## Summary of Input Data and Results

Tables D.22 and D.23 summarize the input data to the LOCA frequency calculation. Tables D.24 through D.26 give the results at T = 25, 40 and 60 years, respectively. Consistent with the LOCA frequency elicitation structure, Table D.27 is summary of mid values (MV, or 50th percentiles) at T = 25 years rather than mean values, however. Figure D.41 shows the time-dependent LOCA frequencies. Figure D.42 shows selected weld failure rates. Figures D.43 through D.46 show the contribution to LOCA frequency by respective Base Case. Note that Figure D.45 includes the contribution to LOCA frequency by PWR-1 (Reactor Coolant System Hot Legs; all 3 loops are accounted for in this figure) and PWR-2 (Pressurizer Surge Line). Note that the Base Case results used in Table 4.1 in the main body can be obtained from Tables D.16, D.17, and D.20 in this report.

Table D.22 Summary of Key BWR Base Case Input Data

 Base Case Input BWR-1 BWR-2 Data NPS12 NPS22 NPS28 NPS12 NPS14 NPS20 Weld count 50 16 56 63 5 53 Weld failure rate Dominant [1/Reactor-yr.] 6.50E-05 1.54E-04 1.44E-04 2.20E-06 2.20E-06 1.58E-06 Weld failure rate Minimum [1/Reactor-yr.] 2.37E-05 3.32E-05 1.29E-05 1.77E-07 1.77E-07 1.73E-07

Table D.23 Summary of Key PWR Base Case Input Data

 Input Data Base Case PWR-1 PWR-2 PWR-3 NPS30 NPS14 NPS3-% Weld count 50 14 9 Weld failure rate Dominant [1/Reactor-yr.] 7.64E-05 1.56E-06 6.56E-04 Weld failure rate Minimum [1/Reactor-yr.] 1.05E-06 4.60E-08 1.58E-06

Table D.24 Calculated LOCA Frequencies (T = 25 Years)

 LOCA Frequency — Statistical Mean [per Reactor-year] Base Flow Rate Threshold Value [gpm Case Catl Cat2 Cat3 Cat4 Cat5 Cat6 V > 100 V > 1,500 V > 5,000 V > 25,000 V > 100,000 V > 500,000 BWR-113 9.46E-06 1.22E-06 4.60E-07 1.53E-07 3.05E-08 N/A14 BWR-215 2.54E-06 3.36E-07 1.25E-07 4.09E-08 7.33E-09 N/A PWR-116 7.42E-07 7.62E-08 2.93E-08 1.09E-08 3.77E-09 1.26E-09 PWR-2 1.29E-07 1.50E-08 5.40E-09 1.56E-09 5.31E-10 N/A PWR-317 1.60E-05 2.32E-06 9.22E-07 N/A N/A N/A

Table D.25 Calculated LOCA Frequencies (T = 40 Years)

 Base Case LOCA Frequency — Statistical Mean [per Reactor-year] Flow Rate Threshold Value [gpm Catl V > 100 Cat2 V > 1,500 Cat3 V > 5,000 Cat4 V > 25,000 Cat5 V > 100,000 Cat6 V > 500,000 BWR-1 1.14E-05 1.47E-06 5.54E-07 1.84E-07 3.78E-08 N/A BWR-2 2.56E-06 3.39E-07 1.26E-07 4.13E-08 7.40E-09 N/A PWR-1 8.96E-07 9.20E-08 3.54E-08 1.32E-08 4.55E-09 1.45E-09 PWR-2 1.60E-07 1.86E-08 6.70E-09 1.93E-09 6.59E-10 N/A PWR-3 1.95E-05 3.30E-06 9.44E-07 N/A N/A N/A

Table D.26 Calculated LOCA Frequencies (T = 60 Years)

 Base Case LOCA Frequency — Statistical Mean [per Reactor-year] Flow Rate Threshold Value [gpm Cat1 V > 100 Cat2 V > 1,500 Cat3 V > 5,000 Cat4 V > 25,000 Cat5 V > 100,000 Cat6 V > 500,000 BWR-1 1.88E-05 2.43E-06 9.16E-07 3.05E-07 6.07E-08 N/A BWR-2 2.56E-06 3.39E-07 1.26E-07 4.13E-08 7.40E-09 N/A PWR-1 9.74E-07 1.00E-07 3.85E-08 1.43E-08 4.95E-09 1.57E-09 PWR-2 1.77E-07 2.06E-08 7.41E-09 2.14E-09 7.29E-10 N/A PWR-3 1.96E-05 3.32E-06 9.50E-07 N/A N/A N/A

13 BWR-1 is the combination of RR Loop A and B.

14 N/A = not applicable.

15 The results are for FW Loop A and B.

16 The results are for 3-of-3 RC hot legs.

17 The results are for 2-of-2 HPI/NMU lines.

 Base Case Median (MV) LOCA Frequency [per Reactor-year] Flow Rate Threshold Value [gpm Cat1 V > 100 Cat2 V > 1,500 Cat3 V > 5,000 Cat4 V > 25,000 Cat5 V > 100,000 Cat6 V > 500,000 BWR-1 8.23E-06 1.08E-06 4.03E-07 1.29E-07 2.19E-08 N/A BWR-2 1.09E-06 1.35E-07 5.03E-08 1.65E-08 2.10E-09 N/A PWR-1 1.54E-07 2.25E-08 8.33E-09 2.85E-09 8.53E-10 1.58E-10 PWR-2 1.37E-08 1.39E-09 5.15E-10 1.54E-10 5.46E-11 N/A PWR-3 6.87E-06 1.15E-06 2.14E-07 N/A N/A N/A

 0 10 20 30 40 50 60 70 Years

Figure D.41 Time-Dependent Cat 1 LOCA Frequencies

 RR NPS28 Pipe-to — RR NPS28 Elbow-to — RR NPS12 Pipe-to — RC-HL NPS30 Nozzle — RC-HL NPS30 Elbow — RC Surge Line HPI/NMU (NPS 3-3/4) Safe-end Pipe Reducer to-Safe-end to-Pump (NPS14) Nozzle-to — Pipe-to-Nozzle Safe- Safe-end end

 0.0001

 1e-05

 1e-06

Figures D.48 (BWR) and D.49 (PWR) show the influence of in-service inspection on the time-dependent LOCA frequency; no ISI and ISI with POD = 0.5 and 0.9, respectively.

Figure D.48 Influence of ISI on Time-Dependent BWR-1 Cat 1 LOCA Frequency

Figure D.49 Influence of ISI on Time-Dependent PWR-1 Cat 1 LOCA Frequency

## REACTOR VESSEL LOCA PROBABILITY BASE CASE ANALYSES. (BWR VESSELS AND PWR TOP HEAD NOZZLES)

I.1 Introduction

The LOCA expert panel elicitation team charter includes estimating the contribution to LOCA frequency from reactor vessels and other non-piping components. Extensive analyses were performed by members of the elicitation panel to develop LOCA frequencies for five piping “base cases” that were formulated by the panel in early meetings (documented as Appendices D, E, F and G to this NUREG). The piping base cases include failures on the piping side of vessel nozzles, including safe-ends. However, they do not include small diameter, partial penetration welded nozzles such as CRDM penetrations and other small nozzles, such as instrument nozzles, that aren’t connected to piping systems. In addition, the piping base cases do not include consideration of a leak from or rupture of other regions of the reactor vessel, such as the irradiated reactor vessel beltline or the low alloy steel portions of large vessel nozzles. LOCA frequency estimates for these cases are presented in this appendix, based on prior PFM analyses performed for PWR top head nozzles [I.1, I.2], the BWR Reactor Vessel Beltline Region [I.3, I.4], and BWR reactor vessel feedwater nozzles [I.5]. These estimates are used to construct a complete set of LOCA frequency tables for BWR and PWR reactor vessels, for all LOCA categories defined in the elicitation, a comparison of them to the aforementioned piping base cases is also presented.

I.2 PWR Reactor Vessel Top Head Nozzles

Extensive PFM analyses have been conducted over the past several years to estimate the probability of leakage and rupture associated with the PWR CRDM penetration PWSCC problem [I.1, I.2]. The analysis model incorporates the following major elements:

• computation of applied stress intensity factors for circumferential cracks in various nozzle geometries as a function of crack length,

• determination of critical circumferential flaw sizes for nozzle failure,

• an empirical (Weibull) analysis of the probability of nozzle cracking or leakage as a function of operating time and temperature of the RPV head,

• statistical analysis of PWSCC crack growth rates in the PWR primary water environment as a function of applied stress intensity factor and service temperature, and

• modeling of the effects of inspections, including inspection type, frequency and effectiveness. The model has been benchmarked with respect to field experience, considering the occurrence of cracking and leakage and of circumferential cracks of various sizes. Figures I.1 and I.2 illustrate the benchmarking. Figure I.1 presents a Weibull analysis of inspection results at thirty plants, of which 14 detected leakage or cracking (data points in the figure). The remaining plants that were inspected and found clean were treated as “suspended tests” according to standard Weibull analysis theory [I.2]. The data are plotted in terms of effective degradation years (EDYs) which are equivalent operating years at 600°F (315°C), using an activation energy (Arrhenius) model [I.1] to adjust for different head operating temperatures. For plants in which multiple cracked nozzles were detected in the inspections, the data were extrapolated back to the expected time of first cracking or leakage, using an assumed Weibull slope of 3. The straight line through the data represents a medium rank Weibull regression (also with a slope of 3) upon which the probability of leakage predictions in the model are based. Figure I.2 illustrates the benchmarking process used for the crack growth analysis algorithm in the model with respect to CRDM nozzles that exhibited circumferential cracks of various sizes. (Eleven (11) nozzles out of a total of 881 inspected nondestructively through the spring of 2003 exhibited circumferential cracking. No additional

circumferential cracking has been detected in more recent inspections.) The figure shows that, when using original analysis parameters, the crack growth model under-predicted the probability of circumferential cracking somewhat, but after adjustment of selected analytical parameters, the PFM model was “benchmarked” so as to very accurately predict the field results, especially for the most important, larger crack sizes.

The benchmarked model was then used to evaluate the probabilities of nozzle failure and leakage in actual plants. A sample of the results is presented in Figures I.3 and I.4. Figure I.3 illustrates the probability of nozzle failure (ejection of a nozzle) for a head operating temperature of 580°F (304°C), the approximate average of U. S. PWRs. No inspections were assumed to be performed during the first 25 years of plant operation, resulting in the probability of nozzle failure constantly increasing with time during that period. The analysis then assumed that inspections begin after 25 years, at intervals and detection levels representative of current requirements [1.6]. It is seen from the figure that the current inspection regimen reduces the nozzle failure probability significantly.

Ejection of a 4 inch CRDM nozzle [2.75 inch (~70 mm) ID] due to a circumferential crack would yield a one-sided LOCA corresponding approximately to Category 2 LOCA [>1,500 gpm (5,700 lpm) but <

5,0 gpm (19,000 lpm)]. If periodic inspections are continued, with any nozzles in which leakage or cracking are detected repaired or the heads replaced (as is common practice), the nozzle ejection probability will be even lower in the future. Table I.1 below provides a summary of the average failure probabilities from Figure I.3, between 0 and 25 years, and from 25 to 40 years. The probability of failure for 40 to 60 years was not calculated, but was assumed to be the same as 25 to 40 years, on the basis that the current inspection regimen will be maintained, or the heads replaced. A Category 3 break was assumed to require multiple nozzle failures, the probability of which was computed via a binomial distribution for the typical number of nozzles in a head. As seen in Table I.1, the probabilities of simultaneous multiple nozzle failures is quite low.

Figure I.4 illustrates similar PFM results (based on the above Weibull model) for the probability of small amounts of leakage from a top head CRDM nozzle. The same inspection regimen was assumed as in the nozzle ejection analysis (no inspections from 0 to 25 years, inspections in accordance with current requirements thereafter). A small leak from a CRDM nozzle was assigned as a Category 0 break [less than 1 gpm (3.8 lpm)] in Table I.1, and the intermediate, Category 1 break size was obtained by logarithmic interpolation between Categories 0 and 2.

Table I.1 Summary of PWR CRDM Nozzle PFM Results

 Break Category Leak Rate >(gpm) Average LOCA Probabilities During Operating Years: 0-25 25-40 40-60 0 1 2.00E-02 5.00E-03 5.00E-03 1 100 1.27E-03 2.75E-04 2.75E-04 2 1,500 2.50E-04 5.00E-05 5.00E-05 3 5,000 4.00E-08 2.00E-09 2.00E-09 4 25,000 — — — 5 100,000 — — — 6 500,000 — — —

## Submitted by Joseph Conen of the BWR Owners Group

Comment: It is apparent that the panel has not given appropriate credit to the IGSCC mitigation measures for the NSSS stainless steel piping that the BWR owners have implemented since the early 1980s. For example, the second paragraph from bottom on page xvii states, in part: “.. .the biggest frequency contributors for each LOCA size tend to be systems having the smallest pipes, or component, which can lead to that size LOCA. The exception to this general rule is the BWR recirculation system, which is important at all LOCA sizes due to lingering IGSCC concerns.” Since the largest pipe size in the recirculation piping system can be up to 28-inches, the preceding statement essentially implies that LB LOCA redefinition is not applicable to BWRs. The panel did not seem to give adequate credit for several effective mitigation measures in terms of better material (e. g., use of nuclear grade stainless steel in replacement lines), stress improvement (e. g., induction heating stress improvement [IHSI], last pass heat sink welding [LPHSW], and mechanical stress improvement process [MSIP]) and water environment (e. g., hydrogen water chemistry [HWC]) and repair measures such as the weld overlays and elimination of creviced geometries. The panel apparently did not consider the report GE-NEA41-00110-00-1, Rev. 0, A Review of NUREG/CR-5750 IGSCC Improvement Factor and Probability of Rupture Given a Through — Wall Crack, provided to NRC by the BWROG on April 25, 2002 (ADAMS Accession NO. ML021210417). In addition, the panel did not recognize the contribution of BWRVIP-75, which provides evidence that IGSCC is effectively managed at BWRs and provides the basis for revising examination frequencies. We consider these significant oversights, given their relevance to the panel’s conclusions. On the other hand, the panel did accept the future effectiveness of mitigating measures for PWSCC issue for the PWR small diameter piping (p. 6-5) in reducing failure rates for this piping. The NUREG should provide similar credit for the BWR IGSCC mitigation measures noted above with regard to break frequencies.

Response: The authors disagree with the contention that the panel has not given appropriate credit to the IGSCC mitigation measures for the NSSS stainless steel piping that the BWR plants have implemented since the early 1980s. The report referenced above (GE-NEA41-00110-00-1) was provided as background documentation to the peer review panel. Additionally, several of the panelists had extensive experience with the assessment of the IGSCC issue and the development of appropriate mitigation strategies. One of the BWR base cases specifically investigated the failure probability of primary recirculation piping due to IGSCC. The base case model plant was assumed to follow the Generic Letter 88-01 inspection technique, used weld overlay to reinforce the flawed piping, and utilized normal water chemistry.

This base case definition was chosen for convenience to improve the expected accuracy of the PFM analysis to be conducted by only considering one easily modeled mitigation strategy, i. e., weld overlay. It was well recognized and stressed during meetings that this base case was generally not representative of BWR plants. It was noted that most BWR plants employ hydrogenated water chemistry and may employ mitigation strategies other than weld overlays. Additional sensitivity analyses of the base case results were conducted to evaluate the effectiveness of BWR mitigation strategies for IGSCC. These sensitivity studies are summarized in Sections 4.3.3 and 4.3.4 of NUREG-1829, while more detail is provided in Appendices D — F. These results were presented to the panelists during the base case review meeting and copies were available to support the preparation of their individual elicitations. This information, the background information provided, and the experience of individual panelists was sufficient to ensure that the panelists were sufficiently informed about the effectiveness of IGSCC mitigation so that it was properly credited during the elicitation.

However, although the mitigation has been effective in reducing the associated LOCA frequencies, there is still risk associated with failure of BWR systems containing pre-existing flaws. As summarized in Section 6.3.2 of NUREG-1829, “The panel consensus is that the susceptibility of BWR piping systems to IGSCC is greatly reduced compared to what it was in the past. Measures such as improved HWC, weld overlay repairs, stress relief, and pipe replacement with more crack resistant materials have led to this reduction. Inspection quality has also improved such that the probability of crack detection is much better than in the past. However, as indicated earlier, there remains concern about the failure likelihood of the large recirculation piping and the RHR lines that have not been replaced. The original piping materials are much more susceptible to IGSCC and many lines retain preexisting cracks that initiated and grew before HWC was adopted.”

Additionally, at least one panelist was also concerned that the more IGSCC-resistant replacement piping materials may still crack under service conditions. This panelist cited the German plant experience with cracking in Type 347 stainless steel. Another panelist raised the possibility that cold work (e. g. due to grinding) could increase the IGSCC susceptibility of the low carbon (L grade) stainless steel that has been used as a replacement material in many U. S. plants. However, the U. S. BWR experience with L grade stainless steel piping was widely recognized by the panel as being very good thus far. For these reasons, many panelists believe that continued vigilance is required through the augmented inspection requirements in Generic Letter 88-01 and NUREG-0313.

Key elements of this response have been used to modify the Executive Summary and Sections 3.5, 4.3, and 6.3 of the revised NUREG.

Comment Number: ES3

## Software

The following discussion provides only a brief review of the PRAISE software. The references cited give the details. The results reported here were generated by use of the PRAISE software, which was developed with NRC support over a period of some 20 years. PRAISE is based on deterministic fracture mechanics, with some of the inputs considered as random variables. This allows the statistical distribution of lifetime to be computed, rather than a single deterministic failure time. The probability of failure (leaks of various sizes) is obtained from the computed lifetime distribution.

Several versions of PRAISE were employed, depending on the nature of the problem. The original version of PRAISE [F.1] considers fatigue crack growth from crack-like weld defects introduced during

fabrication. Semi-elliptical interior surface cracks are considered, usually circumferentially oriented.

The initial crack size and fatigue crack growth properties are the major random variables, and Monte Carlo simulation is used to generate numerical results. Stratified sampling of crack depth and aspect ratio is employed to allow very small probabilities to be obtained without excessive computer time. Figure F. 1 is a schematic representation of the probabilistic fracture mechanics procedures used in the original version of PRAISE. The cumulative probability of a flow (leak) rate exceeding a specified size is generated by PRAISE as a function of time. If the stress history is specified in reactor-years, then the

Figure F.1 Overview of PRAISE Methodology for Probabilistic Analysis of Fatigue Crack Growth

A later version of PRAISE was used the analyze initiation and growth of stress corrosion cracks [F.2]. Both the fatigue crack growth and stress corrosion initiation and growth capabilities are included in pcPRAISE [F.3] and also in WinPRAISE [F.4], which is a Windows version that is much easier to use than the earlier DOS versions. WinPRAISE gives the same results as pcPRAISE for the same problem with the same inputs.

In some cases, the cyclic stresses were such that fatigue crack initiation is the expected dominant degradation mechanism, and it was necessary to use a later version of PRAISE. This version was developed and used in Reference F.5, with additional capabilities described in Reference F.6 (such as the ability to consider detailed circumferential variations of the stresses). The fatigue crack initiation analyses employed probabilistic strain-life relations developed by Argonne National Laboratory [F.7,

F. 8]. Once a fatigue crack initiates, the original growth analysis capabilities in PRAISE are used for the crack growth portion of the lifetime. The depth of an initiated fatigue crack is taken to be 3.0 mm (0.12 inches), in accordance with the ANL correlations, with a random surface length. In some instances, modifications to the source code of PRAISE were made to provide results of particular use for the problem at hand. These instances are discussed in the specific base case problem where they were employed.

 . a uln a *

In cases where inspection was considered, the nondetection probability was represented by the expression

In this expression, є is the probability of not detecting a crack no matter how deep it is, a* is the crack depth having about a 50% chance of being detected, and v controls the slope of the PND curve. “Good” and “outstanding” detection capabilities were considered [F.9], with the parameters given in Table F.1.

Table F.1 Parameters in Non-Detection Relation

 Ferritic Austenitic Fatigue Cracks Fatigue Cracks SCC Cracks outstanding a* 0.05h 0.05h 0.05h v 1.6 1.6 1.6 є 0.005 0.005 0.005 good a* 0.15h 0.15h 0.4h v 1.6 1.6 1.6 є 0.02 0.02 0.10

The operating and hydro pressures of Table F.2 were considered.

Table F.2 Operating and Hydro Pressures, psi

 Operating Pressure Hydro Pressure PWR 2250 3125 BWR 1250 1560

A leak detection capability of 19 lpm (5 gpm) was considered, which means that any through-wall crack with a leak rate greater than 19 lpm (5 gpm) was immediately detected and removed from service.

## WILLIAM (BILL) GALYEAN

The approach taken by Bill Galyean is based on the total operating experience of U. S. commercial nuclear power plants. This experience consists of approximately 2,650 LWR-years with zero category-1 (> 100 gpm [380 lpm]) loss of coolant accidents. The average age of these plants is approximately 25 years, with a number of plants being 30-plus years old. During this time, a number of RCS degradation issues have arisen and been addressed, for example, IGSCC in BWRs and thermal fatigue in PWRs. The operating experience therefore indicates that degradation will occur, but it will likely be detected and corrected before it can lead to a catastrophic failure. Consequently, this data is the basis for estimating an average LOCA frequency using a Bayesian update of a non-informative prior distribution. Since both PWRs and BWRs have zero LOCAs, the reasonable assumption is that the two designs share a similar LOCA frequency. The operating experience for the two designs is therefore pooled (i. e., use zero failures and the total 2,650 LWR-years of experience). Assuming the LOCA frequency has been (and will be) relatively constant over time (again, this seems reasonable given the history of degradation mechanisms being detected and subsequently mitigated), the resulting LOCA frequency of 1.9E-4/LWR-year produces a probability of one or more LOCA events in the 2,650 LWR years of experience of 39% (again not unreasonable, given there have been zero LOCA events). By contrast, separating the PWR and BWR experience and analyzing them separately produces LOCA frequencies of 2.8E-4/PWR-yr and 5.6E — 4/BWR-yr, and a probability of seeing one or more LOCA events (either PWR or BWR) in 2,650 LWR — years experience of 63%. Again, given that there have been no LOCA events, the first (pooled) estimate seems to be the more realistic.

This assumed LOCA frequency (1.9E-4/LWR-yr) was used for the category 1 LOCA (> 100 gpm [380 lpm]). Note that as defined in the elicitation effort, category 1 LOCA includes all larger size categories. So the approach followed by this panel member was to assume a Уг order of magnitude reduction in frequency for each next larger size category. This general approach (if not the precise value of the reduction) has been followed by virtually every LOCA frequency estimate ever made, and is supported by studies on precursor events documented in NUREG/CR-5750, Appendix J.

The time-independent assumption for the LOCA frequency is also based upon the historical experience, if only qualitatively. There seems to be no doubt that the LOCA frequency fluctuates over the age of the plant, but there is reason to believe it will both increase and decrease over time. The IGSCC experience seems to support the assertion that times of increasing frequency will be followed by times of decreasing frequency as degradation mechanisms are identified, understood, and mitigated. Indeed, even the recent RPV-head corrosion event at Davis Besse supports this model of a LOCA frequency increase as degradation occurs undetected, then a decrease as mitigation programs are implemented (e. g., in the case of Davis Besse, replacing RPV head).

The last issue to be addressed is the allocation of the total LOCA frequency among the systems and components that compose the RCS. This aspect again relies upon operating experience data, this time in the form of the relative frequency of crack and leak events (i. e., precursor events). Basically, these precursor data were collected from LER and foreign reactor experience, and then sorted by degradation mechanism and RCS subsystem/component. In many cases the information provided on the precursor event was somewhat unclear or incomplete. Also, there is little assurance that all precursor events have been captured. However, assuming there is no bias in the reporting of the events such that the data samples for each subsystem/component are equally representative of the all events for that subsystem/component, then the data can be used to support estimates of the relative contribution from each subsystem/component. That is, the precursor events do not have to be completely reported, just consistently reported. Further, the RCS subsystem/component boundaries have not been clearly defined. Hence, the relative contributions to the overall LOCA frequency would likely change somewhat if the precursor data were reviewed and categorized by a different analyst. Nevertheless, this aspect of the analysis was performed simply to allocate the total LOCA frequency (described above) to the general subsystems/components that make-up the RCS.

In summary, the entire U. S. LWR operating experience is used to estimate an average industry-wide total LOCA frequency. This frequency is used for both BWRs and PWRs, not because they are believed to be the same, but on the basis that the operating experience does not support different frequencies. Time — independence is assumed using the rationale that variation (both increases and decreases) in the frequency will occur as degradation mechanisms manifest themselves and are subsequently addressed by the industry. This total LOCA frequency is allocated by LOCA size categories using the argument that as pipe-size increases, the LOCA frequency decreases. This argument is supported by a number of studies on precursor data and if nothing else has been reflected in all LOCA frequency estimates since WASH — 1400 (1975). The total LOCA frequency is also allocated by RCS subsystem/component using data collected on primary system crack and leak events (although the details of this allocation are view as somewhat subjective with respect to the boundaries of the different subsystems/components).

## Example Base Case Analysis

As a way of demonstrating the procedure given above, the results for the 14-inch Surge Line elbow are reproduced in this section. Two situations are considered, the elbow and the adjacent weld. The transients were based on data supplied and are reproduced in Attachment G.1

G. 4.1 Probability of Failure Surge Line Elbow — Base Case

This is a failure from base material and so the analysis assumed a fatigue based crack initiation followed by crack growth to failure. As stated earlier, the crack initiation and crack growth are assumed to be positively correlated. This assumption assumes that if the properties of the base material are such as to lead to an early crack initiation, it is very possible that these same properties

could result in a subsequently fast crack growth rate. The results of this analysis are shown in the following table:

Table G.2 Results for PWR Surge Line Elbow Base Case Analysis

 Time Cumulative Probability (years) of Failure 25 6.1×10-6 40 7.8×10-6 60 9.4×10-6

RR-PRODIGAL gave the critical through wall defect length, based on the R6 criterion, as 14 inches.

## BWR Piping

The participants generally thought that the important degradation mechanisms for BWR piping were thermal fatigue, FAC, and IGSCC. It was argued that BWR plants are more prone to thermal fatigue problems than the PWRs because they experience a greater temperature fluctuation during the normal operating cycle. In BWRs, thermal fatigue is a concern for the feedwater lines, the main steam lines, and the RHR system. From a LOCA perspective, thermal fatigue is an important aging mechanism because it does not manifest itself as a single crack, but as a family of cracks over a wide area. As such, it can lead
to a large LOCA. Thermal fatigue cracks also tend to propagate rapidly, and since it is not material sensitive (i. e., it can attack a number of materials), it is difficult to prioritize critical areas for inspections.

Only the feedwater piping system is highly susceptible to FAC. The main steam line is the other major carbon-steel piping system which experiences constant fluid flow. However, it is not as susceptible to FAC because the erosion rates associated with two-phase flow are less severe. While FAC caused a serious accident in the secondary side piping at Surry 15 years ago, the panel members generally thought that the industry had inspection programs in place today to prevent the reoccurrence of such an event, especially for the primary side piping systems. However, a number of panel members expressed the concern that the water chemistry improvements which mitigate IGSCC could lead to unexpected FAC problems.

The panel consensus is that the susceptibility to IGSCC is greatly reduced compared to the past.

Measures such as improved HWC, weld overlay repairs, and pipe replacement with more crack resistant materials had essentially reduced the likelihood of IGSCC. However, there is still residual concern about the failure likelihood of the large recirculation piping material that has not been replaced. Furthermore, even for the pipe which has been replaced, the question was raised as to whether the new replaced pipe was immune to this type of degradation, or is the problem simply been move out into the future. The German experience with Type 347 stainless steel was raised in this regard. There was also concern expressed about the effects of increased sulfate levels in the future due to efforts focused at extending the life of some of the filters in the plants.

Another aging mechanism of concern is mechanical fatigue. This is primarily a problem in smaller diameter piping, especially those with socket welds, and is caused by an adjacent vibration source. From a LOCA perspective, it was noted that locations susceptible to mechanical fatigue damage were not always obvious. It is impossible to eliminate all plant vibrations, and furthermore, changing the configuration of the plant can result in newly susceptible areas.

As part of this elicitation exercise a total of 14 LOCA-susceptible piping systems were considered for the BWR plants. Of these, however, most of the participants focused on a few common systems as being the important LOCA contributors. Figure L.6 shows the Category 1 LOCA frequencies for each of these piping systems at 25 years of plant operation (present day). Note, the results for the HPCS and LPCS systems are combined as a single entry in Figure L.6 (HPCS/LPCS). For these smaller category LOCAs, the main concern is with the smaller diameter lines, such as the instrument and drain lines. Most of the participants believe that it is more likely to have a complete break of a smaller diameter line than a comparable size opening in a larger diameter pipe. One reason for this is that for a given crack size, the crack is a larger percentage of the pipe circumference in the smaller diameter pipes, and it was thought that a small diameter pipe was just as likely to have a crack of a certain length as a larger diameter pipe. Furthermore, smaller diameter lines are often fabricated from socket welded pipe which has a history of mechanical fatigue damage from plant vibrations. These lines may also be susceptible to external failure mechanisms arising from human error (e. g., damaging with equipment, such as fork trucks). Finally, these smaller diameter lines are often subject to fabrication flaws and they are typically more difficult to inspect, if they are inspected at all. In-service inspection is not routinely performed on these lines. Conversely, the larger diameter lines are inspected more rigorously and routinely.

Besides the instrument and drain lines, the recirculation and, to a slightly lesser extent, the CRD and RHR lines are also of concern, primarily as a result of SCC susceptibility.

For larger Category 3 LOCAs, the recirculation system was the largest contributor to the overall LOCA frequencies, see Figure L.7. (Note in this figure that the instrument and drain lines, as well as the CRD lines, are no longer shown in that these smaller diameter lines cannot support a Category 3 LOCA.) The fact that the recirculation system is the largest contributor is a slight departure from the PWR estimates where the smallest diameter piping system that can support a particular LOCA category consistently had the highest LOCA frequencies. The main concern with the recirculation system piping continues to be SCC, even when considering the effective mitigation programs in place today. Of secondary importance were the feedwater, RHR, RWCU, core spray, and SRV systems. There was wide variability expressed for the feedwater system. Several participants thought that its susceptibility was similar to that of the recirculation system while others thought that it would make an inconsequential contribution. This latter group generally thought that the mitigation programs in place for the feedwater system were overall effective. The RHR system was deemed important by some panel members due to the relatively larger number of precursor events reported and the relatively high number of welds. A number of the participants used the weld census data provided to differentiate the relative contributions between systems for those systems that have similar operating experience. The SRV lines were judged to be potentially problematic by four of the eight respondents who addressed the question of BWR piping. They pointed out that the SRV lines are subject to high dynamic loads during the relatively common SRV discharge events, however, only a short section of these lines are actually susceptible to a LOCA event. Overall, in comparing Figure L.6 with Figure L.7, one can see approximately a one order of magnitude reduction in the LOCA frequency between the Category 1 and 3 LOCAs for most of the BWR piping systems considered.

For the largest category BWR piping LOCAs (Category 5), the recirculation system remains the main contributor to the overall LOCA frequencies, see Figure L.8. The RWCU system had about the same median value, however, there was a question expressed as to whether the RWCU system could actually sustain such a high flow rate LOCA. One of the participants thought that the maximum diameter for this system was only 6-inches, not 24-inches as specified in the development of the elicitation questions. Besides the recirculation, and RWCU systems, the next two largest contributors to the BWR Category 5 LOCA frequencies were the feedwater and RHR systems. As for the Category 3 LOCAs, the RHR system was deemed important due to the large number of precursor events reported and the large number of potentially susceptible welds. Several of the participants indicated that these lines are susceptible to SCC

Figure L.9 is a plot of the cumulative BWR piping LOCA frequencies (including contributions from all of the piping systems) for Category 1 through 5 LOCAs. The BWR piping LOCA frequency decrease with LOCA size is relatively shallow, i. e., approximately Уг order of magnitude per LOCA category. The results tend to be governed by the results from the recirculation system. It was noted that for the recirculation system that the mitigation programs in place for controlling IGSCC promote a more uniform residual stress field which can in turn promote longer cracks which are more likely to cause a LOCA.

This effect will potentially offset the overall reduction in crack growth due to the mitigation program. It is also of note from Figure L.9 that the variability in the results as expressed by the interquartile range and the difference between the minimum and maximum values does not vary much with LOCA size. It is also of note that the expert ranking is relatively consistent with LOCA size, i. e., Participant C always predicted the highest LOCA frequencies and Participants E and G consistently predicted the lowest LOCA frequencies.

Figure L.10 shows the effect of operating time on the cumulative Category 1 LOCA frequencies for BWR piping systems. As can be seen in Figure L.10, there is not much of an effect of operating time on the cumulative Category 1 frequency. Similar findings were evident for the larger Category 3 and 5 LOCAs. Obviously, any unabated degradation mechanism would cause an increase in the overall LOCA frequencies. However, it was generally assumed by the panel members that any new degradation mechanism that came on the scene would be aggressively met by the industry and NRC, just like the IGSCC problem in BWRs was met in the past and the PWSCC problem in PWRs is being met today. The minimal changes in LOCA frequencies with time evident in Figure L.10 were the result of a number of compensating factors considered by the panel members. From the perspective of potential decreases in the LOCA frequencies, the recirculation lines should see a decrease in the LOCA frequencies with respect to the current-day estimates that are based on an analysis of operational experience data due to improved mitigation strategies that have been put in place. The panelists generally felt that the IGSCC issue for BWRs had been effectively mitigated for the foreseeable future. In addition, the core spray systems may see a decrease in the LOCA frequencies with time as the segments of stainless steel piping potentially susceptible to IGSCC are replaced with carbon steel piping. Finally, future inspection and mitigation programs are expected to lead to additional decreases in the predicted LOCA frequencies. In this regard, having the industry focus its inspection resources on the more important systems through risk-informed ISI should help reduce the propensity for LOCAs. Counteracting these potential decreases are potential increases due to bigger thermal fatigue and FAC concerns in the future. Concern was expressed about the high usage factors that will exist near the end-of-plant license. Also, there is the concern with new, previously unknown degradation mechanisms that may arise in the future. In this regard, the inspection methods of today may not be reliable for these new mechanisms. Furthermore, these new mechanisms may not manifest themselves in the same locations of concern today. Finally, while timely and proper maintenance programs are always beneficial, there are instances in which they may prove counterproductive. The frequent opening and closing of systems for inspections increases the likelihood

for human error such as having tools and other debris left behind or bolts not being torqued properly. Also, improper service of active components (e. g., valves) can lead to passive system failures.

Figures L.11 and L.12 show the cumulative MV estimates, along with the 5% and 95% bound values for the various participants for the Category 1 and 3 LOCAs, respectively. The uncertainty range (difference between 5% LB and 95% UB values) for the Category 3 LOCAs are comparable (or slightly greater than) for the Category 1 LOCAs. Only participants A, E, and F expressed considerably more uncertainty for the Category 3 LOCAs than they did for the Category 1 LOCAs. Similar findings were found when comparing the Category 5 results with the Category 3 results. Overall, the panelists appeared more confident about their BWR estimates than they did for the corresponding PWR estimates. They had less uncertainty about future and bigger size LOCA frequencies compared with their PWR predictions. There was also less uncertainty among the panelists about the magnitude of the dominant contributing factors.

In addition, the panel members used more consistent approaches and more consistent base case estimates for the BWR estimates than they did

Figure L.14 shows the breakdown of PWR Category 3 LOCA frequencies by piping system at 25 years of plant operations (present day). The small diameter instrument and drain lines, as well as the RH lines, do not appear on this figure in that they are of such size that they could not sustain a Category 3 LOCA. Again, as was the case for the PWR Category 1 LOCAs, the smallest diameter lines that can sustain this size (i. e., category) of LOCA are the dominant contributors. These include the CVCS, SIS-DVI, RHR, surge, and PSL. This is different than what was observed for the BWR Category 3 LOCAs where the larger recirculation system was the dominant contributor, primarily due to its susceptibility to IGSCC.

The two most listed systems as being major contributors to this category of LOCA for PWR piping were the CVCS and SIS-DVI lines. For both, the primary concern was fatigue. One participant commented that the CVCS line was one of the most fatigue sensitive locations in the entire plant. Another commented that they were concerned with environmentally-assisted fatigue for this system. With regard to the SIS-DVI (and the SIS-Accumulator lines for that matter), several participants indicated that both lines had experienced thermal fatigue cracking in the past due to cold water leaking past the check valves. Another line that a number of participants thought would be a major contributor to this category of LOCA was the RHR lines. The concern with these lines was with environmental attack due to the stagnant nature of the flow in these lines. The pressurizer spray lines were of a concern due to the chance for PWSCC at one of the bimetal welds.

n = 9

1e-13 1 e-12 1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2

LOCA Frequency (yr-1)

Figure L.14 Category 3 LOCA Frequencies for PWR Piping Systems at 25 Years of Plant

Operation

For the largest categories of PWR piping LOCAs (Categories 5 and 6), the hot leg, cold leg, surge line, and RHR lines all contribute to the overall LOCA frequencies, see Figure L.15 for the Category 5 LOCAs. Of these, the median value of the LOCA frequency for the cold leg is about a half order of magnitude less than the median values for the other three piping systems. This slight reduction is primarily due to the fact that the cold leg is less susceptible to PWSCC than either the hot leg or surge line at this time (25 years of plant operations) due to the fact that it operates at a slightly lower temperature. Somewhat surprisingly in examining Figure L.15, a number of the participants felt that the hot leg would have a greater propensity for a Category 5 LOCA than the surge line. Both lines are susceptible to PWSCC due to the presence of bimetallic welds and the high operating temperatures, but the surge line was also judged to be susceptible to thermal fatigue due to thermal stratification and thermal striping stresses. Also, the surge line is smaller diameter, which based on the thought that smaller diameter lines are more prone to LOCAs than their larger counterparts, would imply that the Category 5 LOCA frequencies for the surge line should be higher. Finally, at least one participant argued that the surge line to pressurizer bimetallic weld was one of their biggest concerns in the entire plant due to its susceptibility to PWSCC and the fact that it is a very difficult weld to inspect. Counteracting these arguments, however, is the fact raised by a number of the participants that there are more hot leg to RPV bimetal welds (2 to 4 depending on the number of loops) in a plant than there are surge line to pressurizer bimetal welds (one).

LOCA Frequency (yr-1)

Figure L.15 Category 5 LOCA Frequencies for PWR Piping Systems at 25 Years of Plant

Operation

Figure L.16 is a plot of the cumulative PWR LOCA frequencies at 25 years of plant operation. Cumulative frequencies are shown for Category 1, 3, and 6 LOCAs. Based on a review of Figure L.16 there appears to be approximately a one order of magnitude reduction in LOCA frequency between each successive LOCA category.

——— 1——- 1——- 1——- 1——- 1——- 1——- 1—— 1——— 1——- 1——- 1——- 1——-

1e-15 1e-14 1e-13 1e-12 1e-11 1e-10 1e-9 1e-8 1e-7 1e-6 1e-5 1e-4 1e-3 1e-2

PWR LOCA Frequencies at 25 Years (CY-1)

Figure L.16 Cumulative PWR LOCA Frequencies at 25 Years of Plant Operations

Figure L.17 shows the effect of operating time on the cumulative Category 1 LOCA frequencies for PWR piping systems. Several participants felt that the operational experience is sufficient to expect the frequencies to remain relatively constant out to 60 years of life. Degradation and aging will naturally continue to occur. However, the inspection and mitigation strategies will effectively identify and temper the frequency increases caused by this aging. Some panelists expected a short term frequency increase due to PWSCC before effective mitigation is developed. This trend is consistent with the historical response to evidence of emerging degradation by the industry. Also, at least one participant expressed a concern about the high usage factors that will exist at 60 years at many locations. All of these concerns are reflected in the results showing the effects of operating time and aging in Figures L.17 and L.18 for Category 1 and 3 LOCAs, respectively. As can be seen in Figures L.17 and L.18, there is a slight increase in the cumulative Category 1 and 3 LOCA frequencies between 25 and 40 years, but not much of an effect between 40 and 60 years. The median LOCA frequencies for the Category 1 and 3 LOCAs at 40 years are an order of magnitude higher than the median LOCA frequencies for the Category 1 and 3 LOCAs at 25 years. Similar findings were evident for the larger Category 6 LOCAs. The rationale behind this is that this size of LOCA (and associated pipe size) is most affected by aging. These pipes are not as easily inspected, or as leak sensitive, as their larger counterparts and these pipes have not experienced the infant mortality as their smaller counterparts.

 1e-7 1e-6 1e-5 1 e-4 1e-3 1e-2 1 e-1 1e+0 1e+1

PWR Category 1 LOCA Frequencies(CY’1)

Figure L.17 Effect of Operating Time on the Cumulative Category 1 LOCA Frequencies for PWR

Piping Systems

 1e-8 1e-7 1e-6 1e-5 1 e-4 1e-3 1e-2

PWR Category 3 LOCA Frequencies(CY-1)

Figure L.18 Effect of Operating Time on the Cumulative Category 3 LOCA Frequencies for PWR

Piping Systems

 Solid diamonds are 5% Lower Bound, Mid-Value, and 95% Upper Bound Values A o—o^-o * _ * B 1 C ♦——————- ♦————- ♦ *

 H I . 0 ** " J L

Figures L.19 and L.20 show the cumulative MV estimates, along with the 5% and 95% bound values for the various participants for the Category 1 and 3 LOCAs, respectively. The uncertainty range (difference between 5% LB and 95% UB values) for the Category 3 LOCAs are typically greater than for the Category 1 LOCAs for most of the participants. In a similar vein, the level of uncertainty for the Category 6 estimates were much greater than for the Category 1 or 3 estimates, see Figure L.21. All of the panelists had at least two orders of magnitude difference between the LB and UB values for their Category 6 estimates, and some of the panelists (C, E, and J) had greater than four orders of magnitude difference.

1.00E-17 1.00E-15 1.00E-13 1.00E-11 1.00E-09 1.00E-07 1.00E-05 1.00E-03 1.00E-01

PWR Category 1 LOCA Frequencies (CY-1)

Figure L.19 PWR Category 1 LOCA Frequencies Showing MVs, 5% LB, and 95% UB Values for
All Participants Who Responded to the PWR Piping Questions

 ♦—{——— ♦ G H ‘ * . . I J L

 Solid diamonds are 5% Lower Bound, Mid-Value, and 95% Upper Bound Values A * ♦—Ц—- ♦ b C

 1.00E-17 1.00E-15 1.00E-13 1.00E-11 1.00E-09 1.00E-07 1.00E-05 1.00E-03 1.00E-01 PWR Category 3 LOCA Frequencies (CY-1)

 Figure L.20 PWR Category 3 LOCA Frequencies Showing MVs, 5% LB, and 95% UB Values forAll Participants Who Responded to the PWR Piping Questions

 Solid diamonds are 5% Lower Bound, Mid-Value, and 95% Upper Bound Values о———- о—о A * ♦—————— ♦ B * ♦——————————————- ♦ C

 E

 G

 j

 L

 1.00E-17 1.00E-15 1.00E-13 1.00E-11 1.00E-09 1.00E-07 1.00E-05 1.00E-03 1.00E-01

 Figure L.21 PWR Category 6 LOCA Frequencies Showing MVs, 5% LB, and 95% UB Values forAll Participants Who Responded to the PWR Piping Questions

In general, the results for PWR piping appear consistent. The quantitative results and the qualitative opinions and rationale were for the most part consistent. The variability between participants stems from the different approaches used and the basis for their estimates. Several different approaches with different anchoring points were used by the different panelists. The variability between the participants seems reasonable given the frequency magnitudes being computed.

## Benchmarking

A limited scope benchmarking exercise was performed to compare predicted weld failure rates with the reported service experience. The benchmarking was limited to NPS12 BWR reactor recirculation welds susceptible to IGSCC. Probabilistic fracture mechanics (PFM) calculations using the WinPRAISE computer code generated predictions about the weld failure rate for different assumptions about the normal operating stresses (cNO).18 Bayesian reliability analysis was used to derive weld failure rates from service experience data. Figure D.50 shows the results of the benchmarking exercise. Table D.28 includes a description of the different cases of the benchmarking exercise.

D.7.4 Comparison to Historical LOCA Frequency Estimates

Figures D.51 and D.52 compare the Base Case results to historical LOCA frequency estimates. Direct (one — to-one) comparisons are not feasible due to different LOCA definitions and estimation approaches. Listed below are the selected BWR and PWR LOCA frequency references.

BWR Large (> Cat3) LOCA Frequency Estimates (Figure D. 51)

• SKI 98:30 (FW/RR); the displayed value range is taken from Reference [D.18]. It excludes contribution from thermal fatigue in Code Class 1 feedwater system piping. The feedwater system design is unique to the pilot plant in SKI Report 98:30 and it is therefore not applicable to BWR-2.

• NUREG/CR-5750 (Appendix J) provides recommended pipe LOCA frequencies. The given value range accounts for all Code Class 1 pipe failure contributions.

• GRS-98 is a probabilistic safety assessment of the German plant Gundremmingen; a BWR plant designed and built by Kraftwerk Union. This reactor design has no external recirculation loops; the given LOCA frequency value range accounts only for contributions from Code Class 1 feedwater pipe failure.

• BFN-1 (NUREG/CR-2802) is the 1982 probabilistic safety assessment of Browns Ferry Unit 1 performed as part of the NRC-sponsored Interim Reliability Evaluation Program. The given LOCA frequency value range accounts for Reactor Recirculation pump suction piping failure.

PWR Large (> Cat3) LOCA Frequency Estimates (Figure D. 52)

• NUREG/CR-5750 (Appendix J) provides recommended pipe LOCA frequencies. The given value range accounts for all Code Class 1 pipe failure contributions.

• Surry-1 (1990 Expert Elicitation). Surry-1 is a 3-loop Westinghouse reactor, similar to the PWR- 1/PWR-2 reference design. The given LOCA frequency value range applies to RCS pipe failure and resulted from a NRC-sponsored expert elicitation.[13]

• EPRI TR-100380 (Piping Failures in U. S. Commercial Nuclear Power Plants, 1992) includes recommended BWR and PWR LOCA frequencies that are based on statistical analysis of service data. The given LOCA frequency value range accounts for all Code Class 1 pipe failure contributions.

## BWR Reactor Vessels

Analyses have been previously submitted and approved [I.3, I.4] that establish reduced inspection requirements for BWR reactor vessels relative to ASME Section XI requirements. Specifically, BWRVIP-05 [I.3] justifies that only axially-oriented welds in the vessel beltline region need be examined on a ten year interval, versus the Section XI requirement to inspect all axial and circumferential welds on this interval. This relief was based on PFM calculations demonstrating, for the BWR fleet, that circumferential weld inspections contribute negligibly to reduction in the already small failure probability of a BWR vessel. The methodology used for the PFM analysis is a computer program (VIPER [I.7]) developed by Structural Integrity Associates for EPRI and the BWRVIP. To address this LOCA frequency contributor, the VIPER software was run for a typical BWR vessel, extending the analysis period from 40 to 60 years. A modification to the software (VIPER-NOZ) was also used to estimate leakage and failure probabilities for BWR Reactor Vessel feedwater nozzles. Feedwater nozzles were selected because they are subject to thermal fatigue cycling, which caused serious nozzle cracking in the 1970s [I.5]. Both analyses take credit for routine ISI programs that are conducted on these components on ten-year inspection intervals. The feedwater nozzle analysis also takes credit for nozzle modifications and thermal sleeve improvements that were installed in all U. S. BWRs to reduce the severity of the thermal fatigue cycling.

## Comments Related to Section 4 of NUREG-1829

Comment Number: 4-1

Submitted by Joseph Conen of the BWR Owners Group

Comment: Table 4.1 shows a six-order of magnitude difference between the PFM and the field history estimates of through-wall cracking frequencies for the BWR-2 base case. Although the report suggests that service history data could be analyzed to resolve this difference, it is not clear if this was actually done. This also points out to the need for a rigorous examination of the statistical methods used to translate field leak data or service experience data into pipe break frequencies.

Response: The complexity of translating field leak data or operating-experience data into pipe break frequencies was part of the rationale for conducting the elicitation (Section 1). It was also a major consideration for performing multiple base case analyses so that results from different approaches could be compared. The small break LOCAs and through-wall cracking frequencies represent the categories where operating experience and PFM comparisons are most meaningful because the least extrapolation is required. For the small break LOCAs (Category 1), the results are typically within 2 orders of magnitude for most of the base case team members. However, the total spread in both the through-wall cracking leak frequencies (Figure 4.1) and SB LOCA frequencies (Figure 4.2) can be greater than three orders of magnitude. As the commenter indicates, the BWR-2 Category 0 frequencies (Figure 4.1) vary significantly (4-5 orders of magnitude in final NUREG)

Furthermore, while the PFM and the operating-experience predictions of the base case analyses did not always readily agree, this simply reflects the current uncertainty in calculating these estimates.

Differences among these estimates reflect the different assumptions and approaches used in the various analyses. Specifically, the BWR-2 operating-experience frequency estimate was based on 20 reported incidents in BWR feedwater systems. However, none of these events resulted in through-wall leakage. Therefore, the operating-experience estimate is undoubtedly conservative.

The elicitation reported in NUREG-1829 did not attempt to resolve these discrepancies; rather, the elicitation’s goal was to reflect the scientific uncertainty in the technical community. Therefore, the purpose of the base case estimates was not to obtain convergence among the predictions. Rather, it was to understand how the assumptions and approaches associated with particular analyses contributed to the disparity in the results. These analyses and their differences were clearly presented to the elicitation panelists so that they could judge their effect as part of their elicitation. As a result, most panel members chose to anchor their current-day LOCA Category 1 responses using operating experience and used PFM, if at all, to inform their assessments for LOCA sizes greater than Category 1 or for future LOCA frequency estimates. These are the domains where relevant operational experience data does not exist.

Key elements of this response were incorporated into Section 4.2 of the revised NUREG.

Comments Related to Section 5 of NUREG-1829

None