracture mechanics to examine how different inservice inspection strategies could improve component 

reliability (Khaleel and Simonen 1994a, 1994b; Khaleel et al. 1995; Simonen et al. 1998; Simonen and 

Khaleel 1998a, 1998b). These studies evaluated changes in failure probabilities when changes to 

inservice inspection requirements are implemented at operating nuclear power plants. Other work at 

PNNL for the NRC (Khaleel et al. 2000) was directed at fatigue critical components with the potential to 

attain calculated fatigue usage factors in excess of design limits (usage factors greater than unity). 

PRAISE was used also to develop the technical basis for changes to Appendix L of the ASME Code 

Section XI that addresses fatigue critical locations in pressure boundary components (Gosselin et al. 

2005). 

In summary, the PRAISE code has been extensively documented, has been successfully applied to a range 

of structural integrity issues, and has been available since the 1980s as a public domain computer code. 

However, the code has not been systematically maintained and upgraded in an ongoing manner. 

Upgrades have met the needs of immediate applications of the code and, as such, have served to fill very 

specific gaps in capabilities of PRAISE. 

3.2.2 PRO-LOCA 

In 2003, the NRC began the development of a new code called PRO-LOCA (Rudland et al. 2006; 

Rudland et al., NRC, unpublished(a)) that is intended to include advances in fracture mechanics models 

and thereby become the successor to the PRAISE code. The development of the PRO-LOCA code was 

motivated by an NRC need to address issues related to LOCA events. The need for an improved 

probabilistic fracture mechanics code became evident during an expert elicitation process funded by the 

NRC (Tregoning et al. 2005) to establish estimates of LOCA frequencies along with quantification of the 

uncertainties in the estimates. Development of the code is expected to continue. Part of this future work 

would include the preparation of documentation of the code with detailed user instructions needed to 

support the release of the code to external organizations. A report on the status of PRO-LOCA has been 

 

(a) Rudland DL, H Xu, G Wilkowski, N Ghadiali, F Brust, and P Scott. Unpublished. Evaluation of 

Loss-of-Coolant Accident (LOCA) Frequencies Using the PRO-LOCA Code, NUREG/CR-XXXX, 

currently under review. 
3.3 

prepared by EMC2

 (Rudland et al., NRC, unpublished(a)), which describes features and the technical bases 

for PRO-LOCA. Like PRAISE, the PRO-LOCA code addresses the failure mechanisms associated with 

both preexisting cracks and service-induced cracks. Both fatigue and IGSCC are addressed. The model 

for fatigue is essentially the same as that in PRAISE, whereas the IGSCC model differs significantly from 

the model in PRAISE. PRO-LOCA can also predict failure probabilities for PWSCC. Other improved 

capabilities are in the areas of leak rate predictions and the prediction of critical/unstable crack sizes. 

Many of the enhancements are based on the results of some 20 years of NRC-supported research on the 

integrity of degraded piping. PRO-LOCA has also incorporated an improved basis for simulating weld 

residual stresses. 

3.2.3 Other Piping Codes 

A number of other probabilistic fracture mechanics codes have been developed by various organizations 

to calculate failure probabilities for piping. The SRRA code (Westinghouse Owners Group 1997; Bishop 

1993, 1997) as developed by Westinghouse uses approaches similar to those of the PRAISE code. 

However, the fatigue model is limited to failures associated with the growth of preexisting fabrication 

flaws. Fatigue crack initiation has been approximated, in a conservative manner, by assuming very small 

preexisting cracks, but only one crack per weld. Stress corrosion cracking is similarly treated by 

postulating very small preexisting cracks with crack growth predicted using a crack growth law consistent 

with growth rates for stress corrosion cracks (SCCs). The SRRA code includes an importance sampling 

procedure that gives reduced computation times compared to the Monte Carlo approaches used by PROLOCA and PRAISE. The SRRA code also simulates uncertainties in a wide range of parameters, 

including operating stresses and pipe dimensions. 

A European effort (Brickstad et al. 2004, NURBIM) has reviewed and tested a number of codes including 

PRAISE as part of an international benchmarking study. Included were a Swedish code NURBIT 

(Brickstad and Zang 2001), the PRODIGAL code from the United Kingdom (Bell and Chapman 2003), a 

code developed in Germany by GRS (Schimpfke 2003), a Swedish code ProSACC (Dillstrom 2003), and 

another code (STRUEL) from the United Kingdom (Mohammed 2003). A detailed discussion of the 

NURBIM study is beyond the scope of this report. However, it appears that none of the other 

benchmarked codes provides capabilities that are significantly different from or superior to the 

capabilities of PRAISE or PRO-LOCA. In any case, the predictions of all such codes are limited in large 

measure by uncertainties in establishing the input parameters needed to apply the code. 

 

(a) Rudland DL, H Xu, G Wilkowski, N Ghadiali, F Brust, and P Scott. Unpublished. Evaluation of 

Loss-of-Coolant Accident (LOCA) Frequencies Using the PRO-LOCA Code, NUREG/CR-XXXX, 

currently under review. 
4.1 

4 Open Literature Search 

The Proactive Materials Degradation Assessment (PMDA) (USNRC 2007) identified components of 

nuclear power plant systems and ranked their susceptibility or likelihood for damage due to various 

degradation mechanisms. An open literature search was performed to locate technical papers and reports 

with information relevant to failure frequencies for the components and degradation mechanisms 

identified in the NRC PMDA program. The search used computerized databases available through the 

Hanford Technical Library in Richland, Washington, and made use of support from information 

specialists from the library staff. A large list of relevant documents was generated and reviewed to 

identify documents with specific information on component failure frequencies with specific attention to 

results generated by application of probabilistic structural mechanics models. 

4.1 Methodology 

Research and development relevant to the structural integrity of reactor systems (reactor coolant, 

emergency core cooling, steam and power conversion, and support and auxiliary) of PWRs and BWRs 

has been ongoing since the advent of nuclear facilities, both military and civilian. Activities have 

included programs at nuclear laboratories and material laboratories in many countries worldwide. While 

some results such as new material compositions may not be freely available, numerous other results and 

accounts of operating experience have been published internationally in various journals and research 

reports. Most of these publications are electronically listed in various comprehensive bibliographic 

databases such as Compendex, Corrosion Abstracts, METADEX, and Science Research Connection, and 

those maintained by the National Technical Information Service (NTIS) and the NRC ADAMS Public 

Records System and Public Legacy Databases. 

An initial computer-aided search was performed on every subgroup of the different systems of PWRs and 

BWRs. This search yielded very few results. The search criterion was revised by grouping different 

subgroups in PWRs and BWRs to do a broad search of the above databases. For example, different 

subgroups such as the hot leg piping, the crossover leg piping, and the cold leg piping were grouped into a 

single search item. Keywords such as nuclear power plants and major group names such as pipes along 

with the various degradation mechanisms were used. All the results were retrieved and downloaded into 

an EndNote database. EndNote is a bibliographic management software tool for publishing and 

managing bibliographies. This EndNote software was subsequently used for searching specific topics. 

The following open literature databases were searched.

Compendex (computerized engineering index) is a database that provides international coverage of 

literature in the engineering field, including materials science and metallurgy, bioengineering, air and 

water pollution, and solid waste and hazardous waste management. Citations are drawn from 

5000 journals, technical reports, and conference papers and proceedings. 

Corrosion Abstracts provides the world’s most complete source of bibliographic information in the area 

of corrosion science and engineering. International sources of literature are scanned and abstracted in the 

areas of general corrosion, testing, corrosion characteristics, preventive measures, materials construction 

and performance, and equipment for many industries. 
4.2 

METADEX (Metals Abstracts/Alloy Index) is a comprehensive source covering the worldwide 

literature on metals and alloys, including properties, manufacturing, applications, and development. 

The National Technical Information Service (NTIS) provides a multidisciplinary database covering 

U.S. government-sponsored research and worldwide scientific, technical, engineering, and businessrelated information. The Government Report Announcements & Index (GRA&I) is completely covered 

by the NTIS database. 

Science Research Connection provides approximately 4 million bibliographic records and over 

125,000 full-text documents spanning more than six decades of DOE research.

The NRC ADAMS Public Records System and Public Legacy Databases provide technical 

information about nuclear facilities and NRC licensees and is a comprehensive research center for NRC 

documents. 

Search strategies were customized to fit the databases searched. The search strategy for Compendex, 

Corrosion Abstracts, METADEX, and NTIS is very straightforward. Boolean operators like “and” and 

“or” were used to limit the searches. Science Research Connection follows a similar strategy, whereas 

NRC’s ADAMS database search engine is simple and does not handle complex searching so multiple 

simple searches were run to retrieve the relevant important documents. Six sets of different searches 

covering all of the degradation mechanism, were performed, and the results were downloaded into 

EndNote Libraries. 

4.1.1 Search Strategy for Compendex, Corrosion Abstracts, Metadex, and NTIS Search 

The following six sets of keywords were used: 

• “nuclear power plant*” AND (pipe OR pipes OR piping OR nozzle*) AND (erosion OR fatigue OR 

corrosion OR pitting OR MIC OR SCC OR PWSCC OR FAC) AND limit by date to 1980–present 

• “nuclear power plant*” AND (vessel* OR steam generator* OR pressurizer*) AND (corrosion OR 

pitting OR fatigue OR SCC OR PWSCC) AND limit by date to 1980–present 

• (BWR OR boiling water reactor*) AND core shroud* AND (304 OR “stainless steel” OR “alloy 182” 

OR weld*) AND (SCC OR PWSCC or IGSCC OR fatigue of corrosion) AND limit by date to 1980–

present 

• (BWR OR boiling water reactor*) AND jet pump* AND (weld* OR “inconel 600” OR “alloy 600” 

OR “alloy 182” or diffuser* OR bracket* or support* or brace* OR adapter OR riser) AND (SCC OR 

PWSCC or IGSCC OR fatigue of corrosion) AND limit by date to 1980–present 

• (BWR OR boiling water reactor*) AND steam AND (separator OR dryer) AND (SCC OR PWSCC 

or IGSCC OR fatigue OR corrosion) limit by date to 1980–present 
4.3 

• (BWR OR boiling water reactor*) AND condensate storage tank* AND limit by date to 1980–

present. 

4.1.2 Search Strategy for ADAMS 

This search required a specialized strategy. The Public Records System contains documents and records 

from November 1999 to the present, and the Public Legacy System contains records of documents 

published before November 1999. The advanced search option available in ADAMS database is a fulltext search and proved too broad. The simple search used automatic truncation and found many irrelevant 

documents. As a result, multiple searches of the following keywords were done in each database. In the 

Public Legacy System, results were limited by date to the period from 1980 through 1999. The sets of 

keywords used were 

• pipe erosion, pipe fatigue, pipe corrosion, pipe pitting, pipe SCC, piping erosion, piping fatigue, 

piping corrosion, piping pitting, and piping SCC 

• nozzle erosion, nozzle fatigue, nozzle corrosion, nozzle pitting, and nozzle SCC 

• vessel corrosion, vessel pitting, vessel fatigue, and vessel SCC 

• steam generator corrosion, steam generator pitting, steam generator fatigue, and steam generator SCC 

• pressurizer corrosion, pressurizer pitting, pressurizer fatigue, and pressurizer SCC 

• core 304 SCC, core stainless steel SCC, core alloy 182 SCC, core weld SCC, core 304 fatigue, core 

stainless steel and fatigue, core alloy 182 fatigue, core weld fatigue, core 304 corrosion, core stainless 

steel corrosion, core alloy 182 corrosion, and core weld corrosion 

• jet pump weld SCC, jet pump stainless steel SCC, jet pump inconel 600 SCC, jet pump alloy 600 

SCC, jet pump alloy 182 SCC, jet pump diffuser SCC, jet pump bracket SCC, jet pump support SCC, 

jet pump brace SCC, jet pump adapter SCC, jet pump riser SCC, jet pump weld fatigue, jet pump 

stainless steel fatigue, jet pump inconel 600 fatigue, jet pump alloy 600 fatigue, jet pump alloy 182 

fatigue, jet pump diffuser fatigue, jet pump bracket fatigue, jet pump support fatigue, jet pump brace 

fatigue, jet pump adapter fatigue and jet pump riser fatigue, jet pump weld corrosion, jet pump 

stainless steel corrosion, jet pump inconel 600 corrosion, jet pump alloy 600 corrosion, jet pump alloy 

182 corrosion, jet pump diffuser corrosion, jet pump bracket corrosion, jet pump support corrosion, 

jet pump brace corrosion, jet pump adapter corrosion, and jet pump riser corrosion 

• steam dryer fatigue, steam dryer corrosion, and steam dryer SCC, steam separator fatigue, steam 

separator corrosion, and steam separator SCC. 

4.1.3 Search Strategy for Science Research Connection 

In searching Science Research Connection, we used the following six sets of keywords with a date limit 

of 1980–2006: 
4.4 

• All Fields: “nuclear power plant” OR “nuclear power plants” 

Bibliographic Data: erosion OR fatigue OR corrosion OR pitting OR MIC or SCC OR PWSCC OR 

FAC 

Title Data: pipe or pipes or piping 

• All Fields: “nuclear power plant” OR “nuclear power plants” 

Bibliographic Data: corrosion OR pitting OR fatigue OR SCC OR PWSCC 

Title Data: vessel OR vessels OR “steam generator*” OR pressurizer* 

• All Fields: “BWR” OR “boiling water reactor*” 

Bibliographic Data: corrosion OR fatigue OR SCC OR PWSCC OR IGSCC 

Title Data: core shroud 

• All Fields: “BWR” OR “boiling water reactor*” 

Bibliographic Data: corrosion OR fatigue OR SCC OR PWSCC OR IGSCC 

Title Data: jet pump* 

• All Fields: “BWR” OR “boiling water reactor*” 

Bibliographic Data: corrosion OR fatigue OR SCC OR PWSCC OR IGSCC 

Title Data: steam AND (separator* OR dryer*) 

• All Fields: “BWR” OR “boiling water reactor*” 

Title Data: condensate storage tank*. 

4.2 Literature Search Results 

The above open literature searches identified more than 7500 publications from 1980 to the present. 

Approximately 60% of the documents were related to corrosion and pitting mechanisms. The remaining 

documents were equally divided between fatigue and stress corrosion cracking mechanisms. The vast 

majority of the references addressed various mechanistic aspects of degradation mechanisms and applied 

deterministic rather than probabilistic models. With regard to corrosion mechanisms other than SCC 

(e.g., general corrosion, pitting), the literature search did not identify any probabilistic calculations that 

applied mechanistic models to address the types of failures observed in the field. Although the 

probabilistic treatments and characterizations of fatigue and stress corrosion cracking are well 

documented in the open literature, a relatively limited number of publications documented the results of 

plant-specific and component-specific probabilistic evaluations based on design or actual operating 

stresses. The more relevant of these are discussed in Section 5 of this report. 

In many cases, the literature search did not identify numerous plant-specific studies completed by 

industry owners’ groups and by research organizations such as the EPRI Materials Reliability Program. 

These results are typically not available through the open literature. Also, plant-specific calculations, 

such as those performed in the support of risk-informed evaluations, were not publicly available. 

However, portions of these results are sometimes included in topical reports submitted for NRC approval 

or in selected papers in the literature. Examples can be found in the Westinghouse Owners Group topical 

report on RI-ISI (1999) and work reported by Bishop and McNeil (1999). Additionally, failure 
4.5 

probability estimates have been based on statistical analyses of industry service experience. Examples 

include work reported by Fleming et al. (1999); Lydell (1999); Attwood et al. (1999); Fleming (2004); 

and Tregoning et al. (2005). 
5.1 

5 Failure Probabilities for Reactor Components 

This section includes summaries of failure probability estimates obtained by applying probabilistic 

fracture mechanics models to different reactor components and for different degradation mechanisms. 

The information was obtained from reviews of public domain literature and is organized by degradation 

mechanism and plant type (BWR or PWR). 

5.1 Stress Corrosion Cracking 

The three variations of stress corrosion cracking of concern to reactor components are intergranular, 

transgranular, and primary water cracking, with each type discussed separately in this section. 

5.1.1 Intergranular Stress Corrosion Cracking 

In general, intergranular stress corrosion cracking (IGSCC) results from a combination of sensitized 

materials (caused by a depletion of chromium in regions adjacent to the grain boundaries in weld heateffected zones), high stress (residual welding stresses), and a corrosive environment (high level of oxygen 

or other contaminants). IGSCC is encountered most frequently in wrought Type 304 or 316 austenitic 

stainless steels that become sensitized through the welding process and are subjected to BWR operating 

environments. To a lesser extent, IGSCC has been observed also in wrought austenitic stainless steel 

PWR piping having high dissolved oxygen content and stagnant flow (e.g., stagnant, oxygenated borated 

water systems). The susceptible areas extend into the base material a few millimeters beyond either side 

of the weld—the weld heat-affected zone. Generally materials resistant to sensitization from welding are 

not considered susceptible to IGSCC. The cause of and experience with IGSCC in nuclear power plants 

is well documented in Shao et al. (1979, 1980); Kassir et al. (1985); Miraglia (1988); Hazelton and Koo 

(1988); and Strosnider et al. (2000). 

By 1987 and 1988, most operating BWR plant operators had implemented one or more IGSCC mitigation 

programs together with IGSCC inspection programs as outlined in the NRC Generic Letter 88-01 

(Miraglia 1988) and NUREG-0313 Revision 2 (Hazelton and Koo 1988). Today most if not all BWR 

plants operate with hydrogen water chemistry, and weld residual stresses are kept at a minimum by using 

improved welding techniques, applying weld overlays or by induction heat stress improvement of existing 

welds. Although not generally used in the United States, austenitic stainless steels with additions of 

niobium or titanium, referred to as stabilized stainless steels, have become the preferred material in 

German BWR plants. 

Summarized in Table 5.1 is the worldwide service experience involving IGSCC in the BWR operating 

environment. This tabulation is based on nuclear power plant (NPP) failure data reported in the 

proprietary PIPExp database described in Appendix D to NUREG-1829 (Tregoning et al. 2005). The 

weld failure data separated into two groups: stabilized and unstabilized austenitic stainless steels. 
5.2 

Table 5.1 Summary of IGSCC Service Experience in BWR Nuclear Power Plants 

Number of Reported Weld Failures Attributed to IGSCC(a) 

1970–1987 1988–2005 1970–2005 

Type of Austenitic 

Stainless Steel 

Part 

ThroughWall ThroughWall 

Part 

ThroughWall ThroughWall 

Part 

ThroughWall ThroughWall 

Stabilized(b) -- -- 209 12 209 12 

Unstabilized(c) 623 139 195 33 818 172 

(a) Through-wall leak rates range from perceptible (no active leakage during normal operation) to small active 

leakage (Duane Arnold weld failure on June 14, 1978, to date the most significant failure, produced a 11.4-

L/min (3-gpm) leakage. Weld failure at Santa Maria de Garona, a Spanish BWR plant, on February 12, 

1980, produced a 3.0-L/min (0.8-gpm) leakage. 

(b) Stabilized austenitic stainless steel failures report for 15 non-U.S. operating NPPs. 

(c) Unstabilized austenitic stainless steel failures report for 57 plants U.S. and non-U.S. operating NPPs. 

Numerous published reports document results of PFM-based ISGCC failure probabilities for major welds 

in BWR ASME Code Class 1 and 2 piping. The published probabilities span a wide range from lowerbound PFM computer code truncation values of 1 × 10-11 or less per weld per calendar year to upperbound estimates approaching 1 × 10-1 per weld per calendar year. These results reflect inherent modeling 

uncertainties as well as different modeling capabilities and analysis assumptions. For the most part, the 

more significant IGSCC probabilistic fracture mechanics studies in the open literature either employed or 

were benchmarked against versions of IGSCC models developed for the PRAISE computer code. 

The early IGSCC probabilistic fracture mechanics evaluations (Holman and Chou 1988) were based on 

the semi-empirical IGSCC model developed for the PRAISE computer code as described in Harris et al. 

(1985, 1986a, 1986b). In the first paper the crack initiation time and crack growth rates are correlated 

against damage parameters that consolidate the separate influences of environment, applied loads, 

residual stresses, and material sensitization. The models were benchmarked by comparing predicted leak 

rates under nominal BWR applied load conditions against actual crack and leak data. To improve the 

agreement of the PRAISE model predictions with observed field experience, the model was tuned by 

adjusting residual stress magnitudes (using a multiplication factor). 

The original values of residual stress adjustment factors selected by Harris et al. (1985) for large, 

intermediate, and small pipe sizes are provided in Table 5.2. The leak and double-ended guillotine 

(DEGB) probabilities for the large, intermediate, and small pipe sizes and using residual stress adjustment 

factors in Table 5.2 are summarized in Table 5.3. The comparison of the field observations of leak 

probabilities and PRAISE predictions for the large pipe range (outside diameters greater than 508 mm 

[>20 in.]) and various residual stress adjustment factor values is shown in Figure 5.1. 

The results in Figure 5.1 show that it was necessary to substantially reduce the residual stress levels to 

15-20% of the values estimated from measurements. 
5.3 

Table 5.2 Residual Stress Adjustment Factors, f (Harris et al. 1985) 

Line Size Range Outside Diameter Value of f 

Large >508 mm (>20 in.) 0.15 

Intermediate 254–508 mm (10–20 in.) 0.30 

Small <254 mm (<10 in.) 0.20 

Table 5.3 Summary of Leak and DEGB Probabilities as a Function of Time for Three Pipe Size 

Ranges (Harris et al. 1985) 

Time Leak Probabilities DEGB Probabilities 

(Years) Large Intermediate Small Large Intermediate Small 

2 <3.0 × 10-5 2.4 × 10-4 7.0 × 10-4 -- <2.0 × 10-5 <1.0 × 10-4

5 6.0 × 10-4 8.3 × 10-4 1.2 × 10-2 -- 4.0 × 10-5 1.0 × 10-4

10 3.4 × 10-4 5.4 × 10-2 6.7 × 10-2 -- 4.0 × 10-5 3.0 × 10-4

15 7.7 × 10-3 1.2 × 10-1 1.4 × 10-1 <3.0 × 10-5 6.0 × 10-5 4.0 × 10-4

20 1.2 × 10-2 1.8 × 10-1 1.9 × 10-1 3.0 × 10-5 6.0 × 10-5 5.0 × 10-4

The original calibration of the PRAISE model in Harris et al. (1985) was redone by Khaleel et al. (1995). 

The original calculations predicted substantial levels of material damage from loading and unloading 

events (i.e., complete startup and shutdown of the plant) that used a model that applied strain-to-failure 

data from constant extension rate tests. A review of the damage model concluded that these predictions 

were extremely conservative and were inconsistent with more recent insights into stress corrosion 

cracking mechanisms. In the revised calculations, the loading/unloading events were decreased from 

once per year to once per 40 years, which essentially removed the contribution of these events to the 

calculated failure probabilities. Figure 5.2 compares the original PRAISE cumulative leak probabilities 

(Harris et al. 1985) and results obtained by Khaleel et al. (1995) for various values of residual stress 

adjustment factors and plant loading/unloading frequencies. In this case, the adjusted residual stress level 

used to limit the disagreement between predicted and observed leak probabilities was set at 75% of their 

original values (f = 0.75). The resulting predictions had a much more rational basis and were in very 

good agreement with operational data for time periods beyond 6 years. The less satisfactory level of 

agreement for time periods less than 6 years can be attributed in large measure to a lack of observed 

failure events for the early periods of plant operation. 

NUREG/CR-4792 (Holman and Chou 1988) describes early stress corrosion cracking probabilistic 

fracture mechanics evaluations for BWR reactor coolant piping performed for the NRC by Lawrence 

Livermore National Laboratory. In these evaluations, leak and DEGB probabilities were estimated in the 

recirculation loop piping of a representative Mark I BWR plant. The probabilistic fracture mechanics 

evaluations of NUREG/CR-4792 were based on the semi-empirical IGSCC model developed for the 

PRAISE computer code as described in Harris et al. (1985, 1986a, 1986b). 
5.4 

Figure 5.1 Field Observations of Leak Probabilities and PRAISE Predictions at Various Residual 

Stress Adjustment Factor Values (f) for 304 Stainless Steel Pipes Having Outside 

Diameters Greater Than 508 mm (>20 in.) (Harris et al. 1985) 

Figure 5.2 Sample Comparison of PRAISE Cumulative Leak Probabilities (Harris et al. 1985) and 

Results Obtained by Khaleel et al. (1995) for Various Values of Residual Stress 

Adjustment Factors and Plant Loading/Unloading Frequencies 
5.5 

The reactor recirculation system in a Mark I BWR plant has two loops, as shown in Figure 5.3. The 

recirculation water from the reactor vessel is delivered to the reactor coolant recirculation pump via a 

suction line 711 mm (28 in.) in diameter. The reactor coolant recirculation pump discharges into the 

reactor pressure vessel via discharge piping 711 mm (28 in.) in diameter and header piping 559 mm 

(22 in.) in diameter header piping, plus five 305-mm (12-in.) NPS risers. Although not shown in 

Figure 5.3, a bypass line 102 mm (4 in.) in diameter with a bypass valve is connected to the discharge line 

on either side of the reactor coolant recirculation pump discharge valve. 

Figure 5.3 Mark I BWR Reactor Recirculation System Piping 

The evaluation considered three recirculation system configurations: 1) an existing “old” configuration 

with 51 welds per loop and SA-240 TP304 stainless steel piping; 2) an existing “old” configuration with 

51 welds per loop and SA-358 TP316 stainless steel piping; and 3) a replacement “new” configuration 

with SA-358 TP316 stainless steel piping and 30 welds per loop. The material types, dimensions, and 

number of welds for the recirculation piping in the NUREG/CR-4792 evaluations are presented in 

Table 5.4. 

The bar charts in Figure 5.4 show the relative contribution made by each NUREG/CR-4792 weld type to 

the overall system probabilities of leak and DEGB. The relative contribution of different weld types is 

illustrated further in Figure 5.5, which shows weld-by-weld leak probabilities for the existing loop 

configuration. Finally, a comparison of the NUREG/CR-4792 cumulative DEGB probabilities between 

the old configuration with TP304 stainless steel piping and the new configuration with TP316 stainless 

steel piping is shown in Figure 5.6. 
5.6 

Table 5.4 NUREG/CR-4792 Mark I BWR Recirculation System Configurations 

Material Pipe Suction Discharge Header Riser Bypass 

Nominal Pipe Size (in.) 28 28 22 12 4 

Outside Diameter (in.) 28.169 28.519 22.003 12.706 4.5 

Wall Thickness (in.) 1.51 1.326 1.038 0.631 0.337 

SA-240 

TP304 

Stainless 

Steel Number of Welds 10 6 5 20 10 

Nominal Pipe Size (in.) 28 28 22 12 NA 

Outside Diameter (in.) 28.000 28.000 22.000 12.750 NA 

Wall Thickness (in.) 1.209 1.390 1.750 1.688 NA 

SA-358 

TP316 

Stainless 

Steel Number of Welds 11 5 2 12 NA 

25.4 mm = 1 in. 

Figure 5.4 NUREG/CR-4792 Relative Contributions of Various Weld Types to BWR Recirculation 

System Probabilities of (a) Leak and (b) DEGB 

Figure 5.5 NUREG/CR-4792 Cumulative Leak Probabilities for BWR Recirculation System Weld 

Types in the “Old” Configuration with TP304 Stainless Steel Piping 
5.7 

Figure 5.6 NUREG/CR-4792 Cumulative DEGB Probabilities Between the Old Configuration with 

TP304 Stainless Steel Piping and the New Configuration with TP316 Stainless Steel 

Piping 

As mentioned previously, most operating BWR plants have implemented one or more of the IGSCC 

mitigation programs outlined the NRC Generic Letter 88-01 (Miraglia 1988) and NUREG-0313 

Revision 2 (Hazelton and Koo 1988). These include minimizing the weld residual stresses through the 

application of improved welding techniques, use of weld overlays, or by the implementation of induction 

heat stress improvement or mechanical stress improvement process (MSIP) on existing welds. In addition 

most, if not all, operating plants control oxygen concentration below 20 ppb (parts per billion) by 

hydrogen addition to the reactor coolant. 

In work performed by Failure Analysis Associates (1990), enhancements to the original PRAISE stress 

corrosion cracking models were made to account for improved piping reliability associated with the 

midlife implementation of IGSCC mitigation programs. These enhancements to PRAISE included 

mechanistic models that account for mid-life changes in: 1) residual stresses due to the implementation of 

thermal and mechanical stress improvement methods, 2) oxygen concentrations due to implantation of H2 

water chemistry controls, coolant conductivity. Enhancements made to account for mid-life changes 

associated with piping replacements included changes in pipe wall thickness, dead weight stresses, 

thermal expansion stresses, and vibration stresses. 

Figure 5.7 presents an example of the predicted effects of midlife dissolved oxygen concentrations on the 

cumulative leak probability of a BWR 304 stainless steel pipe. These results show that reducing the 

steady-state oxygen levels from 0.2 to 0.05 ppm decreases the cumulative leak probability by a factor of 2 

for the following 10 operating years. Figure 5.8 presents the results for cases in which residual stresses in 

the weld region were changed by treatment with either induction heat stress improvement or mechanical 

stress improvement. 
5.8 

Figure 5.7 Effect of Lowering Oxygen Concentration at Midlife (20 years) in 102-mm (4-in.) 

Schedule 80 304 Stainless Steel BWR Pipe. The numbers by each line are the oxygen 

concentration (ppm) in the coolant during plant startup and steady-state operation, 

respectively (Failure Analysis Associates 1990). 

Figure 5.8 Effect of Reducing Residual Stress at Midlife (20 years) in 102-mm (4-in.) Schedule 80 

Type 304 Stainless Steel BWR Pipe (Failure Analysis Associates 1990) 
5.9 

Many of the early studies focused on the probability of a DEGB; however, the DEGB is widely 

recognized as an extremely unlikely event. Therefore, the NRC is evaluating risk-informed design-basis 

break size requirements for operating commercial nuclear power plants. A central consideration in 

selecting a risk-informed break size is an understanding of conditional failure frequency as a function of 

break size (i.e., leak rate). 

In a recent study (Tregoning et al. 2005), LOCA frequency estimates as a function of break size and 

operating time were developed using an expert elicitation process. This process was designed to 

consolidate service history data and insights from probabilistic fracture mechanics studies with 

knowledge of plant design, operation, and material performance. As part of this study, probabilistic 

fracture mechanics calculations were performed on a 304-mm (12-in.) and 711-mm (28-in.) Schedule 80 

BWR recirculation piping. The recirculation piping was A-358 Class 1 Type 304 stainless steel. For the 

304-mm (12-in.) pipe, analyses were performed assuming no remedial actions were taken and also 

assuming a weld overlay was performed at 20 years. No remedial action was the only case considered for 

the 711-mm (28-in.) line. Normal operating stresses were assumed to be 140.8 MPa (20.41 ksi) in the 

304-mm (12-in.) line and 65.4 MPa (9.48 ksi) in the 711-mm (28-in.) line. The default residual stress 

distributions in NUREG/CR-4792 were assumed for the case when no mitigating actions (i.e., weld 

overlay) were performed. A linear approximation of the axisymmetric through-wall residual stress 

distribution in EPRI (1988) was assumed for the post-weld overlay. Table 5.5 summarizes the results for 

the 304-mm (12-in.) weld and the 711-mm (28-in.) weld in the recirculation system. 

The beneficial effects of a weld overlay at 20 years are shown more clearly in Figure 5.9, which provides 

a plot of the cumulative probability of a leak exceeding 378.5 L/min (100 gpm) as a function of time. The 

slopes of the lines in Figure 5.9 are the leak frequencies; and the slope at 40 years with no overlay is 

approximately seven times that with an overlay at 20 years. 

5.1.2 Transgranular Stress Corrosion Cracking 

In addition to IGSCC, transgranular stress corrosion cracking (TGSCC) failures also have been reported 

for highly strained nonsensitized austenitic stainless steel piping base materials subjected to aggressive 

environments (e.g., oxygen/oxidizing species in conjunction with the presence of halides in PWR and 

BWR nuclear power plant reactor systems). TGSCC has occurred generally in regions of stagnant 

coolant at high elevation, where a high concentration of oxygen might have been present (Shah et al. 

1998). For example, TGSCC was found in penetrations at the top of the reactor pressure vessel head. 

TGSCC pipe failures have also been reported for stainless steel piping subject to external chloride 

contamination from PVC tape applied to piping, and small- to medium-diameter cold bent piping where 

lubricants used in the bending machine contaminated the inside pipe wall and the bending process 

introduced high stresses. Summarized in Table 5.6 are the numbers of worldwide nuclear power plant 

service failures involving TGSCC; as reported in the proprietary PIPExp database described in Appendix 

D to NUREG-1829 (Tregoning et al. 2005). 
5.10 

Table 5.5 Cumulative Probability PRAISE Results for the 304-mm (12-in.) and 711-mm (28-in.) 

Recirculation Line Welds (Tregoning et al. 2005) 

304-mm (12–in.) Schedule 80 (a) 711-mm (28–in.) Schedule 80 (b) Leak 

Size 

(gpm) 

Time 

(yr) No Weld Overlay 

Weld Overlay at 

20 Years No Weld Overlay 

25 0.3674 0.2967 6.23 × 10-3

40 0.5986 0.3803 1.02 × 10-2 >0 

60 0.7435 0.4241 1.46 × 10-2

25 0.1682 0.1427 6.0 × 10-4

40 0.2452 0.1622 8.0 × 10-4 >100 

60 0.2872 0.1693 8.0 × 10-4

25 0.1529 0.1066 6.66 × 10-5

40 0.2193 0.1250 6.66 × 10-5 >1500 

60 0.2534 0.1312 1.25 × 10-4

25 -- -- 6.00 × 10-5

40 -- -- 6.87 × 10-5 >5000 5 

60 -- -- 9.79 × 10-5

25 0.1529 0.0490 3.30 × 10-5

40 0.2193 0.0674 3.30 × 10-5 DEGB 

60 0.2535 0.0736 6.70 × 10-5

(a) σNO = 140.8 MPa (20.41 ksi) + Seismic σ ≅ 0 MPa (0 ksi) 

(b) σNO = 65.4 MPa (9.48 ksi)+ Seismic σ = 7.72 MPa (1.12 ksi) 

Figure 5.9 Cumulative Probability of a Leak Exceeding 378.5 lpm (100 gpm) Versus Time for a 

304-mm (12-in.) Recirculation Line Weld With and Without a Weld Overlay at 

20 Years (Tregoning et al. 2005) 
5.11 

Table 5.6 Summary of Service Experience Involving TGSCC 

PlantType Number of Reported Base Metal Failures (1970–2005) 

 All Part Through-Wall Through-Wall 

BWR 209 149 60 

PWR 123 24 99 

BWR + PWR 332 173 159 

• This tabulation is based on failure data as recorded in the proprietary PIPExp database; 

see Appendix D in NUREG-1829 (Tregoning et al. 2005). 

• Maximum recorded leak rate is 6.4 lpm (1.7 gpm). 

The literature survey did not identify any probabilistic fracture mechanics models or probabilistic 

evaluations to determine failure probabilities for TGSCC degradation of the kind observed in the field and 

reported in Table 5.6. Work reported by Lydell (1999) included a statistical assessment of serviceinduced TGSCC to determine the failure probability of cold bent piping. Depending on system and pipe 

size, the derived failure frequencies ranged from about 9.8 × 10-12 to about 1.5 × 10-7 per piping 

component per calendar year. 

5.1.3 Primary Water Stress Corrosion Cracking 

Excluding the service experience with steam generator tubes, incidents of stress corrosion cracking of 

Alloy 600 components including weld alloys 182/82 in domestic PWR primary systems have been 

observed since the early 1980s. This experience is well documented in numerous NRC bulletins, 

information notices, and generic letters (USNRC 1997, 2001a, 2001b, 2002, 2003a, 2003b, 2003c, 2004). 

An excellent history of foreign and domestic Alloy 600 cracking experience is provided in NUREG-1823 

(Grimmel and Cullen 2005). Summarized in Table 5.7 are the numbers of worldwide nuclear power plant 

service failures involving primary water stress corrosion cracking (PWSCC), as reported in the 

proprietary PIPExp database described in Appendix D to NUREG-1829 (Tregoning et al. 2005). 

Research, both domestically and internationally, is being performed to better characterize Alloy 600 

cracking in light water reactors. The Office of Nuclear Regulatory Research programs include studies of 

1) nondestructive inspection technologies, 2) material properties and crack growth rates, and 3) stress and 

structural integrity analysis. Additional work is being undertaken by the industry including the EPRI 

Materials Reliability Program, vendors, and owners’ groups. With regard to structural integrity, the work 

reported in the literature has focused primarily on the need for qualified PWSCC crack initiation, and 

crack growth rate models have been reported by Garud (1997), Garud and Pathania (1999), EPRI (2002, 

2004), White et al. (2003), and Cattant et al. (2005). 

Garud (1997) proposed a simplified strain rate damage model (SRDM) for short PWSCC initiation times 

in low-temperature mil-annealed and cold-worked Alloy 600 tubing materials. In Figure 5.10, Garud 

(1997) compares SRDM predicted Alloy 600 initiation times with available initiation data in the literature 

for various test conditions in high-purity water environments. 
5.12 

Table 5.7 Summary of Service Experience in Piping Involving PWSCC 

Number of Reported Bi-Metallic Weld Failures (1970-2005) 

Pipe Size All Part Through-Wall Through-Wall 

OD ≤ 25.4 mm (≤1-in.) 62 22 40 

25.4 mm (1 in.) < OD ≤ 

254 mm (10-in.) 21 14 7 

OD > 254 mm (10-in.) 2 1 1 

All Pipe Sizes 85 37 48 

• This tabulation is based on failure data as recorded in the proprietary PIPExp database; see 

Appendix D in NUREG-1829 (Tregoning et al. 2005). 

• The data summary is limited to piping components (non-piping passive components are excluded). 

• Maximum recorded through-wall leak rate is 3.1 L/min (0.8 gpm). 

Figure 5.10 Comparisons of SRDM Estimates with All Available Data on Alloy 600 Initiation Life 

for Various Tests and Conditions in High-Purity Water Environments (Garud 1997) 

EPRI recommended crack growth rate (CGR) curves for PWSCC of thick-wall components, such as 

reactor vessel head nozzles including control rod drive mechanism, control element drive mechanism, 

and in-core instrument nozzles, fabricated from Alloy 600 material (EPRI 2002) and Alloys 82 

and182/132 (EPRI 2004), are shown in Figure 5.11 and Figure 5.12. In accordance with standard practice 

for evaluation of SCC, a power-law dependence on stress intensity factor, based upon the relationship 

originally developed by Scott (1991), was assumed. Tabulation of crack growth rates for both Alloy 600 

and Alloy 182 allowed an evaluation showing that SCC growth rates in Alloy 182 weld metal were 

factors of 3–5 faster than rates in Alloy 600. 
5.13 

Figure 5.11 PWSCC Crack Growth Rate versus Crack Stress Intensity for Alloy 600 Laboratory 

Data, EPRI Material Reliability Program Crack Growth Curve, Modified Scott Curve, 

and CGR Data Points for Cook 2 Nozzle 75 (EPRI 2002). MPa√m = 0.91 ksi√in. 

Figure 5.12 Comparisons of MRP-115 Curves for Alloys 182/132 and 82 with MRP 55 

Westinghouse CGR Data for Weld Material Removed from VC Summer Reactor Hot 

Leg Nozzle Dissimilar Metal Weld (EPRI 2004). MPa√m = 0.91 ksi√in. 
5.14 

Probabilistic fracture mechanics calculations were used to estimate PWSCC failure frequencies in the 

864-mm (34-in.) -diameter, 63.5-mm (2.5-in.) -thick PWR hot leg safe end-to-pressure vessel Alloy 600 

weld (Tregoning et al. 2005). The hot leg material is SA-376 austenitic stainless steel. To model the 

initiation and growth of PWSCC cracks, in these analyses, the initiation kinetics were assumed to be the 

same as for Type 316NG stainless steel as currently in PRAISE (Harris et al. 1985); however, the crack 

growth kinetics in PRAISE were changed to be consistent with the EPRI MRP Alloy 600 crack growth 

rate recommendations (Garud 1997). 

The cumulative probabilities for several leak rates and two modes of cracking are shown in Table 5.8. 

First, PWSCC for the growth of an initial fabrication defect with a depth distribution from Khaleel et al. 

(1999) was considered. In the second case, the initiation and growth of multiple PWSCC cracks were 

considered. We can see that, when a single preexisting fabrication crack is postulated, crack growth 

occurs at the start of plant life and all failures occurred within the first 25 years of plant operation. 

Consequently, the failure probability for small leaks is significantly higher than the larger leak sizes. In 

the second case, a significant operating time is required for PWSCC initiation and the crack growth phase 

is much shorter. Consequently, the small leak probability is much lower. Also, because multiple cracks 

tend to coalesce and grow as longer cracks, the probability of larger leaks is higher than when only a 

single fabrication crack is present. 

As part of an ongoing NRC study, PWSCC probabilistic failure probability estimates for a PWR hot leg 

bi-metallic Alloy 600 weld have been performed using the PRO-LOCA code (Rudland et al., NRC, 

unpublished(a)). Initial results are expected to be published in the summer of 2006. 

5.2 Fatigue 

This section discusses two categories of fatigue degradation–thermal and vibration fatigue. Although 

both forms of fatigue degradation are well documented in the literature, most of the publications in the 

literature search were related to mechanistic aspects for initiation and crack growth modeling. The 

limited number of probabilistic fracture mechanics applications reported for nuclear plant components 

were associated mostly with thermal fatigue mechanisms. 

5.2.1 Thermal Fatigue 

Thermal fatigue has caused through-wall cracking and leakage in BWR and PWR plants from both lowcycle and high-cycle thermal fatigue. The service experience with thermal fatigue is extensively 

documented in reports by the Nuclear Energy Agency (1998) and Shah et al. (1997), as well as numerous 

NRC bulletins and information notices (USNRC 1979, 1988a, 1988b, 1988c, 1988d, 1989). Included in 

Table 5.9 is a summary of the thermal fatigue failures recorded in the proprietary PIPExp database as 

described in Appendix D of NUREG-1829 (Tregoning et al. 2005). 

Experience (EPRI 1994, 2000; USNRC 1992) suggests that service-induced fatigue in operating plants 

are caused primarily by thermal stratification or hot-cold water mixing conditions, such as thermal 

 

(a) Rudland DL, H Xu, G Wilkowski, N Ghadiali, F Brust, and P Scott. Unpublished. Evaluation of 

Loss-of-Coolant Accident (LOCA) Frequencies Using the PRO-LOCA Code, NUREG/CR-XXXX, 

currently under review. 
5.15 

Table 5.8 Cumulative PRAISE Failure Probabilities for Hot Leg Pressure Vessel-to-Safe End 

Alloy 600 Weld (Tregoning et al. 2005) 

Leak Size Time 

PWSCC Growth of Single 

Fabrication Defect 

PWSCC Initiation and 

Growth of Multiple Cracks 

25 0.916 0.001 

40 0.918 0.020 >0 lpm 

(>0 gpm) 60 0.919 0.068 

25 2.167 × 10-7 1 × 10-5

40 2.167 × 10-7 2.69 × 10 >378.5 lpm -4

(>100 gpm) 60 2.167 × 10-7 1.78 × 10-3

25 2.78 × 10-11 <1 × 10-4

40 2.78 × 10-11 1 × 10 >5677 lpm -4

(>1500 gpm) 60 2.78 × 10-11 4.85 × 10-4

25 4.66 × 10-11 <1 × 10-5

40 4.66 × 10-11 9 × 10 >18926 lpm -5

(>5000 gpm) 60 4.66 × 10-11 3.77 × 10-4

25 2.59 × 10-13 <1 × 10-5

40 2.59 × 10-13 9 × 10-5 DEGB 

60 2.59 × 10-13 3.77 × 10-5

Table 5.9 Summary of Service Experience Involving Thermal Fatigue 

(Tregoning et al. 2005)

Number of Pipe Failures Attributed to Thermal Fatigue (1970–2005) 

Plant 

Type 

All Failure 

Types 

Part ThroughWall Crack Small ThroughWall Leak Large ThroughWall Leak 

BWR 69 40 27 2 

PWR 87 42 40 5 

All 156 82 67 7 

striping, and cyclic turbulent mixing not analyzed in the original plant designs. These loading conditions 

are known to impose many cycles, with the cyclic boundary conditions changing rapidly causing cracks to 

initiate and grow at multiple locations on the inside surface of the pipe. Because of the large nonlinear 

gradient stress profiles associated with these service conditions, cracks will tend to grow in the length 

direction where the highest surface cyclic stresses occur. This will cause many of these cracks to coalesce 

and grow as effectively longer defects. Another large body of fatigue cracking events relates to PWR 

piping (USNRC 1990a) and BWR feedwater and control rod drive return line nozzles (Snaider 1980; 

USNRC 1980). In both situations, relatively colder water flows into a nozzle of a hot vessel. For the 

PWRs, the cracking initiated at the weld between the nozzle/safe-end and the attached piping. In the 

BWRs, the cracking occurred very locally in the vessel nozzle and was due to localized mixing of the hot 

and cold water on the nozzle bore and blend-radius regions. Again, these loadings were unknown at the 

time of the original plant design. 
5.16 

In NUREG/CR-2189, Harris et al. (1981) documented the probabilistic fracture mechanics calculations 

performed with the early version of the PRAISE code to estimate the probability of a seismic-induced 

LOCA in the primary piping of a commercial pressurized water reactor. The primary piping system at 

Zion 1 was selected for these analyses. Attention was focused on the girth butt welds in the hot leg, cold 

leg, and cross-over piping. The primary piping was austenitic stainless steel, predominantly centrifugally 

cast stainless steel. Zion 1 was a Westinghouse 1100-MWe four-loop PWR design. A listing of the 

11 Zion 1 primary system design thermal transients considered in these computations along with their 

postulated number of occurrences over the 40-year plant life is shown in Table 5.10. Appendix D of 

NUREG/CR-2189 included the design time-temperature profiles in the Zion 1 hot leg, cold leg, and steam 

lines for each of the design transients in Table 5.10. The Zion 1 primary system piping, with the 14 weld 

locations considered and the resulting failure rates (i.e., leaks per weld per year), is shown in Figure 5.13. 

Figure 5.14 provides an example of the conditional LOCA probabilities for weld joints 1 and 13 for 

various magnitudes of seismic events. 

NUREG/CR-6674 (Khaleel et al. 2000) documents calculations performed with the pc-PRAISE code to 

address fatigue failures for 47 selected components for seven representative plants covering both BWR 

and PWR plants. The calculations also covered all four nuclear steam supply system suppliers and both 

older and newer vintage plants. These calculations predicted probabilities of crack initiation and 

probabilities of through-wall cracks as a function of time for plant operating periods up to 60 years. The 

calculations were made possible by the development of a new version of the pc-PRAISE code that has the 

capability to simulate the initiation of fatigue cracks in combination with a simulation of the subsequent 

growth of these fatigue cracks. Probabilities of crack initiation were calculated using fatigue-life 

correlations from the Argonne National Laboratory (ANL) research for NRC (Keisler et al. 1995). The 

number of cycles to crack initiation in the Argonne equations is a function of the material type, water/air 

Table 5.10 Zion 1 Primary System Piping Design Transients and Number of Postulated 

Occurrences in 40-year Design Life (Harris et al. 1981) 

Design Transient Design Cycles 

Heatup and Cooldown at 37.7°C/hr (100°F/hr) 200 (each) 

Unit Loading and Unloading at 5% of full 

power/minute 18,300 (each) 

Step Load Increase and Decrease at 10% of full power 2,000 (each) 

Large Step Load Decrease 200 

Loss of Load 80 

Loss of Power 40 

Loss of Flow (partial loss) 80 

Reactor Trip from Full Power 400 

Steam Line Break 1 

Turbine Roll Test 10 

Hydrostatic Test Condition 5 
5.17 

Figure 5.13 Zion 1 Primary System Piping Weld Locations and NUREG/CR-2189 Estimated Leak 

Frequencies (Heasler and Simonen 1990) 

environment, temperature, dissolved oxygen content, sulfur content, and strain rate. The alternating 

stresses and anticipated number of cycles were the same as those used for the deterministic calculations of 

fatigue usage factors of NUREG/CR-6260 (Ware et al. 1995). Detailed inputs and results for the 

probabilistic fracture mechanics calculations are given in NUREG/CR-6674. 

Tables 5.11 and 5.12 provide a summary of the results for all the components, with Table 5.11 accounting 

for the effects of reactor water environments and Table 5.12 neglecting the environmental effects. 

Results are given for both a 40-year and a 60-year operating period. Many of the components analyzed in 

NUREG/CR-6674 had cumulative probabilities of crack initiation(a) and cumulative probabilities of 

through-wall cracks that approach unity within the 40-year to 60-year time period. Other components, 

often with similar values of fatigue usage factors, show much lower failure probabilities. The maximum 

failure rate (through-wall cracks per year) is about 5 × 10-2. Tables 5.11 and 5.12 also show estimated 

values for core-damage frequencies, which show maximum values based on the calculated failure rates of 

about 1.0 × 10-6 per year. These maximum values correspond to components with very high cumulative 

failure probabilities, and the failure rates do not change significantly from 40 years to 60 years. Failure 

 

(a) The ANL S-N model associates fatigue life with approximately a 3-mm (0.020-in.) surface flaw depth 

(Gavenda et al. 1997; Chopra and Shack 2001; Chopra 2002) and is based on consideration of the 

25% load drop criteria used to denote fatigue failure of smooth push-pull specimens in fatigue testing 

procedures. 
5.18 

Figure 5.14 NUREG/CR-2189 Conditional LOCA Probabilities as a Function of Time for Two 

Zion I Representative Weld Locations Showing the Influence of Seismic Events 

(Harris et al. 1981) 
5.19 

Table 5.11 Summary of Results for All Seven Plants - Water Environment (Khaleel et al. 2000

DET DET CUMULATIVE CUMULATIVE CUMULATIVE CUMULATIVE TWC/YEAR TWC

USEAGE@40 USEAGE@60 PI@40 YR PI@60 YR PTWC@40 YR PTWC@60 YR @40 YR @6

CE-NEW RPV LOWER HEAD/SHELL LAS 1.40E-02 2.10E-02 7.89E-06 4.82E-05 6.71E-15 1.44E-12 1.13E-15 1.90CE-NEW RPV INLET NOZZLE LAS 4.75E-01 7.12E-01 1.40E-02 4.44E-02 5.90E-05 9.01E-04 7.50E-06 7.59

CE-NEW RPV OUTLET NOZZLE LAS 4.72E-01 7.08E-01 4.22E-01 6.89E-01 1.74E-03 2.90E-02 3.58E-04 2.57

CE-NEW SURGE LINE ELBOW 304/316 2.60E+00 3.90E+00 9.95E-01 9.99E-01 9.81E-01 9.98E-01 7.60E-02 9.38

CE-NEW CHARGING NOZZLE NOZZLE LAS 1.04E-01 1.56E-01 9.56E-04 3.84E-03 2.61E-06 5.50E-05 3.46E-07 5.06

CE-NEW CHARGING NOZZLE SAFE END 304/316 5.02E-01 7.53E-01 1.06E-02 6.75E-02 9.00E-05 1.03E-03 1.75E-05 1.15

CE-NEW SAFETY INJECTION NOZZLE NOZZLE LAS 4.57E-01 6.85E-01 1.01E-03 4.81E-03 1.00E-06 1.90E-05 3.75E-07 1.50

CE-NEW SAFETY INJECTION NOZZLE SAFE E 304/316 2.86E-01 4.29E-01 8.68E-03 3.16E-02 2.61E-06 5.50E-05 3.46E-07 5.06

CE-NEW SHUTDOWN COOLING LINE ELBOW 304/316 4.87E-01 7.30E-01 1.13E-02 5.75E-02 2.00E-05 4.53E-04 7.00E-06 4.40

CE-OLD RPV LOWER HEAD/SHELL LAS 1.30E-02 1.95E-02 2.68E-06 1.93E-05 6.36E-16 1.85E-13 1.07E-16 1.85

CE-OLD RPV INLET NOZZLE LAS 1.72E-01 2.58E-01 1.88E-03 7.89E-03 4.11E-07 1.33E-05 5.87E-08 1.33

CE-OLD RPV OUTLET NOZZLE LAS 5.53E-01 8.29E-01 5.91E-01 8.46E-01 7.05E-02 3.53E-01 8.98E-03 2.27

CE-OLD SURGE LINE ELBOW 304/316 6.61E-01 9.92E-01 9.39E-01 9.87E-01 6.27E-01 8.85E-01 4.36E-02 5.48

CE-OLD CHARGING NOZZLE SAFE END 304/316 5.62E-01 8.43E-01 1.18E-02 5.31E-02 4.10E-05 5.98E-04 8.75E-06 5.05CE-OLD SAFETY INJECTION NOZZLE SAFE E 304/316 3.17E-01 4.75E-01 7.56E-03 3.59E-02 1.40E-05 2.00E-04 2.25E-06 1.85CE-OLD SHUTDOWN COOLING LINE ELBOW 304/316 8.40E-02 1.26E-01 3.94E-02 1.19E-01 2.10E-04 2.36E-03 4.38E-05 1.98B&W RPV NEAR SUPPORT SKIRT LAS 2.23E-01 3.35E-01 8.25E-03 2.50E-02 7.85E-06 1.52E-04 1.04E-06 1.36

B&W RPV OUTLET NOZZLE LAS 4.69E-01 7.04E-01 7.74E-01 8.99E-01 1.83E-01 5.44E-01 1.94E-02 3.35

B&W MAKEUP/HPI NOZZLE SAFE END 304/316 1.05E+00 1.58E+00 1.30E-01 4.79E-01 2.10E-03 3.09E-02 5.88E-04 2.22

B&W DECAY HEAT REMOVAL/REDUCING T 304/316 5.30E-01 7.95E-01 5.72E-02 2.08E-01 3.00E-03 2.54E-02 4.26E-04 1.79

W-NEW RPV LOWER HEAD/SHELL LAS 1.80E-02 2.70E-02 3.21E-05 1.71E-04 7.52E-13 9.64E-11 1.23E-13 1.21

W-NEW RPV INLET NOZZLE LAS 2.90E-01 4.35E-01 2.49E-03 1.05E-02 9.17E-07 2.84E-05 1.30E-07 2.83

W-NEW RPV OUTLET NOZZLE LAS 6.58E-01 9.87E-01 8.62E-01 9.49E-01 3.65E-01 7.42E-01 3.17E-02 4.50

W-NEW CHARGING NOZZLE NOZZLE 316NG 3.37E+00 5.06E+00 9.51E-01 9.83E-01 8.72E-01 9.63E-01 5.38E-02 5.66

W-NEW SAFETY INJEC NOZZLE NOZZLE 316NG 1.46E+00 2.19E+00 4.34E-03 3.69E-02 5.00E-04 1.09E-02 5.33E-05 1.30

W-NEW RESIDUAL HEAT INLET TRAN 304/316 2.73E+00 4.10E+00 9.58E-01 9.99E-01 7.80E-01 9.80E-01 6.25E-02 1.18

W-OLD RPV LOWER HEAD SHELL LAS 8.91E-01 1.34E+00 1.11E-01 1.28E-01 7.20E-07 1.11E-05 8.38E-08 9.08

W-OLD RPV INLET NOZZLE INNER SURFACE LAS 3.02E-01 4.53E-01 3.91E-01 6.44E-01 4.38E-03 5.04E-02 7.53E-04 3.96

W-OLD RPV INLET NOZZLE OUTER SURFACE LAS 4.96E-01 7.44E-01 6.81E-02 1.11E-01 4.48E-04 3.32E-03 4.75E-05 2.18

W-OLD RPV OUTLET NOZZLE INNER SURF LAS 4.99E-01 7.48E-01 4.90E-01 7.53E-01 9.33E-03 9.60E-02 1.56E-03 7.54

W-OLD RPV OUTLET NOZZLE OUTER SURF LAS 3.47E-01 5.20E-01 1.63E-01 2.38E-01 7.77E-03 3.60E-02 6.99E-04 1.83

W-OLD CHARGING NOZZLE NOZZLE 304/316 3.19E-01 4.79E-01 4.67E-04 3.75E-03 3.00E-07 5.20E-06 7.50E-08 6.00W-OLD SAFETY INJECTION NOZZLE NOZZLE 304/316 3.27E-01 4.90E-01 1.88E-03 1.31E-02 4.00E-06 8.80E-05 8.75E-07 1.05

W-OLD RESIDUAL HEAT REMOVAL TEE 304/316 2.05E-01 3.08E-01 1.34E-02 5.16E-02 1.15E-04 1.14E-03 1.63E-05 9.26GE-NEW RPV NEAR CRDM PENETRATION LAS 6.28E-01 9.42E-01 7.89E-05 3.49E-04 7.88E-12 6.82E-10 1.25E-12 8.26

GE-NEW FEEDWATER NOZZLE SAFE END LAS 1.88E+00 2.82E+00 1.04E-01 2.53E-01 1.31E-03 1.47E-02 2.38E-04 1.23

GE-NEW RECIRC SYS - TEE SUCTION PIPE 304/316 8.30E-01 1.25E+00 4.23E-02 1.39E-01 4.80E-04 4.67E-03 7.13E-05 3.66

GE-NEW CORE SPRAY LINE SAFE END EXT LAS 4.36E-01 6.54E-01 3.83E-04 1.27E-03 1.45E-07 3.25E-06 1.97E-08 3.04

GE-NEW RHR LINE STRAIGHT PIPE LAS 1.13E+01 1.69E+01 4.73E-01 6.71E-01 4.10E-01 6.21E-01 1.35E-02 2.25

GE-NEW FEEDWATER LINE ELBOW LAS 3.69E+00 5.53E+00 1.59E-01 3.65E-01 1.01E-03 1.46E-02 1.69E-04 1.35GE-OLD RPV LOWER HEAD TO SHELL LAS 7.90E-02 1.19E-01 2.71E-10 2.76E-08 0.00E+00 0.00E+00 0.00E+00 0.00

GE-OLD RPV FEEDWATER NOZZLE BORE LAS 3.17E+00 4.75E+00 7.27E-02 2.42E-01 1.00E-05 8.80E-04 2.50E-06 9.76

GE-OLD RECIRC SYSTEM RHR RETURN LINE 304/316 3.90E+00 5.85E+00 9.43E-01 9.99E-01 7.12E-01 9.85E-01 7.20E-02 1.23

GE-OLD CORE SPRAY SYSTEM NOZZLE LAS 5.20E-01 7.80E-01 1.41E-04 7.89E-04 1.91E-08 8.84E-07 2.85E-09 9.51

GE-OLD CORE SPRAY SYSTEM SAFE END 304/316 1.77E+00 2.65E+00 3.33E-01 7.64E-01 1.46E-02 1.10E-01 2.08E-03 8.04

GE-OLD RESIDUAL HEAT TAPERED 304/316 4.78E-01 7.17E-01 1.47E-03 7.89E-03 9.21E-05 1.02E-03 1.07E-05 7.82

GE-OLD FEEDWATER LINE - RCIC TEE LAS 6.98E+00 1.05E+01 3.86E-01 7.82E-01 2.99E-03 5.92E-02 6.96E-04 5.54

PLANT COMPONENT MAT
5.20 

Table 5.12 Summary of Results for All Seven Plants - Air Environment (Khaleel et al. 2000) 

ASME DESIGN ASME DESIGN CUMULATIVE CUMULATIVE CUMULATIVE CUMULATIVE TWC/YEAR TWC/Y

USEAGE@40 USEAGE@60 PI@40 YR PI@60 YR PTWC@40 YR PTWC@60 YR @40 YR @60 

CE-NEW RPV LOWER HEAD/SHELL LAS 7.00E-03 1.05E-02 1.30E-06 1.00E-05 8.40E-23 3.96E-19 1.64E-23 6.37ECE-NEW RPV INLET NOZZLE LAS 1.82E-01 2.73E-01 4.24E-03 1.64E-02 1.75E-07 7.64E-06 2.57E-08 8.18ECE-NEW RPV OUTLET NOZZLE LAS 3.34E-01 5.01E-01 2.21E-01 4.55E-01 1.00E-07 7.70E-06 2.50E-08 1.02ECE-NEW SURGE LINE ELBOW 304/316 9.81E-01 1.47E+00 6.48E-01 8.26E-01 4.52E-01 7.11E-01 2.69E-02 5.16ECE-NEW CHARGING NOZZLE NOZZLE LAS 5.00E-02 7.50E-02 2.27E-04 1.16E-03 6.48E-11 5.96E-09 1.04E-11 7.22ECE-NEW CHARGING NOZZLE SAFE END 304/316 7.78E-01 1.17E+00 3.83E-04 3.16E-03 1.62E-04 1.76E-03 1.90E-05 1.33ECE-NEW SAFETY INJECTION NOZZLE NOZZLE LAS 8.98E-01 1.35E+00 2.15E-03 8.69E-03 1.22E-06 3.18E-05 1.66E-07 3.09ECE-NEW SAFETY INJECTION NOZZLE SAFE E 304/316 3.60E-01 5.40E-01 2.12E-05 1.64E-04 1.04E-06 1.59E-05 1.29E-07 1.37ECE-NEW SHUTDOWN COOLING LINE ELBOW 304/316 8.94E-01 1.34E+00 1.56E-04 1.00E-03 1.83E-05 2.01E-04 2.10E-06 1.55ECE-OLD RPV LOWER HEAD/SHELL LAS 8.00E-03 1.20E-02 4.03E-07 3.47E-06 4.85E-24 3.80E-20 9.51E-25 5.95E

CE-OLD RPV INLET NOZZLE LAS 7.30E-02 1.10E-01 4.44E-04 2.27E-03 7.99E-11 7.84E-09 1.26E-11 9.69E

CE-OLD RPV OUTLET NOZZLE LAS 2.84E-01 4.26E-01 7.89E-02 1.28E-01 6.72E-04 4.79E-03 7.07E-05 3.09ECE-OLD SURGE LINE ELBOW 304/316 7.05E-01 1.06E+00 2.62E-01 4.91E-01 1.74E-02 9.10E-02 2.02E-03 5.46ECE-OLD CHARGING NOZZLE SAFE END 304/316 2.66E-01 3.99E-01 5.62E-04 3.16E-03 9.92E-05 8.92E-04 1.09E-05 6.45ECE-OLD SAFETY INJECTION NOZZLE SAFE E 304/316 8.80E-02 1.32E-01 2.39E-04 1.34E-03 2.03E-05 2.18E-04 2.36E-06 1.65ECE-OLD SHUTDOWN COOLING LINE ELBOW 304/316 - - 8.69E-04 3.16E-03 5.13E-05 3.78E-04 5.28E-06 2.53E

B&W RPV NEAR SUPPORT SKIRT LAS 3.60E-01 5.40E-01 2.61E-03 9.56E-03 4.07E-09 2.64E-07 6.23E-10 3.01EB&W RPV OUTLET NOZZLE LAS 1.43E-01 2.15E-01 7.89E-02 1.05E-01 2.92E-03 1.27E-02 2.49E-04 6.50EB&W MAKEUP/HPI NOZZLE SAFE END 304/316 7.40E-01 1.11E+00 3.03E-03 1.98E-02 1.91E-03 1.48E-02 2.07E-04 1.02EB&W DECAY HEAT REMOVAL/REDUCING T 304/316 - - 1.05E-03 4.89E-03 1.70E-04 1.34E-03 1.80E-05 9.13EW-NEW RPV LOWER HEAD/SHELL LAS 1.00E-02 1.50E-02 6.19E-06 4.10E-05 1.04E-19 8.98E-17 1.79E-20 1.39EW-NEW RPV INLET NOZZLE LAS 1.10E-01 1.65E-01 5.62E-04 3.03E-03 7.03E-10 6.54E-08 1.11E-10 7.93EW-NEW RPV OUTLET NOZZLE LAS 3.98E-01 5.97E-01 7.26E-01 8.76E-01 3.00E-04 2.20E-03 2.50E-05 2.00EW-NEW CHARGING NOZZLE NOZZLE 316NG 8.29E-01 1.24E+00 6.18E-01 8.11E-01 3.17E-01 6.06E-01 2.18E-02 2.75EW-NEW SAFETY INJEC NOZZLE NOZZLE 316NG 9.66E-01 1.45E+00 2.89E-04 2.61E-03 1.93E-04 2.06E-03 2.25E-05 1.50EW-NEW RESIDUAL HEAT INLET TRAN 304/316 8.96E-01 1.34E+00 1.66E-01 3.79E-01 2.90E-02 1.32E-01 3.09E-03 8.11EW-OLD RPV LOWER HEAD SHELL LAS 2.90E-01 4.35E-01 9.56E-02 1.11E-01 1.06E-09 3.34E-08 1.37E-10 3.27EW-OLD RPV INLET NOZZLE INNER SURFACE LAS 1.35E-01 2.03E-01 2.06E-01 4.21E-01 2.00E-07 1.03E-05 1.00E-07 1.15EW-OLD RPV INLET NOZZLE OUTER SURFACE LAS 2.08E-01 3.12E-01 6.82E-02 1.11E-01 4.48E-04 3.33E-03 4.75E-05 2.19EW-OLD RPV OUTLET NOZZLE INNER SURF LAS 1.93E-01 2.90E-01 2.67E-01 5.22E-01 2.00E-05 6.00E-05 1.25E-06 2.50EW-OLD RPV OUTLET NOZZLE OUTER SURF LAS 4.31E-01 6.47E-01 1.63E-01 2.39E-01 7.77E-03 3.60E-02 7.00E-04 1.83EW-OLD CHARGING NOZZLE NOZZLE 304/316 4.60E-01 6.90E-01 1.53E-05 1.48E-04 8.90E-07 1.68E-05 1.15E-07 1.52EW-OLD SAFETY INJECTION NOZZLE NOZZLE 304/316 1.99E+00 2.99E+00 4.10E-05 3.67E-04 4.77E-06 7.46E-05 5.95E-07 6.40EW-OLD RESIDUAL HEAT REMOVAL TEE 304/316 2.20E-02 3.30E-02 4.89E-04 2.27E-03 3.01E-05 2.86E-04 3.38E-06 2.07EGE-NEW RPV NEAR CRDM PENETRATION LAS 4.07E-01 6.11E-01 1.94E-05 1.00E-04 3.59E-18 1.55E-15 6.03E-19 2.18E

GE-NEW FEEDWATER NOZZLE SAFE END LAS 3.01E-01 4.52E-01 4.31E-02 1.19E-01 2.00E-06 2.00E-05 1.25E-07 1.75EGE-NEW RECIRC SYS - TEE SUCTION PIPE 304/316 2.98E-01 4.47E-01 2.61E-03 1.00E-02 6.50E-04 3.59E-03 5.96E-05 2.16EGE-NEW CORE SPRAY LINE SAFE END EXT LAS 5.00E-02 7.50E-02 1.22E-04 4.65E-04 6.45E-12 5.21E-10 1.02E-12 6.26EGE-NEW RHR LINE STRAIGHT PIPE LAS 4.07E-01 6.11E-01 3.71E-01 4.99E-01 2.08E-01 3.02E-01 7.21E-03 6.09E

GE-NEW FEEDWATER LINE ELBOW LAS 4.35E-01 6.53E-01 6.77E-02 1.82E-01 1.00E-05 6.00E-05 1.25E-06 5.00EGE-OLD RPV LOWER HEAD TO SHELL LAS 3.20E-02 4.80E-02 0.00E+00 4.72E-10 0.00E+00 0.00E+00 0.00E+00 0.00E+GE-OLD RPV FEEDWATER NOZZLE BORE LAS 7.58E-01 1.14E+00 3.83E-02 9.13E-02 4.53E-05 6.76E-04 5.79E-06 5.58E

GE-OLD RECIRC SYSTEM RHR RETURN LINE 304/316 3.97E-01 5.96E-01 1.11E-01 4.18E-01 1.44E-02 1.33E-01 2.38E-03 1.04E

GE-OLD CORE SPRAY SYSTEM NOZZLE LAS 2.30E-02 3.45E-02 2.50E-05 1.79E-04 4.44E-14 7.67E-12 7.63E-15 1.01EGE-OLD CORE SPRAY SYSTEM SAFE END 304/316 1.82E-01 2.73E-01 6.11E-03 4.30E-02 3.40E-06 1.08E-04 7.12E-07 1.19EGE-OLD RESIDUAL HEAT TAPERED 304/316 1.56E+00 2.34E+00 6.19E-06 7.19E-05 2.62E-07 6.36E-06 3.56E-08 6.10EGE-OLD FEEDWATER LINE - RCIC TEE LAS 5.84E-01 8.76E-01 1.37E-01 4.35E-01 4.30E-06 2.06E-04 1.33E-06 2.29E

PLANT COMPONENT MAT
5.21 

rates for other components having much lower probabilities of failure are seen to increase by as much as 

an order of magnitude from 40 years to 60 years, but these components make relatively small overall 

contributions to core damage frequencies. 

The pc-PRAISE code was applied in the calculations, with crack initiation depending only on the peak 

local stresses. Crack initiation was modeled as occurring at multiple sites with the pc-PRAISE model 

using a 5-cm (2-in.) -long initiation site. The Monte Carlo calculations with the pc-PRAISE code were 

run to a maximum of 106

 simulations. Because some components had very low failure probabilities, this 

number of simulations was sometimes inadequate to establish probabilities of through-wall cracks. 

Rather than reporting a probability as less than 10-6, additional calculations were performed with a Latin 

Hypercube approach developed by PNNL on a previous NRC research project (Simonen and Khaleel 

1998b). The italicized values in Table 5.11 were derived from these supplementary calculations. The 

importance sampling procedure of the Latin Hypercube code permitted calculations of through-wall crack 

probabilities as small as 10-15. 

The pc-PRAISE calculations were based on a number of conservative modeling assumptions and input 

parameters, balanced by other non-conservative assumptions and inputs. In the balance, the calculations 

are believed to provide realistic predictions of through-wall crack frequencies for the assumed cyclic 

stress pairs. The inputs for stress cycles as taken from NUREG/CR-6260 (Ware et al. 1995) were 

believed in most cases to conservatively bound the stresses experienced during actual plant operation. 

These stresses were taken from design stress reports that assumed bounding conditions for thermal stress 

transients and other loads. In addition, the method used to derive load pairs from the transient 

descriptions assumed worst-case sequencing of loads. The method used in the current calculations to 

estimate through-wall stress distributions (uniform tension versus through-wall gradient) were intended to 

overestimate the fraction of the stress assigned to the uniform tension category. Inputs for strain rates, 

oxygen, and sulfur were all assigned as bounding values that are unlikely to be present simultaneously for 

any given component. 

In summary, the calculations of NUREG/CR-6674 indicated that the components with the very high 

probabilities of failure can have through-wall crack frequencies that approach about 5 × 10-2 per year. In 

contrast, other components with much lower failure probabilities can have their failure frequencies 

increase by factors of about 10 from 40 years to 60 years. Calculations were also performed to address 

the effects of reactor water environments (versus air) and to compare these effects to the effects of 

extended plant operation from 40 years to 60 years. The environmental effects were predicted to increase 

through-wall crack probabilities by as much as two orders of magnitude. The calculated failure 

frequencies of NUREG/CR-6674 were sufficiently low to make relatively small contributions of fatigue 

failures to core-damage frequencies for the seven representative plants addressed by the study. However, 

it was also observed that the calculated failure frequencies were higher than the corresponding 

frequencies based on reported events at operating plants. As discussed above, the differences can be 

explained in terms of conservative inputs for cyclic stresses and assumptions regarding plant operating 

environments. 

In a more recent study (Tregoning et al. 2005), probabilistic fracture mechanics models in pc-PRAISE 

were applied to estimate LOCA frequencies as a function of break size and operating time for a PWR 

stainless steel pressurizer surge line, a stainless steel high-pressure injection nozzle safe-end weld, and a 
5.22 

BWR carbon steel feedwater line elbow. In all cases, the dominant failure mode was assumed to be 

thermal fatigue. Examples of the cumulative probability results for each line are summarized in 

Table 5.13. 

Table 5.13 NUREG-1829 Appendix F Cumulative Probabilities for PWR Surge Line Elbow, PWR 

HPI Nozzle Safe End (with and without thermal sleeve) and BWR Feedwater Line 

(Tregoning et al. 2005) 

Leak Size 

(gpm) Time 

PWR Surge 

Line Elbow 

PWR HPI Nozzle 

Safe End 

(thermal sleeve) 

PWR HPI Nozzle 

Safe End 

(no thermal sleeve) 

BWR 

Feedwater Line 

5 -- -- 5.67 × 10-5 -- 

25 0.216 1.01 × 10-5 3.69 × 10-3 <1.00 × 10-8

40 0.593 6.80 × 10-4 1.26 × 10-2 5.69 × 10-6

>0 L/min 

(>0 gpm) 

60 0.889 1.04 × 10-2 2.98 × 10-2 2.57 × 10-4

25 1.13 × 10-4 4.50 × 10-8 6.49 × 10-4 -- 

40 4.00 × 10-4 4.90 × 10-7 2.68 × 10-3 1.03 × 10

>378.5 L/min -11

(>100 gpm) 60 8.02 × 10-4 1.79 × 10-5 5.31 × 10-3 6.44 × 10-7

25 3.41 × 10-8 2.00 × 10-8 -- -- 

40 3.56 × 10-7 2.10 × 10-7 -- 5.29 × 10

>5677 L/min -21

(>1500 gpm) 60 1.52 × 10-6 4.56 × 10-6 -- 7.84 × 10-11

25 2.42 × 10-9 -- -- 

40 3.73 × 10-8 -- 3.88 × 10 >18926 L/min -24

(>5000 gpm) 60 2.04 × 10-7 -- 3.74 × 10-12

25 9.86 × 10-13 -- 6.49 × 10-4 -- 

40 4.86 × 10-11 -- 2.68 × 10-3 2.44 × 10-35 DEGB 

60 5.43 × 10-10 -- 5.31 × 10-3 7.14 × 10-17

5.2.2 Vibration Fatigue 

Vibration fatigue failures are normally a result of poor piping design or installation and welding practices. 

A relatively large share of vibration fatigue failures initiates at the fillets of socket and support attachment 

welds due to a high stress concentration at the juncture of the weld and base metal. Small-bore pipe 

socket welded vent and drain connections less than 25.4 mm (<1.0 in.) in the immediate proximity of 

vibration sources tend to be most susceptible to this failure mechanism (Olson 1985; EPRI 1994; 

Riccardella et al. 1997; Shah et al. 1998). 

Unlike the previously discussed mechanisms, vibration fatigue may not always lend itself to periodic 

inservice examinations (i.e., volumetric, surface) as a means of managing this degradation mechanism. 

This is especially true if the inspections occur at the normal inservice inspection frequency of once every 

10 years. The nature of this mechanism is such that, generally, almost the entire fatigue life of the 

component is expended during the initiation phase. Once a crack initiates, failure quickly follows. 

Therefore, the absence of any detectable crack may not ensure reliable component performance. In 

addition, for many of these components, the plant conditions when vibration levels are unacceptable may 
5.23 

be very difficult to predict since they may be limited to short time periods associated with unique plant or 

system configurations. This would explain why we continue to observe cases where failures occur late in 

the plant’s operating life. Therefore, the fact that a vibration fatigue failure has not occurred within the 

first few years of plant operation does not preclude future failures. 

A review of domestic fatigue failures in U.S. nuclear power plants (EPRI 1994) indicated that although a 

few well-known thermal fatigue failures had affected numerous plants, by far the majority of all fatigue 

failures (approximately 80%) were a result of high-cycle vibration fatigue. Simonen and Gosselin (2001) 

report that vibration fatigue accounts for approximately 30% of all reported piping failures in nuclear 

power plant systems. EPRI (1994) and Shah et al. (1998) report that industry service experience (reported 

failure events, industry surveys, and discussions with plant maintenance personnel) suggest that, on 

average, small-bore vibration failures occur as often as 0.2–0.3 times per plant year. 

Figure 5.15 provides a perspective on the service experience with socket weld failures in safety-related 

piping as reported in the proprietary PIPExp database described in Appendix D of NUREG-1829 

(Tregoning et al. 2005). For the 30-year period 1976–2006, this chart displays the annual number of 

socket weld failures as a percentage of the total number of pipe or weld failures in safety-related systems. 

It includes U.S. and foreign experience. Only piping of nominal diameter 102 mm (4 in.) or less are 

accounted for in this service experience summary. The chart does not convey information about general 

failure trends. There is significant plant-to-plant variability with respect to vibration fatigue 

susceptibilities. Also noteworthy is the observation that some national design practices and standards no 

longer allow the use of socket welds. Table 5.14 is a summary of vibration fatigue service experience 

involving failures of butt welds and socket welds. 

1.0%

10.0%

100.0%

1976197719781979198019811982198319841985198619871988198919901991199219931994199519961997199819992000200120022003200420052006

Socket Weld Failures as % of All Pipe

Failures in 4-inch or Less Safety-Related Piping

550 recorded socket weld failures for period

1976 through 1st quarter 2006 - all degradation mechanisms

Figure 5.15 Socket Weld Failure as Percentage of All Failures of Safety Related Piping of Nominal 

Pipe Size ≤102-mm (≤4-in.) (Tregoning et al. 2005)
5.24 

Table 5.14 Service Experience Involving Butt Weld and Socket Weld Failure Due to Vibration 

Fatigue

Recorded Weld Failures Due to Vibration-Fatigue 

Weld Type Pipe Size All Part Through-Wall Leak Severance 

Butt Weld/ ≤25.4 mm (≤ 1 in.) 622 50 508 64 

Attachment Weld >25.4 mm (> 1 in.) 205 23 172 10 

≤25.4 mm (≤ 1 in.) 290 28 253 9 Socket Weld 

>25.4 mm (> 1 in.) 116 48 68 0 

This tabulation is based on failure data as recorded in the proprietary PIPExp database; see Appendix D in 

NUREG-1829 (Tregoning et al. 2005). 

The current study could not identify any probabilistic fracture mechanics models that can address the 

types of vibration fatigue failures as observed in the field. The field failures have largely been associated 

with socket welds and from vibration stresses. Major uncertainties relate to when and where the 

vibrations occur and the stress amplitudes when the vibrations do occur, which are factors not addressed 

by any of the probabilistic fracture models discussed in this report. There are some limited-scope 

treatments of vibration stresses that have been part of the other probabilistic models designed to 

specifically address other failure mechanisms such as fatigue and intergranular stress corrosion cracking 

(e.g., PRAISE and SRRA computer codes). Pipe failure probabilities reflective of high-cycle vibration 

fatigue can quite readily be obtained through application of service experience data in a statistical analysis 

framework. 

5.3 Flow-Assisted Degradation 

This section addresses three forms of flow-assisted pipe degradation: 1) flow-accelerated corrosion(a)

(FAC) of high-energy carbon steel and low alloy piping; 2) erosion-corrosion of moderate energy carbon 

steel piping; and 3) cavitation erosion of high- or moderate-energy carbon steel and stainless steel piping. 

All three degradation mechanisms cause localized or global internal pipe wall thinning. Depending on the 

specific system and location, cavitation erosion is also known to cause flow-induced vibratory-fatigue 

induced damage. 

5.3.1 Flow-Accelerated Corrosion(b)

Flow-accelerated corrosion (FAC) is defined as a chemical process whereby the normally protective 

oxide layer on carbon or low-alloy steel dissolves into a stream of flowing water or water-steam mixture. 

FAC corrosion rate controlling conditions are primarily electrochemical. FAC occurs in high-energy 

piping systems and can occur in single- and two-phase flow regions. The cause of FAC is a specific set 

of water chemistry conditions (for example, pH, level of dissolved oxygen), and absent of any mechanical 

 

(a) The term “flow-assisted corrosion” also has been used to describe this process. 

(b) In the United States, flow-accelerated corrosion (FAC) is commonly but incorrectly known as 

“erosion-corrosion.” Unlike FAC, the accelerated corrosion rates in the erosion-corrosion process are 

dominated by mechanical factors such as the impact of water droplets on the surface in two-phase 

flow steam systems, cavitation effects, or entrained particles. 
5.25 

contribution to the dissolution of the normally protective iron oxide (magnetite) layer on the inside pipe 

wall (as in pipe degradation by erosion-corrosion). The cause and effect of FAC is well understood, and 

the industry has implemented FAC inspection programs, as well as piping replacements using FACresistant materials such as stainless steel, carbon steel clad on the inside diameter with stainless steel, or 

chrome-molybdenum alloy steel. 

Comprehensive reviews of FAC-induced carbon steel degradation are documented in Brown (1984); 

Cragnolino et al. (1988); Fultz et al. (1988); and Shah et al. (1997). Included in Table 5.15 is a summary 

of the pipe failure experience attributed to FAC. It shows the pre-1987 and post-1987 service experience 

as an indication of the effectiveness of FAC mitigation programs implemented by industry in the 

aftermath of lessons learned from FAC-induced pipe failures at Trojan in 1985 and Surry Unit 2 in 1986. 

It is noted that BWR piping is less susceptible to FAC than corresponding systems in the PWR operating 

environment. This difference is attributed to the inherently different water chemistries. The service 

experience that is summarized in Table 5.15 covers the following plant systems: 

• condensate piping from the condensate booster pump and to the low pressure feedwater heater(s) 

• extraction steam piping 

• main feedwater piping from low-pressure heater(s) to outboard containment isolation valves; for 

PWR plants, also feedwater piping from inboard containment isolation valves to the steam generators 

• main steam piping from outboard containment isolation valves to the high-pressure turbine steam 

admission valve and turbine cross-over/cross-under piping 

• feedwater heater drain and vent piping 

• moisture separator reheater piping. 

Table 5.15 Summary of Service Experience Involving Flow-Accelerated Corrosion 

Number of Reported Pipe Failures Attributed to FAC 

1970–1987 1988–2005 1970–2005 

Plant 

Type Part ThroughWall 

ThroughWall Part ThroughWall ThroughWall Part ThroughWall ThroughWall 

BWR 85 94 61 100 146 194 

PWR 818 89 195 150 972 239 

• This tabulation is based on failure data as recorded in the proprietary PIPExp database; see Appendix D in 

NUREG-1829 (Tregoning et al. 2005). 

• “Through-wall” includes small leaks to structural failure (for example, Surry-2 in December 1986 and 

Mihama-3 in August 2004). 
5.26 

The current study did not identify any published applications of probabilistic structural mechanics models 

with a mechanistic treatment of material degradation in the open literature that address FAC failures as 

observed in the field. However, nonmechanistic treatments for wall-thinning rates have been used for 

probabilistic structural mechanics calculations. Plant-specific calculations have been performed to 

estimate leak and break probabilities for piping subject to user-specified rates of FAC degradation. These 

assessments were performed as part of RI-ISI evaluations. These calculations are normally not publicly 

available; however, some results have been reported in topical reports submitted for NRC approval or 

selected papers in the literature. Examples can be found in work reported by Westinghouse Owners 

Group (1999) and Bishop and McNeil (1999). 

Table 5.16 shows small leak probability results for a 305-mm (12-in.) carbon steel feedwater system 

(FW-12) line at Surry Unit 1. The dominant failure mechanism for this line was pipe material wastage 

due to FAC. The analysis was performed using the Westinghouse SRRA computer code (Westinghouse 

Owners Group 1997; Bishop 1993, 1997). In the SRRA code, the reduction in wall thickness by FAC 

was conservatively assumed to have the same impact on structural integrity as a circumferential crack that 

penetrated the wall to the same depth. Calculations were based on user-provided estimated inputs for 

material wastage rates. The results of this evaluation were included in an RI-ISI topical report 

(Westinghouse Owners Group 1997) submitted to the NRC for its review and approval. 

Other perspectives on the application of probabilistic models to estimate failure probability in wallthinned piping are found in Lee et al. (2006) and Vinod et al. (2003). The latter applies the models to 

FAC-susceptible piping to determine Markov model transition rates for use in RI-ISI program 

development (Fleming 2004). 

5.3.2 Erosion-Corrosion 

Erosion is the deterioration of pipe material because of the relative movement between a corrosive fluid 

and the metal surface. Generally the movement is quite rapid, and wear effect and abrasion are involved. 

The abrasive action of moving fluids is usually accelerated by the presence of solid particles or matter in 

suspension. When corrosion occurs simultaneously, the term erosion-corrosion is used. Erosioncorrosion is characterized in appearance by grooves, gullies, waves, rounded holes, and valleys, and 

usually exhibits a directional appearance. Most metals and alloys are susceptible. In nuclear power 

plants, the observed failures have largely been associated with pipe-wall degradation and is primarily of 

concern for raw water piping systems such as service water. Summarized in Table 5.17 is relevant 

service experience for carbon steel service water piping in U.S. BWR and PWR plants. It is organized by 

type of ultimate heat sink. 

The open literature search did not identify any published applications of probabilistic structural 

mechanics models with a mechanistic treatment of material wastage rates that address the erosioncorrosion failures as observed in the field. However, failure probability estimates for erosion-corrosion 

have been made based on structural reliability and risk assessment (SRRA) wall-thinning models 

(Westinghouse Owners Group 1997, 1999; Bishop and McNeil 1999) for RI-ISI applications. Bishop and 

McNeil (1999) point out that leak probabilities were found to be extremely sensitive to the potential for 

erosion-corrosion wastage. 
5.27 

Table 5.16 Surry Unit 1 305-mm (12-in.) Carbon Steel Feedwater System Piping Segment Small 

Leak Failure Probability by the Westinghouse SRRA Wastage Model Output 

(Westinghouse Owners Group 1999) 

Table 5.17 Summary of Service Experience Involving Erosion-Corrosion of Service Water Piping 

Number of Pipe Failure Records 1972–2005 

Ultimate Heat Sink All Part Through-Wall Through-Wall 

Brackish Water 49 2 47 

Lake/Pond Water 39 6 33 

River Water 71 26 45 

Sea Water 97 3 94 

All 256 37 219 

This tabulation is based on failure data as recorded in the proprietary PIPExp database; see 

Appendix D in NUREG-1829 (Tregoning et al. 2005). 
5.28 

5.3.3 Cavitation-Erosion 

Mechanical degradation of a solid material caused by cavitation is called cavitation-erosion. Cavitation 

erosion can be formed when cavity implosions are violent enough and they take place near the solid 

material. Cavitation-erosion can be identified from a specific rough mark on surfaces of component flow 

paths. 

Cavitation results from the formation and collapse (implosion) of vapor cavities in a liquid. As the liquid 

flows into a region where the pressure is reduced to the vapor pressure, it boils, and vapor pockets 

(bubbles) are formed. The bubbles are carried with the liquid until a region of higher pressure is reached, 

where they collapse. When the vapor bubbles are near (or in contact with) a solid boundary when they 

collapse, pitting and fatigue cracking have been observed on the surface. Cavitation can occur in a 

variety of nuclear plant system components such as pumps, turbines, elbows, valves, and flow orifices. 

With regard to piping pressure boundary degradation, the last two locations are considered more 

significant. 

The observed failures have largely been associated with piping base materials downstream of flow control 

valves and orifices of such systems as chemical and volume control, component cooling, decay heat 

removal, and service water. A summary of relevant service experience is given in Table 5.18. 

Table 5.18 Summary of Service Experience Involving Cavitation-Erosion 

Number of Pipe / Weld Failures 1974–2005 

Material All Part Through-Wall Through-Wall 

Carbon Steel 43 12 31 

Stainless Steel 37 11 26 

All 80 23 57 

This tabulation is based on failure data as recorded in the proprietary PIPExp database; 

see Appendix D in NUREG-1829 (Tregoning et al. 2005). 

The current study could not identify any applications of probabilistic models with mechanistic treatments 

of wastage rates that can address the cavitation-erosion failures as observed in the field. 

5.4 Local Corrosion Mechanisms 

Local corrosion damage generally occurs over relatively small areas; however, it can occur at high 

penetration rates. Examples of these types of degradation mechanisms in nuclear systems include crevice 

corrosion, pitting (oxygen or chloride), and microbiologically induced corrosion (MIC); the most 

prevalent of these mechanisms is MIC. 

MIC is normal electrochemical corrosion in which the microorganisms either chemically or physically 

change the conditions at the inner pipe wall to be favorable to corrosion. The microorganisms have the 

capability of changing the local environment, which then can be totally different from the bulk 

environment. MIC is usually the result of a complex interaction between different microorganisms, and a 

pure culture alone is seldom responsible for the corrosion. MIC appears as localized corrosion. MIC is of 
5.29 

concern in raw water piping systems, such as fire protection and service water, and where low-flow or 

stagnant-flow conditions occur. MIC could be a problem in closed-loop piping systems in the presence of 

contaminants. Carbon steel and, under certain conditions, stainless steels are susceptible to MIC. Where 

stainless steel material is used, MIC can attack welds and weld heat-affected zones. The susceptibility of 

raw water piping systems is also a function of water quality. As an example, systems using brackish 

water or seawater are not conducive to MIC attack. Summarized in Table 5.19 is relevant service 

experience for two raw water piping systems, namely fire protection water system and service water 

system. 

Table 5.19 Summary of Service Experience Involving MIC 

Number of Pipe Failures 

System All Part Through-Wall Through-Wall 

Fire Protection 30 -- 30 

Service Water 212 43 169 

All 242 43 199 

This tabulation is based on failure data as recorded in the proprietary PIPExp database; 

see Appendix D in NUREG-1829 (Tregoning et al. 2005). 

The current study could not identify any applications of probabilistic models with mechanistic treatments 

of wastage rates that can address the types of local corrosion failures as observed in the field. 
6.1 

6 Uncertainty in Estimated Failure Probabilities 

The estimated failure probability for any specific component is subject to large uncertainties, whether the 

estimate is based on data from operating experience or based on an application of a probabilistic structural 

mechanics model. When only operating experience is used, uncertainties are particularly large for lowprobability events such as ruptures or large breaks and/or failure predictions for specific piping locations. 

From Section 5, it is clear that there have been significant differences between independent estimates of 

failure frequencies that are intended to address the same component or a category of components. This 

section presents results from the literature that describe formal attempts to quantify levels of uncertainty 

in estimated failure probabilities. In one case (Khaleel and Simonen 1999), the approach estimated the 

uncertainties associated with individual inputs and modeling assumptions used for a probabilistic fracture 

mechanics calculation for a fatigue failure and then established the effects of these uncertainties on the 

calculated failure probabilities. The other case (Tregoning et al. 2005) reflects uncertainties assigned by 

participants in an expert elicitation process used to estimate LOCA frequencies. 

6.1 Uncertainties in Fatigue Calculations with PRAISE 

A methodology was developed and applied for quantifying the level of uncertainty associated with 

calculated fatigue failure probabilities for piping components (Khaleel and Simonen 1999). This study 

showed large uncertainties in calculated failure probabilities, with the uncertainties being particularly 

large when the calculated probabilities were very small, such as for pipe ruptures. The objective was to 

quantify the uncertainties associated with individual inputs to the PFM calculations and the uncertainties 

in the PFM model itself and then to determine the associated uncertainties in the resulting calculated 

piping failure probabilities. 

All calculations presented used the pc-PRAISE computer code (Harris and Dedhia 1992) to address both 

leak and break probabilities for piping components. A two-step process was used to quantify the 

uncertainties in calculated failure probabilities. The first step was a sensitivity study that identified those 

uncertainties having the greatest effect on the results from pc-PRAISE. The second step was a 

quantitative uncertainty analysis that addressed the most critical parameters as identified by the sensitivity 

calculations. 

While the selected methodology is generally applicable to PFM calculations for piping, the calculations 

addressed specific examples as follows: 

• mechanical and thermal fatigue of stainless steel piping due to the presence of fabrication defects (the 

initiation of fatigue cracks is not addressed) 

• a single 152.4-mm (6-in.) pipe with an inner radius of 70 mm (2.75 in.) and a wall thickness of 

14.3 mm (0.562 in.) 

• A wide range of cyclic stresses is addressed using the Q-factor approach of Khaleel and Simonen 

(1994a, b). The Q-factor depends both on the number of stress cycles (Nf) and on the corresponding 

magnitude of the cyclic stress (σ). It is defined by the equation Q = A•Nf •σ

4

 where A is a constant or 

normalizing parameter. The Q-factor has the same functional form as the Paris law for fatigue crack 
6.2 

growth and thereby reflects the predicted amount of fatigue crack growth. Q factors of 1, 100, and 

10,000 were addressed. The lowest value corresponds to typical locations in piping systems for 

which fatigue failures are not expected, whereas the highest value corresponds to locations exposed to 

thermal fatigue by which the pipe experiences unexpected and excessive levels of cyclic stress. 

While probabilities of both leaks (through-wall cracks) and pipe breaks were calculated, the main interest 

was in the uncertainties for leak probabilities. Uncertainties in leak probabilities were of particular 

interest to the development of a regulatory guide on RI-ISI (Parry 1997). 

The methodology and the resulting estimated uncertainties in the calculated probabilities were found to be 

similar to those for PFM calculations for reactor pressure vessels subject to pressurized thermal shock 

(Bozarth et al. 1985). While additional uncertainty evaluations were recommended to address other pipe 

sizes and failure mechanisms in addition to fatigue, the uncertainties as characterized by the results are 

believed to be representative of results expected for piping components in general. 

Inputs to pc-PRAISE consist of both probabilistic and deterministic parameters. Flaw depth, flaw-aspect 

ratio, crack-growth rate, and flow and ultimate stresses are probabilistic, while the pipe radius and wall 

thickness, internal pressure, dead-weight load, thermal stresses, and the number and level of cyclic 

stresses are deterministic. For preliminary sensitivity analyses, probabilities were first calculated using 

best-estimate values for all input parameters. These analyses were followed by calculations in which the 

best estimate inputs changed one by one. The numerical change for each input corresponded to the 

estimated uncertainty associated with that input parameter. Calculations indicated that leak probabilities 

were most sensitive to changes in the flaw depth, flaw-aspect ratio, crack-growth rate, and level and 

number of cyclic stresses. Changes in some of these parameters resulted in leak probabilities 

higher/lower than the baseline by as much as 2.5 orders of magnitude. The flow stress of the pipe 

material was nevertheless included as one of the more important input parameters due to large 

uncertainties in the simplified fracture-mechanics model used to predict pipe breaks (Wilkowski et al. 

1994). In this regard, the scatter in flow stress was used to represent the scatter in measured failure loads 

from degraded pipes as compared to predictions based on limit load theory. 

The objective of the uncertainty analysis was to estimate the mean, median, and standard deviation of the 

calculated leak and break probabilities. We estimated these statistical parameters using Monte Carlo 

simulations on the variables described above. Many of the eight most critical parameters were related to 

probabilistic inputs, whereas others (thermal stress, cyclic-stress level, and number of stress cycles) are 

deterministic inputs to pc-PRAISE. The uncertainty analyses required that both probabilistic and 

deterministic inputs be randomized. 

Calculations were performed for different sets of inputs obtained by sampling from triangular 

distributions that described the estimated uncertainties in the pc-PRAISE input parameters. Each 

Monte Carlo trial was a pc-PRAISE run that gave values of the failure probabilities. There were 

100 trials for each of the three Q values, which provided a sample of 100 failure probabilities from which 

means, medians, and standard deviations for the calculated probabilities were established. There were 

100 pc-PRAISE runs for each of the three Q values. This gave a total of 300 sets of piping-failure 

probabilities (leaks and breaks) as a function time, with the time span going from the start of plant 

operation to the end of life at 40 years. The focus was limited to cumulative failure probabilities at 
6.3 

40 years. The results gave conditional probabilities, corresponding to the assumption that one 

pre-existing crack exists in the weld. 

Figure 6.1 shows an example histogram (leak probabilities for Q = 100) along with the best-estimate 

probabilities. From the 100 calculated values of leak and break probabilities, it was possible to calculate 

mean and median values for comparison with the corresponding values from the best-estimate calculation 

from pc-PRAISE. The uncertainty calculations generally agreed (within an order of magnitude) with the 

best-estimate calculations. Figure 6.2 summarizes the results from all of the uncertainty analyses. 

Whereas there are relatively small differences between the best-estimate and median values of leak 

probabilities, Figure 6.2 shows more significant differences for mean values. The mean values for leak 

probabilities were about a factor of ten greater than the best estimate. The mean values for break 

probabilities were several orders of magnitude greater than the best estimate. It is noted that 

• Median values of probabilities correlated relatively well with the best estimates. 

• Differences between best-estimate and mean values for leak probabilities were 1 to 3 orders of 

magnitude, and 4 to 6 orders of magnitude for break probabilities. 

• The greater uncertainties for pipe-break probabilities compared to leak probabilities were largely due 

to the sensitivity of calculations to inputs for flaw depths and aspect ratios. There are little data to 

support the estimates for the low occurrence rates for very deep and long flaws. Estimates of 

probabilities for these rare defects are based on extrapolations having high levels of uncertainty. 

• The uncertainties in calculated leak probabilities become relatively small for leak probabilities greater 

than 10-4. Given a 40-year operating life for a reactor piping system, this 10-4 cumulative probability 

corresponds to a failure frequency of 2.5 × 10-6 leaks per weld per year. 

Figure 6.3 shows that the scatter in leak probabilities and break probabilities follows similar trends. It is 

noted that 

1. The uncertainties in calculated failure probabilities are highest when the probabilities themselves are 

very low (e.g., 10-8 per weld per 40-year life) and are much smaller when the failure probabilities 

become larger (e.g., 10-2). 

2. Large uncertainties for components with low failure probabilities may have a relatively small impact 

on the conclusions of risk-based assessments, because risk-based decisions are dominated by 

components with the higher values of failure probabilities. 

3. Failure probabilities for components with higher calculated failure probabilities can be more readily 

checked for consistency with plant operating experience; the ability to make such comparisons helps 

to minimize the uncertainties in these calculated probabilities. 
6.4 

0

2

4

6

8

10

12

14

16

18

20

-16 -15 -14 -13 -12 -11 -10 -9 -8 -7 -6 -5 -4 -3 -2 -1 0

Log(Leak Probability)

Count

best-estim ate

m edian mean

Figure 6.1 Histogram for Probability of Leak for Q = 100 

1E-20

1E-18

1E-16

1E-14

1E-12

1E-10

1E-08

1E-06

1E-04

1E-02

1E+00

0 1 10 100 1000 10000 100000

Q-Factor

Cumulative Failure Probability at 40 Years

Leak Probabilities

Mean of Uncertainty Analysis

Break Probabilities

Best Estimate Calculation

Median of Uncertainty Analysis

Mean of Uncertainty Analysis

Median of Uncertainty Analysis

Best Estimate Calculation

Figure 6.2 Comparisons of Probabilities from Uncertainty Analyses with Best-Estimate 

Calculations 
6.5 

1.E-15

1.E-14

1.E-13

1.E-12

1.E-11

1.E-10

1.E-09

1.E-08

1.E-07

1.E-06

1.E-05

1.E-04

1.E-03

1.E-02

1.E-01

1.E+00

1.E-11 1.E-10 1.E-09 1.E-08 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00

Best Estimate Failure Probability

Uncertainty in Failure Probability

Fatigue of Stainless Steel Piping

6-Inch Schedule 120 (2.75 inch OD x 0.562 inch Wall)

 - Cumulative Failure Probabiity Over 40-Years

c:\wptext\isi-rg\uncert1.xls

3\24\98

Median

 Failure Probability

Upper Bound

 Failure Probability

Lower Bound

 Failure Probability

Figure 6.3 Maximum Failure (Leak and Break) Probabilities from Sample of 100 Cases Evaluated 

by Uncertainty Analyses (25.4 mm = 1 in.) 

In summary, the results clearly show large uncertainties in calculated failure probabilities, with the 

uncertainties being particularly large when the best-estimate probabilities are relatively small. The 

median values of leak probabilities from the uncertainty calculations generally agree (within an order of 

magnitude) with the best-estimate calculations. This correlation supports the viability of using bestestimate calculations for risk-informed decision making. Initial flaw-size distributions were the greatest 

source of uncertainty in calculated failure probabilities because of the unavoidable difficulty in estimating 

the very low probabilities for the large fabrication flaws, which (if present) have a major impact on piping 

integrity. 

6.2 Uncertainties in LOCA Frequencies from Expert Elicitation 

Uncertainties were assigned by participants in an expert elicitation process used to estimate LOCA 

frequencies (Tregoning et al. 2005). Best-estimate values and the uncertainties for the failure frequencies 

from the experts served as inputs to a procedure that combined the individual inputs to establish overall 

trends that reflected the judgments of the group as a whole. These trends are shown by Figures 6.4 and 

6.5 for BWR and PWR plants, respectively. 

The objective of the study was to develop separate frequency estimates for BWR and PWR piping and 

non-piping passive system LOCAs. The sole focus was on event frequencies that initiate in unisolable 

primary system failures that can be exacerbated by age-related material degradation. While the central 

LOCA frequencies to reflect the consensus of the experts were the objective, quantifying the uncertainties 

and variability among the individual panel members was also an important objective. 
6.6 

Threshold Break Diameter (in)

0.1 1 10 100

LOCA Frequency ({cal-yr}-1) 10-10

10-9

10-8

10-7

10-6

10-5

10-4

10-3

10-2

Median

Mean

95th Percentile

BWR: Baseline Results

median and 95th percentile results offset slightly for clarity

Figure 6.4 Total BWR LOCA Frequencies (means, medians, and 95th percentile values) as a 

Function of the Threshold Break Diameter at 25 Years of Plant Operations 

Threshold Break Diameter (in)

0.1 1 10 100

Frequency ({cal-yr}-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: Baseline Results

median and 95th percentile results offset for clarity

Figure 6.5 Total PWR LOCA Frequencies (means, medians, and 95th percentile values) as a 

Function of the Threshold Break Diameter at 25 Years of Plant Operations 

The elicitation developed generic, or average, values for the commercial fleet of nuclear power plants. 

Thus, the uncertainty bounds do not represent LOCA frequency estimates for individual plants that may 

differ from the generic values. This study partitioned LOCA sizes into three smaller categories, which 
6.7 

were consistent with historical definitions of small break, medium break, and large break flow rates. 

Additionally, three larger LOCA categories were defined within the classical large break LOCA regime to 

examine trends with increasing break size, up to and including, a DEGB of the largest piping system in 

the plant. 

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

processes used in NUREG-1150 (USNRC 1990b). To begin, the project staff first identified the issues to 

be evaluated before a panel of 12 experts was selected. 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 that impact LOCA 

frequencies into fundamental elements. A subset of the panel was created to develop quantitative 

estimates of LOCA frequencies associated for selected 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 

refine the questionnaire. A second meeting was held with the entire panel to review the base-case results, 

review the elicitation questions, and finalize the formulation of remaining technical issues. After the 

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

elicitation questionnaire. 

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) and select appropriate basecase conditions for comparison. The panelists would then provide 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 

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

true answer lays 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 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. After this meeting, the panel members provided updated responses, and 

the project staff compiled the panel’s responses to develop preliminary estimates of the LOCA 

frequencies. These preliminary estimates were presented to the panel at a wrap-up meeting. Panel 

members were invited to modify any of their responses based on group discussions. Final estimates of 

the LOCA frequencies were then calculated and provided to the panel members for final review and 

quality assurance. 

The quantitative responses were analyzed separately for each panel member to develop individual LOCA 

frequency estimates. A unified analysis format was developed to ensure consistency and commonalty in 

processing the panelists’ inputs. The panelists’ mid-value and upper bound and lower bound were 

assumed to represent the median, 95th, and 5th percentiles, respectively, of their subjective uncertainty 
6.8 

distributions for each elicitation response. The final output for each panelist was BWR- and PWRspecific total LOCA frequency estimates of the mean, median, 5th and 95th percentiles. 

While there was general qualitative agreement among the panelists about important technical issues and 

LOCA contributing factors, the individual quantitative estimates were much more variable. The LOCA 

frequencies for the individual panelists were aggregated to obtain group LOCA frequency estimates, 

along with measures of panel diversity. The methodology (Tregoning et al. 2005) was consistent with the 

elicitation objectives and structure and ensures that the LOCA frequency parameter estimates were not 

dominated by outlier estimates. 

The median, mean, and 95th percentile baseline estimates are graphically presented in Figures 6.4 and 

6.5. The 95% confidence interval calculated for these parameters are also illustrated. A measure of the 

individual uncertainty is illustrated by the difference between the median and 5thth or 95th percentile 

estimates. Differences of group opinion are reflected in the confidence bounds. 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 greater 

extrapolation required from available service data. 

In summary, the plots of Figures 6.4 and 6.5 reflect both the expert-to-expert variations in the estimated 

LOCA frequencies and the uncertainties expressed by the individual experts. Small breaks (e.g., throughwall cracks) are seen to have much greater estimated frequencies (e.g., 10-2 per plant per year) than the 

corresponding frequencies (e.g., 10-7 per plant per year) for larger breaks (pipe ruptures). In terms of the 

logarithmic scale used to plot the frequencies, the uncertainties are much greater for low values of failure 

frequencies (about one order of magnitude) than for the higher values of failure frequencies (about three 

orders of magnitude). 
7.1 

7 Conclusions 

A search of the open literature was performed, which identified more than 7500 publications from 1980 to 

present that have addressed the degradation of reactor pressure boundary components. Approximately 

60% of the documents were related to corrosion mechanisms (e.g., FAC, erosion-corrosion, cavitation 

erosion, boric acid corrosion, crevice corrosion, pitting). The remaining documents are divided equally 

between fatigue and stress corrosion cracking mechanisms. The vast majority of the references address 

various mechanistic aspects of degradation mechanisms and apply deterministic rather than probabilistic 

models. 

The literature search did not generally identify the numerous plant-specific studies completed by industry 

owners’ groups and by research organizations such as the EPRI Materials Reliability Program. These 

results are typically not publicly available. In addition, many plant-specific probabilistic fracture 

mechanics calculations have been performed in support of RI-ISI evaluations. Although these 

calculations are normally not publicly available, portions of these results are sometimes included in 

topical reports submitted for NRC approval or in selected papers in the open literature. 

Although the probabilistic treatments and characterizations of fatigue and stress corrosion cracking are 

well documented in the literature, only a limited number of these publications document plant-specific 

and component-specific probabilistic evaluations based on actual design and operating stresses. 

Examples from the more relevant of these studies were discussed in Section 5 of this report. 

The literature search did not identify any applications of probabilistic structural mechanics models for 

local corrosion. Failure probability estimates have, however, been reported in plant-specific RI-ISI 

evaluations. In some applications, failure frequency estimates were based on statistical evaluations of 

service failure data. Failure probability estimates have also been based on assumed estimates of corrosion 

rates and qualitative evaluations made by plant expert panels. 

The literature search did not identify any applications of probabilistic structural mechanics models for 

flow-assisted degradation mechanisms (FAC, erosion-corrosion, and cavitation erosion) known to be 

active in nuclear power plant systems. As with other corrosion mechanisms, failure probability estimates 

have been performed in support of plant-specific RI-ISI evaluations. In these cases, failure frequency 

estimates have been based on estimated inputs for material wastage/wall thinning rates. In these models, 

wall thinning was simplistically assumed to have the same impact on structural integrity as a 

circumferential crack that penetrated the wall to the same depth. 

Formal attempts to quantify uncertainties in estimated failure probabilities show that the estimated failure 

probabilities for particular components are subject to large uncertainties, whether the estimate is based on 

data from operating experience or based on application of probabilistic structural mechanics models. The 

reported uncertainties are particularly large when the best-estimate probabilities are relatively small. 

Initial flaw size distributions were identified as particularly large sources of uncertainty in calculated 

failure probabilities because of the unavoidable difficulty in estimating the very low probabilities for the 

large fabrication flaws, which (if present) have a major impact on piping integrity. 

There can also be differences in calculated probabilities related both to the probabilistic fracture 

mechanics models as implemented in the various computer codes and the judgments made in defining 
7.2 

input parameters for plant-specific calculations. Ongoing efforts are needed for improved data on 

material degradation rates and damage models that describe the performance of components under field 

conditions. In many cases, such as for damage mechanisms associated with wall thinning, there are no 

recognized probabilistic structural mechanics codes to predict failure probabilities. 
8.1 

8 References 

10 CFR 50.61. “Fracture toughness requirements for protection against pressurized thermal shock 

events.” Code of Federal Regulations, Part 50.61, Title 10, 1996. 

Attwood CL, VN Shah, and WJ Galyean. 1999. “Analysis of Pressurized Water Reactor Primary 

Coolant Leak Events Caused by Thermal Fatigue.” In Proc. ESREL ‘99 – The Tenth European 

Conference on Safety and Reliability. AA Balkema, Rotterdam (The Netherlands). 

Bell CD and OJV Chapman. 2003. Description of PRODIGAL. NURBIM Report D4/Appendix F, 

Rolls-Royce plc., Derby, United Kingdom. 

Bishop BA. 1993. Benchmarking of Probabilistic Fracture Mechanics Analyses of Reactor Vessels 

Subjected to Pressurized Thermal Shock Loading. EPRI Research Project 2975-5, Electric Power 

Research Institute, Palo Alto, California. 

Bishop BA. 1997. “An Updated Structural Reliability Model for Piping Risk Informed ISI.” In Fatigue 

and Fracture - 1997, Volume 2, PVP-Vol. 346, pp. 245–252. American Society of Mechanical 

Engineers, New York. 

Bishop BA and A McNeil. 1999. “Use of Structural Reliability Models for Piping Risk Informed ISI at 

40 and 60 Years.” In Pressure Vessel and Piping Codes and Standards, PVP-Vol. 383, pp. 61–67. 

American Society of Mechanical Engineers, New York. 

Bozarth D, CL Mason, and JW Minarick. 1985. “Sensitivity and Uncertainty Analyses of Estimated 

Through-Wall Crack Frequencies for H.B. Robinson Unit 2.” Chapter 7 in Pressurized Thermal Shock 

Evaluation of the H.B Robinson Unit 2 Nuclear Power Plant, NUREG/CR-4183, Volume 1, U.S. Nuclear 

Regulatory Commission, Washington, D.C. 

Brickstad B and W Zang. 2001. NURBIT Nuclear RBI Analysis Tool, A Software for Risk Management 

of Nuclear Components. Technical Report No 10334900-1, DNV, Stockholm, Sweden. 

Brickstad B, OJV Chapman, T Schimpfke, H Schulz, and A Muhammed. 2004. Review and 

Benchmarking of SRM and Associated Software. NURBIM Final Report D4, Contract FIKS-CT-2001-

00172, DNV, Stockholm, Sweden. 

Brown EJ. 1984. Erosion in Nuclear Power Plants. AEOD/E4 16, Office for Analysis and Evaluation 

of Operational Data, U.S. Nuclear Regulatory Commission, Washington, D.C. 

Cattant F, F Vaillant, JM Boursier, and S De-Vito. 2005. “Contribution of Hot and Cold Laboratory 

Investigations to the Resolution of EDF Alloys 600/182/82 PWSCC Issues.” In Materials Reliability 

Program: Proceedings of the 2005 International PWSCC of Alloy 600 Conference and Exhibit Show 

(MRP-154), TR-1012089, Electric Power Research Institute, Palo Alto, California. 
8.2 

Cheverton RD and DG Ball. 1984. OCA-P, A Deterministic and Probabilistic Fracture-Mechanics Code 

for Applications to Pressure Vessels. NUREG/CR-3618, U.S. Nuclear Regulatory Commission, 

Washington, D.C. 

Chopra OK. 2002. “Mechanism of Fatigue Crack Initiation in Austenitic Stainless Steel in LWR 

Environments.” In Pressure Vessel and Piping Codes and Standards, PVP-Vol. 439, pp. 133–142, 

RD Rana (ed.). American Society of Mechanical Engineers, New York. 

Chopra OK and WJ Shack. 2001. Environmental Effects on Fatigue Crack Initiation in Piping and 

Pressure Vessel Steels. NUREG/CR-6717, U.S. Nuclear Regulatory Commission, Washington, D.C. 

Cragnolino G, C Czajkowsi, and WJ Shack. 1988. Review of Erosion-Corrosion in Single-Phase Flows. 

NUREG/CR-5156, U.S. Nuclear Regulatory Commission, Washington, D.C. 

Dickson TL. 1994. FAVOR: A Fracture Mechanics Analysis Code for Nuclear Reactor Pressure Vessel, 

Release 9401. ORNL/NRC/LTR/94/1, Oak Ridge National Laboratory, Oak Ridge, Tennessee. 

Dillstrom P. 2003. A Short Description of ProSACC. NURBIM Report D4/Appendix G, DNV, 

Stockholm, Sweden. 

EPRI. 1988. Assessment of Remedies for Degraded Piping. NP-5881-LD, Electric Power Research 

Institute, Palo Alto, California. 

EPRI. 1994. EPRI Fatigue Management Handbook. TR-104534, Electric Power Research Institute, 

Palo Alto, California. 

EPRI. 2000. Operating Experience Regarding Thermal Fatigue of Unisolable Piping Connected to PWR 

Reactor Coolant System (MRP-25). TR-1001006, Electric Power Research Institute, Palo Alto, 

California. 

EPRI. 2002. Crack Growth Rates for Evaluating Primary Water Stress-Corrosion Cracking (PWSCC) of 

Thick-Wall Alloy 600 Materials (MRP-55) Revision 1. TR-1006695, Electric Power Research Institute, 

Palo Alto, California. NRC ADAMS Accession No. ML02301050. 

EPRI. 2004. Materials Reliability Program: Crack Growth Rates for Evaluating Primary Water Stress 

Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115NP). TR-1006696, Electric 

Power Research Institute, Palo Alto, California. NRC ADAMS Accession No. ML051450555. 

Failure Analysis Associates. 1990. PRAISE Enhancements to Include General Strain Hardening 

Exponents and Mid-Life Residual Stress and Water Chemistry Changes. FaAA-SF-R-90-06-06, 

UCRL-CR-105339. Prepared for Lawrence Livermore National Laboratory, Livermore, California. 

Fleming KN. 2004. “Markov Models for Evaluating Risk-Informed In-service Inspection Strategies for 

Nuclear Power Plant Piping Systems.” Reliability Engineering and System Safety 83:27–45. 
8.3 

Fleming KN, SR Gosselin, and J Mitman. 1999. “Application of Markov Models and Service Data to 

Evaluate the Influence of Inspection on Pipe Rupture Frequencies.” In Pressure Vessel and Piping Codes 

and Standards, PVP-Vol. 383, pp. 69–75. American Society of Mechanical Engineers, New York. 

Fultz KO, MA Kruslicky, WD McDowell, Jr., KR Bartlett, WD Campbell, and RL Coleman. 1988. 

Nuclear Regulation: Action Needed to Ensure That Utilities Monitor and Repair Pipe Damage. RCED88-73, U.S. General Accounting Office, Washington, D.C. 

Garud YS. 1997. A Simplified Model for Assessment of SCC Initiation Time in Alloy 600. TR-109137, 

Electric Power Research Institute, Palo Alto, California. 

Garud YS and RS Pathania. 1999. “A Simplified Model for SCC Initiation Susceptibility in Alloy 600 

with the Influence of Cold Work Layer and Strength Characteristics.” In Ninth International Symposium 

on Environmental Degradation of Materials in Nuclear Power Systems—Water Reactors, pp. 261–268. 

PP Ford, SM Bruemmer, and GS Was (eds). The Minerals, Metals & Materials Society (TMS), 

Warrendale, Pennsylvania. 

Gavenda DJ, PR Luebbers, and OK Chopra. 1997. “Crack Initiation and Crack Growth Behavior of 

Carbon and Low-Alloy Steels.” In Fatigue and Fracture – Volume 1, PVP-Vol. 350, pp. 243–255. 

S Rahman, KK Yoon, S Bhandari, R Warke, and JM Bloom (eds). American Society of Mechanical 

Engineers, New York. 

Gosselin SR, FA Simonen, RG Carter, JM Davis, and GL Stevens. 2005. “Enhanced ASME Section XI 

Appendix L Flaw Tolerance Procedure.” In Proceedings of the PVP2005, 2005 ASME Pressure Vessels 

and Piping Conference. July 17–21, 2005, Denver, Colorado. PVP2005-71100, American Society of 

Mechanical Engineers, New York. 

Grimmel B and WH Cullen Jr. 2005. U.S. Plant Experience with Alloy 600 Cracking and Boric Acid 

Corrosion of Light-Water Reactor Pressure Vessel Materials. NUREG-1823, U.S. Nuclear Regulatory 

Commission, Washington, D.C. 

Harris DO and DD Dedhia. 1992. Theoretical and User’s Manual for pc-PRAISE, A Probabilistic 

Fracture Mechanics Computer Code for Piping Reliability Analysis. NUREG/CR-5864, U.S. Nuclear 

Regulatory Commission, Washington, D.C. 

Harris DO and DD Dedhia. 1998. WinPRAISE 98 PRAISE Code in Windows. Engineering Mechanics 

Technology, Inc., San Jose, California. 

Harris DO, DD Dedhia, and ED Eason. 1986a. “Probabilistic Analysis of Initiation and Growth of Stress 

Corrosion Cracks in BWR Piping.” ASME Paper 86-PVP-11. American Society of Mechanical 

Engineers, New York. 

Harris DO, DD Dedhia, ED Eason, and SD Patterson. 1985. Probabilistic Treatment of Stress Corrosion 

Cracking in Sensitized 304 Stainless Steel Weldments in BWR Piping. Failure Analysis Associates Report 

to Lawrence Livermore National Laboratory, Livermore, California. 
8.4 

Harris DO, DD Dedhia, ED Eason, and SD Patterson. 1986b. Probability of Failure in BWR Reactor 

Coolant Piping: Probabilistic Treatment of Stress Corrosion Cracking in 304 and 316NG BWR Piping 

Weldments. NUREG/CR-4792, Vol. 3, U.S. Nuclear Regulatory Commission, Washington, D.C. 

Harris DO, EY Lim, and DD Dedhia. 1981. Probability of Pipe Fracture in the Primary Coolant Loop 

of a PWR Plant Volume 5: Probabilistic Fracture Mechanics Analysis - Load Combination Program 

Project 1 Final Report. NUREG/CR-2189, Vol. 5, U.S. Nuclear Regulatory Commission, Washington, 

D.C. 

Hazelton WS and WH Koo. 1988. Technical Report on Material Selection and Processing Guidelines 

for BWR Coolant Pressure Boundary Piping. NUREG-0313, Rev. 2, U.S. Nuclear Regulatory 

Commission, Washington, D.C. 

Heasler PG and FA Simonen. 1990. “A Probabilistic Evaluation of Alternative Weld Inspection 

Sampling Plans.” In Uncertainty Modeling and Analysis – Proceedings of the First International 

Symposium on Uncertainty Modeling and Analysis (ISUMA ’90), BM Ayyub (ed), pp. 432–436. IEEE 

Computer Society Press, Los Alamitos, California. 

Holman GS and CK Chou. 1988. Probability of Failure in BWR Reactor Coolant Piping. 

NUREG/CR-4792, Vol. 1, U.S. Nuclear Regulatory Commission, Washington, D.C. 

Kassir M, S Sharma, M Reich, and M-T Chang. 1985. An Evaluation of Stress Corrosion Crack Growth 

in BWR Piping Systems. NUREG/CR-4221, U.S. Nuclear Regulatory Commission, Washington, D.C. 

Keating RF, P Hernalsteen, and JA Begley. 1995. Burst Pressure Correlation for Steam Generator 

Tubes with Throughwall Axial Cracks. TR-105505, Electric Power Research Institute, Palo Alto, 

California. 

Keisler J, OK Chopra, and WJ Shack. 1995. Fatigue Strain–Life Behavior of Carbon and Low–Alloy 

Steels, Austenitic Stainless Steels, and Alloy 600 in LWR Environments. NUREG/CR-6335, U.S. Nuclear 

Regulatory Commission, Washington, D.C. 

Khaleel MA and FA Simonen. 1994a. “A Parametric Approach to Predicting the Effects of Fatigue on 

Piping Reliability.” In Service Experience and Reliability Improvement: Nuclear, Fossil, and 

Petrochemical Plants, ASME PVP-Vol. 288, pp. 117–125. American Society of Mechanical Engineers, 

New York. 

Khaleel MA and FA Simonen. 1994b. “The Effects of Initial Flaw Sizes and Inservice Inspection on 

Piping Reliability.” In Service Experience and Reliability Improvement: Nuclear, Fossil, and 

Petrochemical Plants, PVP-Vol. 288, pp. 95–07. American Society of Mechanical Engineers, New York. 

Khaleel MA and FA Simonen. 1999. “Uncertainty Analyses for Calculated Failure Probabilities or 

Piping Welds.” In Pressure Vessel and Piping Codes and Standards - 1999, PVP-Vol. 383, pp. 77–90. 

American Society of Mechanical Engineers, New York. 
8.5 

Khaleel MA, OJV Chapman, DO Harris, and FA Simonen. 1999. “Flaw Size Distribution and Flaw 

Existence Frequencies in Nuclear Piping.” In Probabilistic and Environmental Aspects of Fracture and 

Fatigue, ASME PVP-Vol. 386, pp. 127–144. American Society of Mechanical Engineers, New York. 

Khaleel MA, FA Simonen, DO Harris, and D Dedhia. 1995. “The Impact of Inspection on Intergranular 

Stress Corrosion Cracking for Stainless Steel Piping.” In Risk and Safety Assessments: Where Is the 

Balance? PVP-Vol. 296, pp. 411–422. American Society of Mechanical Engineers, New York. 

Khaleel MA, FA Simonen, HK Han, DO Harris, and D Dedhia. 2000. Fatigue Analysis of Components 

for 60-Year Plant Life. NUREG/CR-6674, U.S. Nuclear Regulatory Commission, Washington, D.C. 

Lee SM, YS Chang, JB Choi, and YJ Kim. 2006. “Failure Probability Assessment of Wall-Thinned 

Nuclear Pipes Using Probabilistic Fracture Mechanics.” Nuclear Engineering and Design 

236(4):350–358. 

Lydell BOY. 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. 

Lydell BOY, E Mathet, and K Gott. 2004. “Piping Service Life Experience in Commercial Nuclear 

Power Plants: Progress with the OECD Pipe Failure Data Exchange Project.” In Proceedings of the 2004 

ASME Pressure Vessels and Piping Conference, PVP2004-2961. American Society of Mechanical 

Engineers, New York. 

Miraglia F J Jr. 1988. “NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping.” Generic 

Letter 88-01, U.S. Nuclear Regulatory Commission, Washington, D.C. 

Mohammed AA. 2003. A Short Description of STRUEL. NURBIM Report D4/Appendix H. The 

Welding Institute, Cambridge, United Kingdom. 

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. 

Nuclear Energy Agency. 2006. “NEA Joint Projects – OECD/NEA OPDE Project.” Available at 

http://www.nea.fr/html/jointproj/opde.html (September 2006). 

Olson DE. 1985. “Piping Vibration Experience in Power Plants.” In Pressure Vessel and Piping 

Technology 1985: A Decade of Progress, pp. 689–705. Pressure Vessel and Piping Division, American 

Society of Mechanical Engineers, New York. 

Parry GW. 1997. “Quality of PRA for Use in Regulatory Decision Making.” In Risk Informed Decision 

Making, PVP-Vol. 358, pp. 81–85, American Society of Mechanical Engineers, New York. 

Poloski JP, DG Marksberry, CL Atwood, and WJ Galyean. 1999. Rates of Initiating Events at U.S. 

Nuclear Power Plants: 1987-1995. NUREG/CR-5750, U.S. Nuclear Regulatory Commission, 

Washington, D.C. 
8.6 

Riccardella PC, DC Rosaio, and SR Gosselin. 1997. “Fracture Mechanics Analysis of Socket Welds 

under High Cycle Vibrational Loading.” PVP-Vol. 353, pp. 23–29. American Society of Mechanical 

Engineers, New York. 

Rudland DL, H Xu, G Wilkowski, P Scott, N Ghadiali, and F Brust. 2006. “Development of a New 

Generation Computer Code (PRO-LOCA) for the Prediction of Break Probabilities for Commercial 

Nuclear Power Plants Loss-of-Coolant Accidents.” In 2006 Proceedings of the ASME Pressure Vessels 

and Piping Conference, July 23–27, 2006, Vancouver, British Columbia, Canada. American Society of 

Mechanical Engineers, New York. 

Schimpfke T. 2003. A Short Description of the Piping Reliability Code PROST. NURBIM Report 

D4/Appendix C, Gesellschaft fur Anlagen-und Reaktorsicherheit (GRS), Berlin, Germany. 

Scott PM. 1991. “An Analysis of Primary Water Stress Corrosion Cracking in PWR Steam Generators.” 

Presented at NEA/CSNI Specialist Meeting on Operating Experience with Steam Generators, 

September 16–20, 1991, Brussels, Belgium. 

Shah VN, AG Ware, and AM Porter. 1997. Review of Industry Efforts to Manage Pressurized Water 

Reactor Feedwater Nozzle, Piping, and Feedring Cracking and Wall Thinning. NUREG/CR-6456, U.S. 

Nuclear Regulatory Commission, Washington, D.C. 

Shah VN, MB Sattison, CL Atwood, AG Ware, GM Grant, and RS Hartley. 1998. “Assessment of 

Pressurized Water Reactor Primary System Leaks. NUREG/CR-6582, U.S. Nuclear Regulatory 

Commission, Washington, D.C. 

Pipe Crack Study Group. 1979. Investigation and Evaluation of Stress-Corrosion Cracking in Piping of 

Light Water Reactor Plants. NUREG-0531, U.S. Nuclear Regulatory Commission, Washington, D.C. 

Pipe Crack Study Group. 1980. Investigation and Evaluation of Cracking Incidents in Piping in 

Pressurized Water Reactors. NUREG-0691, U.S. Nuclear Regulatory Commission, Washington, D.C. 

Simonen FA, KI Johnson AM Liebetrau, DW Engel, and EP Simonen. 1986. VISA-II - A Computer 

Code for Predicting the Probability of Reactor Pressure Vessel Failure. NUREG/CR-4486, U.S. Nuclear 

Regulatory Commission, Washington, D.C. 

Simonen FA, DO Harris, and DD Dedhia. 1998. “Effect of Leak Detection on Piping Failure 

Probabilities.” In Fatigue Fracture and Residual Stress, PVP-Vol. 373, pp. 105–113. American Society 

of Mechanical Engineers, New York. 

Simonen FA and MA Khaleel. 1998a. “A Probabilistic Fracture Mechanics Model for Fatigue Crack 

Initiation in Piping.” In Fatigue, Fracture and Residual Stress, PVP-Vol. 373, pp. 27–34. American 

Society of Mechanical Engineers, New York. 

Simonen FA and MA Khaleel. 1998b. “Effects of Flaw Sizing Errors on the Reliability of Vessels and 

Piping.” ASME Journal of Pressure Vessel Technology 120:365–373. 
8.7 

Simonen FA and MA Khaleel. 1998c. “On the Contribution of Buried Flaws to Piping Failure 

Probabilities.” In Fatigue Fracture and Residual Stress, PVP-Vol. 373, pp. 69–75. American Society of 

Mechanical Engineers, New York. 

Simonen FA and SR Gosselin. 2001. “Life Prediction and Monitoring of Nuclear Power Plant 

Components for Service-Related Degradation,” Transactions of the ASME, J. PVT, Vol. 123, 1:58–64. 

Snaider R. 1980. BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking: 

Resolution of Generic Technical Activity A-10. NUREG-0619-Rev-1, U.S. Nuclear Regulatory 

Commission, Washington, D.C. 

Strosnider JR Jr, WH Koo, JA Davis, and MC Modes. 2000. Pipe Cracking in U.S. BWRs: A 

Regulatory History. NUREG-1719, U.S. Nuclear Regulatory Commission, Washington, D.C. 

Tang SS, PC Riccardella, ML Herrera, D Covill, R Carter, and L Steinert. 1999. “The Effect of Adjusted 

Reference Temperature on the Probability of Failure in Boiling Water Reactor Vessel Welds.” In 

Fracture, Design Analysis of Pressure Vessels, Heat Exchangers, Piping Components, and Fitness for 

Service – 1999, PVP-Vol. 388, pp. 121–126. American Society of Mechanical Engineers, New York. 

Tregoning R, L Abramson and P Scott. 2005. Estimating Loss-of-Coolant Accident (LOCA) Frequencies 

Through the Elicitation Process (NUREG-1829) - Draft Report for Comment. Draft NUREG-1829 

(June 30, 2005). U.S. Nuclear Regulatory Commission, Washington, D.C. www.nrc.gov/readingrm/doc-collections/nuregs/staff/sr1829/index.html 

USNRC. 1975. Reactor Safety Study: An Assessment of Accident Risks I U.S. Commercial Nuclear 

Power Plants. NUREG-1150, U.S. Nuclear Regulatory Commission, Washington, D.C. 

USNRC. 1979. “Cracking in Feedwater System Piping.” Bulletin 79-13, Rev. 2, U.S. Nuclear 

Regulatory Commission, Washington, D.C., October 16, 1979. 

USNRC. 1980. BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking. NUREG0619, U.S. Nuclear Regulatory Commission, Washington, D.C. 

USNRC. 1988a. “Safety Injection pipe Failure.” Information Notice 88-01, U.S. Nuclear Regulatory 

Commission, Washington, D.C., January 27, 1988. 

USNRC. 1988b. “Thermal Stresses in Piping Connected to Reactor Coolant Systems.” Bulletin 88-08, 

U.S. Nuclear Regulatory Commission, Washington, D.C., June 22, 1988. 

USNRC. 1988c. “Thermal Stresses in Piping Connected to Reactor Coolant Systems.” Bulletin 88-08, 

Supplement 1, U.S. Nuclear Regulatory Commission, Washington, D.C., June 24, 1988. 

USNRC. 1988d. “Thermal Stresses in Piping Connected to Reactor Coolant Systems.” Bulletin 88-08, 

Supplement 2, U.S. Nuclear Regulatory Commission, Washington, D.C., August 4, 1988. 
8.8 

USNRC. 1989. “Thermal Stresses in Piping Connected to Reactor Coolant Systems.” Bulletin 88-08, 

Supplement 3, U.S. Nuclear Regulatory Commission, Washington, D.C., April 11, 1989. 

USNRC. 1990a. Investigation and Evaluation of Cracking Incidents in Piping in Pressurized Water 

Reactors. NUREG-0691, U.S. Nuclear Regulatory Commission, Washington, D.C. 

USNRC. 1990b. “Severe Accident Risks: An Assessment of Five U.S. Nuclear Power Plants.” WASH1400, U.S. Nuclear Regulatory Commission, Washington, D.C. 

USNRC. 1992. “Cracking of Valves in the Condensate Return Lines in a BWR Emergency Condenser 

System.” Information Notice 92-50, U.S. Nuclear Regulatory Commission, Washington, D.C., July 2, 

1992. 

USNRC 1997. “Degradation of Control Rod Drive Mechanism Nozzle and Other Vessel Closure Head 

Penetrations.” Generic Letter 97-01, U.S. Nuclear Regulatory Commission, Washington, D.C., April 1, 

1997. 

USNRC. 1998. “Cavitation Erosion of Letdown Line Orifices Resulting in Fatigue Cracking of Pipe 

Welds.” Information Notice 98-45, U.S. Nuclear Regulatory Commission, Washington, D.C., 

December 15, 1998. 

USNRC. 2001a. “Through-Wall Circumferential Cracking of Reactor Pressure Vessel Head Control Rod 

Drive Mechanism Penetration Nozzles at Oconee Nuclear Station, Unit 3.” Information Notice 2001-05, 

U.S. Nuclear Regulatory Commission, Washington, D.C., April 30, 2001. 

USNRC. 2001b. “Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles.” 

Bulletin 2001-01, U.S. Nuclear Regulatory Commission, Washington, D.C., August 3, 2001. 

USNRC. 2002. “Reactor Pressure Vessel Head and Vessel Head Penetration Nozzle Inspection 

Programs.” Bulletin 2002-02, U.S. Nuclear Regulatory Commission, Washington, D.C., August 9, 2002. 

USNRC. 2003a. “Leakage Found on Bottom-Mounted Instrumentation Nozzles.” Information Notice 

2003-11, U.S. Nuclear Regulatory Commission, Washington, D.C., August 13, 2003. 

USNRC. 2003b. “Leakage Found on Bottom-Mounted Instrumentation Nozzles.” Information Notice 

2003-11, Supplement 1, U.S. Nuclear Regulatory Commission, Washington, D.C., August 13, 2003. 

USNRC. 2003c. “Leakage from Reactor Pressure Vessel Lower Head Penetrations and Reactor Coolant 

Pressure Boundary Integrity.” Bulletin 2003-02, U.S. Nuclear Regulatory Commission, Washington, 

D.C., August 21, 2003. 

USNRC. 2004. “Inspection of Alloy 82/182/600 Materials Used in the Fabrication of Pressurizer 

Penetrations and Steam Space Piping Connections at Pressurized-Water Reactors.” Bulletin 2004-01, 

U.S. Nuclear Regulatory Commission, Washington, D.C., May 18, 2004. 
8.9 

USNRC. 2007. Expert Panel on Proactive Materials Degradation Assessment. NUREG/CR-6923. U.S. 

Nuclear Regulatory Commission, Washington, D.C. 

Vinod G, SK Bidhar, HS Kushwaha, AK Verma, and A Srividya. 2003. “A Comprehensive Framework 

for Evaluation of Piping Reliability Due to Erosion-Corrosion for Risk-Informed Inservice Inspection.” 

Reliability Engineering and System Safety 82:187–193. 

Vo TV, PG Heasler, SR Doctor, FA Simonen, and BF Gore. 1991. “Estimates of Rupture Probabilities 

for Nuclear Power Plant Components: Expert Judgment Elicitation.” Nuclear Technology 96:259–271. 

Wachter O and G Brümmer. 1997. “Experiences with Austenitic Stainless Steels in Boiling Water 

Reactors.” Nuclear Engineering and Design 168:35–52. 

Ware AG, DK Morton, and ME Nitzel. 1995. Application of NUREG/CR-5999 Interim Fatigue Curves 

to Selected Nuclear Power Plant Components. NUREG/CR-6260, U.S. Nuclear Regulatory Commission, 

Washington, D.C. 

Westinghouse Owners Group. 1997. Application of Risk-Informed Methods to Piping Inservice 

Inspection Topical Report. WCAP-14572, Rev. 1, Westinghouse Electric Co., Pittsburgh, Pennsylvania. 

Westinghouse Owners Group. 1999. Westinghouse Owners Group Application of Risk Informed 

Methods to Piping Inservice Inspection Topical Report. WCAP-14572, Revision 1-NP-A, Westinghouse 

Electric Co., Pittsburgh, Pennsylvania. 

White GA, J Hickling, and LK Mathews. 2003. “Crack Growth Rates for Evaluating PWSCC of ThickWall Alloy 600 Material.” In Proceedings of the Eleventh International Conference on Environmental 

Degradation of Materials in Nuclear Systems – Water Reactors, pp. 166-179. August 10–14, 2003, 

Stevenson, Washington. American Nuclear Society, LaGrange Park, Illinois. 

Wilkowski G, G Kramer, P Veith, R Francini, and P Scott. 1994. “The Effect of Dynamic and Cyclic 

Loading During Ductile Tearing on Circumferentially Cracked Pipe: Analytical Results.” In Fatigue, 

Flaw Evaluation and Leak Before Break Assessments, PVP–Vol. 280, pp. 221–239. American Society of 

Mechanical Engineers, New York.