PIPING BASE CASE RESULTS OF DAVID HARRIS

Probabilistic Fracture Mechanics Analyses
Performed in Support of LOCA Frequency
Re-evaluation Effort

D. O. Harris

Engineering Mechanics Technology, Inc.
San Jose, California

F. 1 Introduction

The purpose of this document is to report the procedures used and the results obtained in probabilistic fracture mechanics analyses of the base case systems considered in the LOCA Re-evaluation effort performed by use of expert elicitation by the Nuclear Regulatory Commission in the period February 2003 — March 2004. The base case systems, which were defined in the kick-off meeting of the expert panel that was held in Rockville, Maryland in February 2003, consisted of the following:

These were identified as key systems that could serve as benchmarks for use by members of the expert panel in their estimation of LOCA frequencies.

Piping isometrics of the base case systems and other systems identified in the kick-off meeting as important to estimations of flow rate probabilities were included in the FTP site that was set up for the use of panel members. Times in this appendix are in reactor-years (1 calendar year ~ 0.8 reactor years).

GENERAL APPROACH AND PHILOSOPHY OF EACH PANEL MEMBER

In this appendix the general approach and philosophy that each panelist followed as part of this elicitation exercise is presented.

K-1

BRUCE BISHOP

For PWR piping frequencies, the median probability of a 5,000 gpm (19.000 lpm) leak after 40 years of operation comes from the average point estimate for 7 plants that used the PFM methodology for the WOG Piping RI-ISI (WCAP-14257, Rev. 1-NP-A, Supplement 1). These seven plants were selected to provide a representative sampling of all plants with a Westinghouse NSSS design. Characteristics considered in the sampling included number of primary loops, old and new design vintage and foreign and domestic utility operators. The variability in 40-year probability with leak-rate comes from a WOG supported sensitivity study that reflected both the decrease in probability with increasing leak rate of one pipe size and the number of pipes of a given size that could contribute to a given leak rate. All piping leak probabilities consider the effects of LBB with a minimum detectable leak of 1 gpm (3.8 lpm) per typical plant tech-spec requirements. The increase in failure probability in going from 40 years to 60 years of operation is based upon another WOG sensitivity study. This study and its results are described in a paper presented at the 1999 Pressure Vessel and Piping Conference of ASME and included in PVP — Volume 383.

Non-Piping Frequencies are based upon the degradation mechanism of PWSCC initiation and through — wall growth, which is currently the primary cause of unexpected leaks in non-piping components in the primary system. Most other degradation mechanisms are being effectively mitigated. The relative frequencies by component type are based upon a proprietary best-estimate of PWSCC susceptibility by Westinghouse experts for unmitigated Alloy 600/182 base/weld metal. The uncertainties are based upon the variability between the best-estimate susceptibility for PWSCC and observed leak experience.

VIC CHAPMAN

In order to derive a basic set of failure probabilities, the values generated by the ‘Base Cases’ analysis were initially considered. However, in the end, a decision was made to use the results from some previous work that involved a ‘Risk-Informed ISI’ application. That work considered a full plant assessment using fatigue as the basic degradation mechanism. Initially, the results from this full plant assessment were compared with the appropriate base cases in order to ascertain whether they were in general agreement with each other. Once it was decided that the two sets of results were in agreement, it was decided to proceed with using the full plant assessment results. These results provided a set of pipe weld failures over a full range of pipe weld sizes that could be considered as a form of global values for each weld size. Factoring would then be from this base set.

Since the leak rate, given a failure, is independent of the failure probability, this can be evaluated separately to obtain a conditional probability. The basic method developed for the base case was expanded to include lower and upper estimates at each step. These basic steps are as follows:

1 Use expert judgment to estimate the COD, up to full rupture, as a function of defect size.

2 Evaluate the defect cross-sectional area for a given defect size using its associated COD.

3 Evaluate the leak rate from a given defect size using some data supplied by the USNRC.

4 Use expert judgment to assess the distribution of the defect length at failure.

5 Combine Steps 3 and 4 to obtain the conditional probability of a leak rate greater than the prescribed leak rate.

The final probabilities were obtained by combining the conditional probability above with the basic fatigue failure probability.

The effect of leak detection was introduced via a factor that was a function of the leak rate. This reduction factor varied from about 5 for a Category 1 leak, up to about 50 for a Category 6 leak.

For non-weld areas, such as the pump bowls and nozzle crotch corners, the basic probabilities were first factored. Next, the basic steps to derive the conditional leak rates as discussed above were followed to adjust the distributions as appropriate.

The effect of PWSCC was introduced as a multiplying factor on the basic fatigue failure rates. It was assumed that for small pipes, 2 inch diameter, that they would still have a significant contribution from fatigue, but that for the largest pipes, the full three orders of magnitude implied by the PWR-1 Base Case (i. e., hot leg base case) should be applied.

Finally, the failure rates for each system were derived by simply summing over all the elements within a given system.

When estimating RR-PRODIGAL leak rates through the final through wall defect in a pipe weld, evaluations were made using an elastic crack opening displacement (COD) analysis. However, it was felt that the uncertainties associated with assessing both the defect length around the pipe circumference as well as the COD needed for estimating the flow rate through the crack, were too great and too subject to ongoing development, to allow a suitable analysis of the leak rate. Thus, RR — PRODIGAL does not contain, within itself, a routine for evaluating the flow rate from the final defect size.

Instead, it was concluded that the leak rate from a through wall defect could be considered independently of the probability of the breach, i. e. the leak rate from the defect is not dependent on the probability of the defect cracking through the pipe wall. Note, however, that the COD, crack length, and hence leak rate is not independent of the mechanism that led to the failure, only the probability of the failure itself.

Within the Naval Nuclear program, computer programs have been developed to assess the leak rate from different defects based primarily on the ‘SQUIRT’ model. However, for consistency within this program, the data on leak rate against defect area provided by the USNRC were used, as shown in Figure G.1.

G. 3 Procedure

The procedures used to develop the base case numbers are as follows:

1 Evaluate the basic fatigue failure probability using RR-PRODIGAL code using the transient data supplied[16].

2 Evaluate an elastic COD as a function of defect size.

3 Use expert judgement to extend this COD beyond the elastic limit.

4 Evaluate a mean defect cross-sectional area for a given defect size using its associated COD.

5 Evaluate the mean leak rate from a given defect size using the data supplied by the USNRC, see Figure G.1.

[Note for Steps 2, 3, 4 and 5 above a defect length is given. Thus, Steps 2, 3, 4 and 5 provide a

mapping from a given defect size at failure to the mean leak rate in gpm, given this defect exists.]

6 Use expert judgement to assess the distribution of the defect length at failure.

7 Combine Steps 5 and 6 to obtain the conditional probability of a leak rate greater than the given leak rates for Categories 1 through 6. These categories being as follows;

Table G.1 Leak Category Leak Rates

8 Combining the conditional probability of Step 7 with the basic fatigue failure probability in Step 1 gives the required final probability of a leak greater than each of the categories.

Figures L.2 and L.3 show the effect of the industry and regulatory safety culture, respectively, on the LOCA Ratio (i. e., the ratio of the LOCA frequency in the future to the LOCA frequency at 25 years) for Category 1 LOCAs. Figures L.4 and L.5 show the effect of industry and regulatory safety culture on the

LOCA Ratio for Category 4 LOCAs. Ratios less than 1.0 are indicative of a perceived reduction in the LOCA frequency as a result of improvements in the safety culture mindset. As can be seen in these figures, the panel members overwhelming expected the safety culture to either improve or remain constant over the next ten to fifteen years and beyond. The panel felt that the industry as whole was acting in a consistent manner. However, a few plants with a less diligent safety culture mindset would provide the greatest challenge from a LOCA perspective. It was thought that these outlier plants may not affect the mean trends, but could strongly influence the bounds. The Davis-Besse experience was frequently cited as an example of this effect. The panel also expressed the opinion that the industry and regulatory safety culture are highly positive correlated. Therefore, regulatory and industry changes are expected to occur virtually simultaneous.

Figure L.2 Effect of Utility Safety Culture on Category 1 LOCAs

LOCA Ratio: Future/Current

Figure L.3 Effect of Regulatory Safety Culture on Category 1 LOCAs

LOCA Ratio: Future/Current

Figure L.4 Effect of Utility Safety Culture on Category 4 LOCAs

L-4

LOCA Ratios: Future/Current

Figure L.5 Effect of Regulatory Safety Culture on Category 4 LOCAs

As can be seen in comparing Figures L.2 with L.4 and L.3 with L.5, the panel members felt that any improvements in safety culture would be more beneficial for the smaller LOCA categories than their larger counterparts because the smaller LOCA categories constitute the bulk of the experience base. The frequency of the larger LOCA categories due to safety culture effects is expected to remain relatively constant over time.

The bottom line from this discussion is that because the panel members felt that the effect of safety culture was relatively minor, the LOCA frequencies developed during this exercise were not modified to account for this effect. The main caveat to this general conclusion is the previously mentioned concern that the LOCA frequencies developed through the elicitation process could be significantly degraded by a safety-deficient plant operating philosophy. The other concern frequently expressed was that the industry safety culture mindset may deteriorate near the end of a plant’s license as management tries to “squeeze out” the final few years of operations without investing in the necessary maintenance activities. Also, near the end of the plant’s license there was a concern expressed that the morale of the plant’s operating staff may begin to erode as they foresee a potential loss of employment. These concerns are manifested in the higher LOCA Ratios for the 60/25 year results when compared with the 40/25 year results in Figures L.2 through L.5.

Discussion: Rob Tregoning commented that the panel needed to define baseline LOCA frequencies in order to benchmark relative responses during the elicitation. He also mentioned that the SKI-PIPE database could be used to develop baseline frequencies if the group could develop well-defined “base case(s)”. The base case(s) will represent a set of conditions and physical phenomena. In theory, the absolute LOCA frequencies associated with each base case are not important for the elicitation session because all elicitation responses will be judged relative to the base case conditions. The absolute frequencies are only required to reconstruct the final results. However, the panel members decided that their elicitation responses might change depending on the exact LOCA frequencies associated with the base case conditions. That is, if a base case frequency was 10-8/year, the elicitation responses might be quite different than if the frequency was 10-2/year. The group therefore agreed that they will define rigorous conditions for each base case and also associate absolute LOCA frequencies with these conditions.

For this analysis all of the crack growth is from a PTS event; not fatigue.

For plants with multiple pass cladding there is a very low probability of flaw penetrating multiple passes; exception was Oconne that was single pass cladding.

Big driver were flaws between plate and axial flaw region.

For those that want to anchor against PTS, Rob will use updated results base on some average values for Oconee, Beaver Valley, and Palasdies

FOREWORD

Estimated frequencies of loss-of-coolant accidents (LOCAs; i. e., pipe ruptures as a function of break size) are used in a variety of regulatory applications, including probabilistic risk assessment (PRA) of nuclear power plants. Currently, the U. S. Nuclear Regulatory Commission (NRC) is using such information to establish a risk-informed alternative to the emergency core cooling system (ECCS) requirements in Title 10, Section 50.46, of the Code of Federal Regulations (10 CFR 50.46). Current requirements consider pipe breaks in the reactor coolant pressure boundary, up to and including breaks equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system. One aspect of this risk-informing activity is to evaluate the technical adequacy of redefining the design-basis break size (the largest pipe break to which 10 CFR 50.46 applies) to a smaller size that is consistent with updated estimates of pipe break frequencies.

To provide the technical basis for a risk-informed definition of the design-basis break size, this study developed LOCA frequency estimates using an expert elicitation process. This process consolidated operating experience and insights from probabilistic fracture mechanics studies with knowledge of plant design, operation, and material performance. Expert elicitation is a well-recognized technique for quantifying phenomenological knowledge when modeling approaches or data are insufficient.

The results from the expert elicitation provide LOCA frequency estimates for piping and non-piping passive systems, as a function of effective break size and operating time through the end of the plant license — renewal period, for both boiling — and pressurized-water reactors (BWRs and PWRs, respectively). The panelists generally agreed on the important technical issues and LOCA-contributing factors. However, as expected, the panelists’ estimates exhibit both significant uncertainty and diversity. The uncertainty is reflected in the estimated parameters (mean, median, 5th and 95th percentiles) of the individual LOCA frequency distributions, and the diversity is captured by the confidence bounds on the group estimates. In addition, this study considered the sensitivity of the results to various analysis approaches. The results are most sensitive to the method used to aggregate the individual panelists’ estimates to obtain group estimates. In this study, geometric-mean aggregation produces group frequency estimates that approximate the medians of the panelists’ estimates and are also generally consistent with both operating experience and prior LOCA frequency estimates except where increases are supported by specific material aging-related concerns. However, arithmetic-mean and mixture-distribution aggregation are alternative methods that lead to significantly higher mean and 95th percentile group estimates.

Because the alternative aggregation methods can lead to significantly different results, a particular set of LOCA frequency estimates is not recommended for all risk-informed applications. The purposes and context of the application must be considered when determining the appropriateness of any set of elicitation results. In particular, during the selection of the BWR and PWR transition break sizes for the proposed 10 CFR50.46a rulemaking, the NRC staff considered the totality of the results from the sensitivity studies, rather than only the summary frequency estimates from this study. The NRC anticipates that a similar approach will be used in selecting appropriate replacement frequencies for the estimates provided in NUREG/CR-5750, “Rates of Initiating Events at U. S. Nuclear Power Plants: 1987 — 1995,” and for other applications that require frequencies for break sizes other than those in NUREG/CR-5750.

Jennifer L. Uhle, Director

Division of Engineering

Office of Nuclear Regulatory Research

U. S. Nuclear Regulatory Commission

By Bill Galyean, INEEL

Bill employed a “top down” approach in his analysis of operating experience. His database represents approximately 2,600 LWR-years of operating experience. The resultant average age of a plant is 23 years. In that 2,600 years of LWR operating experience there have been no passive system LOCAs with a resultant leak rate greater than 100 gpm (380 lpm) (Category 1 LOCAs).

As part of this analysis, Bill assumed that cracks and leak events are indicators of LOCA frequencies. They indicate system susceptibility. In order to get a LOCA, Bill’s analysis assumed that a piping system must first have a leak or a crack. In the first 2,647 years of US LWR experience represented in Bill’s database there have been approximately 1,100 crack and leak events, but no LOCAs. Note, at the first elicitation meeting the demarcation between leaks and breaks was set at a 100 gpm (380 lpm).

A comment was made that small pipes are more susceptible to LOCAs than large pipes, i. e., large pipes are less likely to fail catastrophically. A question was raised as to why limit the analysis to US operating experience only. Bill did not categorically know whether there have been any 100 gpm (380 lpm)

LOCAs worldwide. Another reason to limit his analysis to US operating experience is that there are some fundamental design differences between US and other overseas plants. Pete Riccardella (Structural Integrity Associates) thought that if Bill had included foreign experience that the number of years of operating experience would have about doubled so Bill’s LWR LOCA frequency number of 1.9E-04/year be would reduced by a factor of two to approximately 1.0E-04/year. It is important to understand the basis for this 1.9E-04/year number since everything else is referenced to this number. This number is the total number of LOCAs of all sizes.

It was noted that this is a different approach than followed in NUREG/CR-5750. This analysis was not an attempt to update NUREG/CR-5750. Pete Riccardella asked if this database included all of the small diameter socket weld cracks that occur due to vibration fatigue. Bill indicated that this was the case, even though these small diameter lines could not result in a 100 gpm (380 lpm) leak. It was pointed out that there was no distinction between cracks and leaks in Bill’s analysis. Any crack deeper than 10 percent of the wall thickness was included in the analysis.

Gery Wilkowski (Emc2) asked if the analysis of the feedwater system (BWR-2) included FAC as a failure mechanism. Gery noted that on the secondary side there have been large breaks in some piping systems due to FAC. Bill indicated that he limited his database search to those systems that affected reactor coolant pressure boundary integrity.

Karen Gott was surprised at the low number of incidences for the feedwater system. Most of the problems seen to date with the feedwater systems have been outside the primary portion of system. Furthermore, cracks in nozzles are associated with the RPV and not piping. For the PWR systems there has been a lot of feedwater cracking, but those cracks have been on the secondary side. It was also pointed out to the panel that everyone has access to the SLAP database that Bill used for his analysis. It is now on the ftp site. Rob encouraged everyone to use it as part of their elicitation exercises. The SLAP database is current up to the end of 1998.

A question was raised about the validity of the single IGSCC failure reported in the carbon steel feedwater system. Bill indicated that this was the reported database value. Karen Gott said they’ve seen such cracking in Sweden as well.

The BWR recirculation system provided a unique problem for the base case analysis. From a materials standpoint, the base case was the old system. Bill segregated the data by old pipe (Type 304 stainless) versus new pipe (Type 316 nuclear grade [NG]). For the old pipe (Type 304 stainless) there were 127 events in 550 years of operating experience versus 3 events in 410 years of operating experience for the new pipe (Type 316NG).

The resultant leak/crack frequency for the old pipe (127 events/550 years = 0.231 events per year) is about a factor of 2 greater than the overall leak/crack frequency for the overall recirculation system history (old plus new), i. e., 130 events/960 years = 0.135 events per year. It was pointed out though that this improvement may be more due to other factors than pipe replacement only. The improvement could also be due to changes in water chemistry, or the installation of weld overlay repairs. Hence, it may be more appropriate to refer to the pipe systems as mitigated (new) or unmitigated (old) pipe systems.

It was stated that the base case is unrealistic in that it is for the old pipe case (Type 304 stainless) and no one uses that material anymore. Also, the base case does not account for the incorporation of water chemistry improvements which all plants have already implemented.

As part of his analysis, Bill made an assumption that the LOCA categories/sizes (e. g., 100, 1500, 5000 gpm, etc. [380, 5,700, 19,000, lpm, etc.]) are related on a logarithmic sense (1, 0.3, 0.1, 0.03, 0.01, 0.003). Half likelihood on logarithmic sense realizing that larger LOCAs are a subset of Category 1 (100 gpm [380 lpm]) LOCAs.

The 40 welds for the PWR-1 case (hot leg) include the cold leg and cross over leg welds. This is inconsistent with the assumption stated above that the PWR-1 case only considers the hot leg (not the cold leg or cross over leg). (Note, there are typically 3 loops in a PWR plant and there can be 5 to 7 welds per loop, but the loading is not the same for all these welds.)

Rob Tregoning indicated that the correlation between pipe size and LOCA size (gpm) that were originally supplied are subject to change.

Bill’s aging correction factor is for thermal fatigue and should not be used for the other failure mechanisms.

The non-pipe LOCAs that Bill included are for passive systems only (bolted flanges, etc.). He did not include active system contributions, e. g., stuck open valves, to the non-pipe break frequencies.

Bill’s results for the “Current Estimate” are significantly smaller than the LOCA frequencies reported by others (WASH-1400, NUREG-1150, NUREG/CR-5750) due to the larger database (more years of operating experience without a LOCA). The best agreement is with the NUREG/CR-5750 results. Bill employed a Bayesian approach as part of his LOCA frequency analysis by assuming a half of failure for these very low occurrence events. This is a very common data analysis practice. Since Bill’s analysis is based on the analysis of past passive-system failure data, the data implicitly include ISI and other mitigation experience implemented by industry.

Each plant has a PRA which includes LOCA frequency estimates for the plant. These LOCA frequency estimates are often based on WASH-1400 or NUREG-1150 and the ranges shown on pages 36 and 37 of Bill’s presentation are the ranges for the IPEs. Thus, the IPE range, WASH-1400 and NUREG-1150 frequencies shown by Bill on pages 36 and 37 of his handout are closely related/interlinked. As described in Slides 36 and 37, Bill expressed the uncertainty in his analysis by assuming an error factor of 10. This is somewhat crude and somewhat arbitrary, but the scope of the base case analysis did not ask for uncertainty, just a best estimate.

Bill indicated that he has no strong technical basis for extending his “Current Estimate” analysis out to 40 or 60 years. With Bill’s approach, the LOCA frequencies will go down as more years of operating experience are accumulated, unless a LOCA event occurs in the future. Bill presented information predicting 8 years to double the frequencies for thermal-fatigue events. This estimate conservatively assumes no industry-wide mitigation programs. Rob Tregoning indicated that he did not include results for Bill in the summary table for 40 and 60 years of operation because they are most appropriate for current estimates.

Bruce Bishop was very uncomfortable predicting the future out to 40 or 60 years. He felt that the uncertainties are going to increase dramatically. Lee Abramson (USNRC) responded that this is a natural experience and likely shared by others on the panel. However, it’s incumbent that each member to attempt these predictions. Rob and Lee stressed again that the panel will not be forced to answer any questions that they are very uncomfortable with. Rob further stressed that the March 2003 SRM specified that the NRC staff needs to revisit the LOCA frequency estimates every 10 years and the effort will be most concerned with the next 10 years. However, it is still important to gain longer-term insights.

A question was raised as to whether or not to have Bill extend his analysis out to 40 and 60 years? Gery Wilkowski said, yes, he wanted to see Bill’s assumptions. Fred Simonen (PNNL) would like to see more partitioning by pipe size as part of Bill’s analysis. Dave Harris (Engineering Mechanics Technology) concurred. Sam Ranganath (formerly of General Electric) specifically indicated that it would be helpful for the BWR recirculation system because they replaced the smaller diameter pipes, but not the 28-inch diameter pipes. Bengt Lydell is to check on the statistics in his database to see if any of the 28-inch diameter BWR recirculation lines had leaks and provide this information in his final base case report.

Pete Riccardella and Sam Ranganath were unaware of any General Electric large diameter recirculation pipes that leaked.

Plant B is a BWR/4 assumed to have been in commercial operation for at least 25 years. Similar to many other operating BWR/4 plants in the USA, Plant B is also assumed to be operating with a combination of IGSCC Category D and E welds, according to the nomenclature of U. S. NRC Generic Letter 88-01 [D.3, D.4]. In other words, the plant has experienced some IGSCC and the affected welds have been reinforced by weld overlays. It is further assumed that none of the IGSCC susceptible welds have been subjected to any stress improvement (SI) process such as induction heat stress improvement (IHSI) or mechanical stress improvement process (MSIP). It is also assumed that the weld overlay repairs (WOR) were all performed in the 1982-1988 timeframe. Finally, Plant B is assumed to have been operating with normal water chemistry (NWC) at all time.

The system descriptions in this section are extracted from design information supplied by members of the Expert Elicitation Panel. The BWR-specific system information is included in the following documents and drawings:

• Document No. EPRI-156-310: Degradation Mechanisms Evaluation for Class 1 Piping Welds at Plant B [D.5].

• Excel-file entitled “PlantBWelds.” This Excel-file includes weld lists with locations for the RR and FW ASME Section XI Code Class 1 piping. The lists are organized by weld identification numbers (as they appear on the isometric drawings identified below) nominal pipe size and pipe schedule. The Excel file forms the basis for the LOCA frequency model used to derive the LOCA frequency distributions.

• Isometric drawing numbers 6M721-5358-5 (RR System Loop B Ring Header), 6M721-5359-5 (RR Loop B Suction & Discharge Piping), 6M721-2336-1 (FW System Inside Drywell), and 6M721- 3537-5 (FW System Inside Drywell).

D.1.3.1 Reactor Recirculation (RR) System — The RR System evaluated in this study consists of two recirculation pump loops external to the reactor pressure vessel (RPV). These loops provide the piping path for the driving flow of water to the RPV jet pumps. Each loop contains a variable speed recirculation pump and two motor operated isolation valves (one on each side of each pump). The recirculation loops are part of the nuclear system process barrier and are located inside the drywell containment structure. The pipe segments that are subject to evaluation in this study consist of:

Loop A: The Class 1 portion starts at the RPV nozzle N1A and is reconnected to the RPV at nozzles

N2F, N2G, N2H, N2J, and N2K. Class 1 lines for the Residual Heat Removal (RHR) and Reactor Water Cleanup (RWCU) Systems are connected to this loop. These particular Class 1 lines are excluded from the study scope, however. Loop A is excluded from the BWR Base Case.

Loop B: The Class 1 portion starts at RPV nozzle N1B and is reconnected to the RPV at nozzles

N2A, N2B, N2C, N2D, and N2E. Part the original design, a NPS4 bypass line at valve F031B has been removed from the system. Class 1 lines for the RHR and RWCU Systems are connected to this loop. These particular Class 1 lines are excluded from the study scope, however.

D. 1.3.2 Feedwater (FW) System — The FW System provides feedwater to maintain a pre-established water level in the RPV during normal plant operation. The Condensate and the FW Systems take water from the main condenser and deliver it to the RPV after passing it through the feedwater heaters and demineralizer system. The Class 1 portion of the FW System consists of two loops:

Loop A: Loop A starts at valve F076A and a connection to the High Pressure Coolant Injection

(HPCI) discharge line (at valve F006), and connects to the RPV at nozzles N4A, N4B, and N4C. The HPCI discharge line is excluded from the study scope. Loop A is excluded from the BWR Base Case.

Loop B: Loop B starts at valve F076B, connection to the Reactor Core Isolation Cooling (RCIC)

discharge line at valve F013, and a discharge from the RWCU System (at valve F220), and connects to the RPV at nozzles N4D, N4E, and N4F. The RCIC and RWCU discharge lines are excluded from the study scope.

Pete Riccardella received his PhD. from Carnegie Mellon University in 1973 and is an expert in the area of structural integrity of nuclear power plant components. He co-founded Structural Integrity Associates (SIA) in 1983, and has contributed to the diagnosis and correction of several critical industry problems, including:

• Feedwater nozzle cracking in BWRs

• Stress corrosion cracking in BWR piping and internals

• Irradiation embrittlement of nuclear reactor vessels

• Primary water stress corrosion cracking in PWRs

• Turbine-generator cracking and failures.

Dr. Riccardella has been principal investigator for a number of EPRI projects that led to advancements and cost savings for the industry. These include the FatiguePro fatigue monitoring system, the RRingLife software for turbine-generator retaining ring evaluation, RI-ISI methodology for nuclear power plants, and several PFM applications to plant cracking issues. He has led major failure analysis efforts on electric utility equipment ranging from transmission towers to turbine-generator components and has testified as an expert witness in litigation related to such failures.

He has also been a prime mover on the ASME Nuclear ISI Code in the development of evaluation procedures and acceptance standards for flaws detected during inspections. In 2002 he became an honorary member of the ASME Section XI Subcommittee on ISI, after serving for over twenty years as a member of that committee.

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