Limited to the BWR Base Case systems, this section summarizes the worldwide service experience with Code Class 1 piping. The results of this review are input to the pipe failure rate estimation.

D.3.2.1 RR Piping Service Experience — The original piping material in BWR plants commissioned prior to mid-1980 is austenitic stainless steels that contain more than 0.03% carbon. During welding these steels are susceptible to sensitization that results in a loss of corrosion resistance. Intergranular stress corrosion cracking (IGSCC) occurs when the sensitized steel is subjected to stresses and corrosive environment. Sensitization can be avoided by controlling the carbon content to below 0.03%. Another approach to controlling sensitization is to add strong carbide formers such as titanium or niobium to the steel. Stainless steels with additions of titanium or niobium are called “stabilized.” It is noted that low-carbon content unstabilized stainless steel or stabilized stainless steels are not completely immune to IGSCC, however [D.18].

For Plant B, IGSCC is the predominant degradation mechanism acting on the RR piping welds, including heat-affected zones. During early plant life some weld reinforcements were performed where the inservice inspection revealed presence of surface penetrating, and subsurface cracking due to IGSCC. Since the analysis of LOCA frequency distributions is based on a degradation mechanism evaluation, the PIPExp database is queried for service data including IGSCC. The database queries are summarized in a set of charts and tables below. The database currently includes a total of about 1000 records on IGSCC in BWR piping. Figure D. 1 shows the number of weld failures due to IGSCC by calendar year and Figure D. 2 shows the number of weld failures by year of operation. Here a weld heat affected zone with a/t > 10% is characterized as a “weld failure.”

Figure D.1 Weld Failures Due to IGSCC in Code Class 1 & 2 Piping (1970-2002)

Figure D.2 IGSCC Experience by Year(s) of Operation

In Figure D. 3 the IGSCC data is organized by mode of failure (crack — pinhole leak — leak) and pipe size. Figure D. 4 shows the IGSCC data by size and material type.

100.00%

90.00%

80.00%

70.00%

60.00%

50.00%

40.00%

30.00%

20.00%

10.00%

0.00%

NPS12 NPS22 NPS28 All

Figure D.3 IGSCC Data by Failure Mode

300

In Figure D.5, the combination of a/t = 100% and L/C = 100% would indicate a case of DEGB where the pipe ends are separated from each other. As a rule-of-thumb, a through-wall crack (a/t = 100%) with L/C > 40% is unstable and may exhibit unstable crack growth if it were to be left in place.[3]

As seen from the above, there have been a limited number of cases of leaks in large-diameter Reactor Recirculation piping. Only a small fraction of the total number of through-wall flaws have been active leaks; i. e., leaks that have developed during routine power operation. The majority of the through-wall flaws have been “non-active leaks.” That is, leaks that have developed while shutting down for drywell inspection, during performance of weld crown grinding in preparation for ultrasonic examination (“ISI-leaks”), or during the performance of induction heat stress improvement (IHSI — “IHSI-leaks”). There are also some cases where leaks have been discovered during hydrostatic pressure testing to verify the integrity of weld repairs.

Like Figure D.5, Figure D. 8 includes data on all IGSCC-susceptible, Code Class 1 and 2 piping systems in BWR plants. While Figure D.5 includes approximately 300 data points, Figure D.8 includes on the order of 500 data points. This difference in the number of reports represented in respective chart is due to the fact that not all reports on IGSCC include complete details on the crack morphology (dimensions, orientation).

Where through-wall flaws have been observed leak rates have been small. In terms of leak rate and operational impact, so far the two most significant instances of IGSCC occurred at Duane Arnold in 1978 and at the Spanish plant Santa Maria de Garona in 1980. In the former case the leak rate was about 11 lpm (3 gpm) with L/C = 22%. In the latter case the observed leak rate was about 3.0 lpm (0.8 gpm) with L/C = 4.5%.

D.3.2.2 FW Piping Service Experience — Figures D.9 and D.10 summarize the service experience with FW piping. With respect to plant designed by General Electric, the Code Class 1 portion of BWR carbon steel feedwater piping has performed well in the field. There are no reported leaks in medium-or large — diameter RCPB piping. Foreign plants have experienced (and in some cases, continue to experience) thermal fatigue damage due to thermal mixing and stratification. In fact, 80% of the degradation of the RCPB portions of FW piping has occurred in foreign plants with a piping system design that differs from that of U. S. BWR plants.

The U. S. service experience includes a few instances of non-through wall cracking of FW nozzle-to-safe-end (bimetallic) welds. The root cause of the cracking is attributed to weld defects from original construction. As documented in Information Notice 92-35 [D.19], Susquehanna Unit 1 has experienced flow-accelerated corrosion damage about 250 mm (10 inches) from a weld connecting NPS12 piping to a 20-inch by 12-inch reducing tee. There

Figure D.10 Service Experience with FW Piping (ii)

Each participant filled out an elicitation questionnaire dealing with age related health issues. The results from this exercise were reviewed with the meeting participants on Wednesday morning. As part of this exercise, Lee Abramson indicated that it is usually easier to determine relative rates versus absolute rates. Various absolute and relative questions were posed in order to demonstrate this concept.

Definition of Terms for Elicitation

Terms used during the elicitation must be commonly understood by the group in order to foster discussion, issue development, and subsequent elicitation. Certain key terms must be defined.

Rob Tregoning indicated that, for this exercise, all definitions should be kept generic, not plant specific. The first term to be defined is LOCA. Rob Tregoning presented the NUREG/CR-5750, Appendix J definition as a starting point. This report defines a LOCA as “an unisolable breach of the Reactor Coolant Primary Boundary (RCPB) requiring ECCS initiation.”

The group felt that the term “unisolable” was not appropriate because the main point is to limit the scope to Class 1 piping. Also, the merits of the phase ECCS initiation were debated because the ECCS response in some plants requires use of normally operating plant equipment. Therefore, some plants might require a large leak before implementation of standby ECCS systems. There was also a discussion on the merits of using break instead of breach, but the term breach was determine to be more generic than break. The addition of the term “sudden breach” instead of just “breach” was also neglected because of the vagueness of the word sudden.

The group agreed to a definition of a general LOCA as follows. A LOCA is “a breech of the reactor coolant pressure boundary which results in a leak rate beyond the normal makeup capacity of the plant”.

The next definitions are required to determine the size classifications of LOCAs. Once again, Rob Tregoning presented the definitions used in NUREG/CR-5750, Appendix J as a starting point. These definitions were also used in NUREG-1150 and form the basis of plant PRA event trees. This document defined three LOCA size categories: SB, MB, and LB. The NUREG/CR-5750 definitions are as follows:

• SB LOCA — A break that does not depressurize the reactor quickly enough for the low pressure systems to automatically inject and provide sufficient core cooling to prevent core damage. However, low capability systems (i. e., 100 to 1,500 gpm [380 to 5,700 lpm]) are sufficient to make up the inventory completion. For a BWR, this translates to a pipe in the primary system boundary with a break size less than 0.004 ft2 (370 mm2), or a 1 inch (25 mm) equivalent inside pipe diameter, for liquid, and less than 0.05 ft2 (4,600 mm2), or an approximately 4 inch (100 mm) inside diameter pipe equivalent, for steam. For a PWR, this equates to a pipe break in the primary system boundary with an inside diameter between Уг to 2 inches (13 to 50 mm).

• MB LOCA — A break that does not depressurize the reactor quickly enough for the low pressure systems to automatically inject and provide sufficient core cooling to prevent core damage. However, the loss from the break is such that high capability systems (i. e., 1,500 to 5,000 gpm [5,700 to 19,000 lpm]) are needed to makeup the inventory depletion. For a BWR, this translates to a pipe in the primary system boundary with a break size between 0.004 to 0.1 ft2 (370 to 9,300 mm2), or an approximately 1 to 5 inches (25 to 125 mm ) inside diameter pipe equivalent, for liquid, and between 0.05 to 0.1 ft2 (4,600 to 9,300 mm2), or an approximately 4 to 5 inches (100 to 125 mm) inside pipe diameter equivalent, for steam. For a PWR, this equates to a pipe break in the primary system boundary with an inside diameter between 2 to 6 inches (50 to 150 mm).

• LB LOCA — A break that depressurizes the reactor to the point where the low pressure system injection automatically provides sufficient core cooling to prevent core damage. For a BWR, this translates to a pipe in the primary system boundary with a break size greater than 0.1 ft2 (9,300 mm2), or an approximately 5 inch (125 mm) inside diameter pipe equivalent, for liquid and steam. For a PWR, this equates to a pipe break in the primary system boundary with an inside diameter greater than 6 inches (150 mm).

The elicitation panel questioned the basis of the equivalent pipe diameter relationships to break size provided in the NUREG/CR-5750 Appendix J. Bill Galyean thought that they could be traced back to NUREG 1150 and possibly WASH-1400. It was quickly determined that “break” should be replaced by “breech” everywhere for consistency with the general LOCA definition. Also, the group decided that the formal definitions should be based on leak rate, and not equivalent break area or size.

At this point, the need for additional LOCA size classification was revisited. This request was promulgated by Bruce Bishop based on discussions with the Westinghouse Owner’s Group (WOG). The System response and mitigation procedures for a 5,000 gpm (19,000 lpm) LOCA (lower limit LB LOCA within NUREG/CR-5750, Appendix J) and a DEGB of the largest class 1 pipe (flow rate up to 860,000 gpm [3,250,000 lpm] according to WOG) are significantly different. Because, the elicitation results will used in existing PRAs, it was also stressed that the original leak rate classifications in NUREG/CR-5750, Appendix J should also be maintained. While the group also agreed that the leak rate threshold should ideally be based on the equipment needed to mitigate a specific event, this information is highly plant specific and could not be approximated generically.

For the reasons stated in the above paragraph, the group decided to keep the NUREG/CR-5750, Appendix J leak thresholds of 100 gpm (380 lpm), 1,500 gpm (5,700 lpm), and 5,000 gpm (19,000 lpm), but to add several leak rate categories above 5,000 gpm (19,000 lpm). The highest category was set at 500,000 gpm (1,900,000 lpm) to capture the DEGB events of the largest primary system pipes. Additional ranges of 25,000 gpm (95,000 lpm) and 100,000 gpm (380,000 lpm) were chosen to span the range from 5,000 gpm (19,000 lpm) to 500,000 gpm (1,900,000 lpm) in roughly equivalent magnifications. These leak rate categories were also chosen because they tend to group DEGBs by primary system functionality.

The LOCA size classification thresholds adopted by the group are summarized in Table B.1.1[2]. A category 1 LOCA is defined as “a breach of the reactor coolant pressure boundary which results in a leak rate which is greater than 100 gpm (380 lpm). Similarly, a category 6 LOCA is a breach of the RCPB which results in a leak rate which is greater than 500,000 gpm (1,900,000 lpm). It should be stressed that category 1 LOCAs include contributions from all categories. The group preferred the threshold classification of LOCA sizes instead of partitioning the sizes into ranges as in NUREG/CR-5750, Appendix J. Care will be needed during the elicitation to ensure that these definitions are understood.

It was determined by the group that these leak rates should be roughly correlated to breach area, and converted into an equivalent pipe diameter so that the UB leak rates for various piping systems could be determined. There was some concern about the feasibility of developing generic estimates. It was suggested that equivalent pipe sizes could be based on 250 gpm/in2 (1.47 lpm/mm2) for liquid PWR lines and 175 gpm/in2 (1.03 lpm/mm2) for liquid BWR lines. However, these estimates did not agree with the Westinghouse equivalent pipe diameter estimates.

Pete used the VIPER program to predict beltline failure frequencies (per vessel year) for typical BWRs. VG3 and 4 are frequency plots and not probability plots.

For VG7 for large Category LOCAs, see big impact with time between 40 and 60 years; attributed to effect of radiation embrittlement; for smaller LOCAs don’t see much effect of time.

EDY stands for Effective Degradation Years; used to normalizes degradation to a reference of a 600oF operating temperature.

For CRDM nozzle ejection probability the assumption in VG11 that immediately have circumferential TWC of 30 degrees is highly conservative according to Bruce in that most are axially oriented.

Of 30 plants, there were 11 nozzles that had circumferential cracks; all of these plants were at about 20 EDY so Pete could take time factor out; total number of nozzles in 30 plants was 881

POD curve for NDE; cracks were EDM notches that were compressed to make them tight; eventually will get some real cracks from the North Anna head that can be used for calibration/validation

VG15 shows the probability of leak in one of 98 nozzles in this plant per vessel year; shows effect of NDE on probability of leakage

VG15 and 16 show effectiveness of NDE and how PFM models can account for inspection in their analyses.

VG17 shows decreasing frequency with time which reflects benefit of inspection.

To get 5,000 gpm (19,000 lpm) leakage, need ejection of 2 nozzles; most likely scenario is for collateral damage as one ejects and causes damage to adjacent nozzles.

Day 1 — June 4, 2003 — Base Case Review

Dr. Rob Tregoning (USNRC) welcomed everyone to the Second Elicitation meeting and reviewed the agenda for the two days. Everyone in attendance introduced themselves. A package of the Day 1 presentations was provided to everyone. Rob warned the group that there was a lot of material to cover in each presentation. Furthermore, he indicated that we should not treat what is presented at this meeting as final, but more of a snapshot of where we are presently. Next, Rob reviewed the objectives for the first day of the meeting. The objectives for the first day were:

1. Review base case conditions

2. Understand assumptions, methodologies, and results calculated by each base case member

3. Understand important factors and variables that lead to differences among results

4. Determine what additional calculations are required to complete the base case analysis

Rob also indicated that he was not asking everyone to agree with the results to be presented, but simply to understand what was done. The panel members can state differences of opinion during their individual elicitations.

Next, the second day agenda (Elicitation Coordination) was discussed. The second day agenda has been adjusted slightly to ensure adequate time for the topic of non-piping LOCA frequency determination since it received less attention at the last meeting

Rob then reviewed the meeting objectives for Day 2. The Day 2 objectives were:

1. Finalize elicitation question sets and provide consistent understanding of each question.

2. Determine methodology for evaluating non-piping LOCAs and identification of non-piping base case data.

3. Determine methodology for evaluating conditional seismic loading including determination of seismic loading magnitude.

4. Determine what information panelists will require prior to their elicitations and assign action items for providing information.

5. Develop final schedule and time-frame for upcoming elicitations.

Bruce Bishop (Westinghouse) asked if the panel would get a status report on the new PFM code being developed as part of the USNRC program. Rob indicated that time was not available at this meeting, but that sometime in the fall or winter an initial meeting will be set up where this new code being developed by Battelle and Emc2 could be presented.

PIPING BASE CASE RESULTS OF BENGT LYDELL

An Application of the Parametric Attribute-
Influence Methodology to Determine Loss of
Coolant Accident (LOCA) Frequency Distributions

Report No. 2 to the NRC Expert Panel on
LOCA Frequency Distributions

Prepared for

U. S. Nuclear Regulatory Commission
Washington (DC)

June 2004

D-1

ACKNOWLEDGEMENTS

The work documented in this report was performed for the U. S. Nuclear Regulatory Commission under Subcontract No. 177115 (Battelle Memorial Institute).

Mr. Karl N. Fleming (Technology Insights, Inc., San Diego, CA) provided constructive review comments on a draft of this report. Mr. Fleming also provided the Markov model solutions supporting the calculation of time-dependent LOCA frequencies.

D. 1 Background

Limited to consideration of Code Class 1 piping failures, Base Case Report Number 2 documents an assessment of BWR and PWR loss of coolant accident (LOCA) frequency distributions. The assessment is a demonstration of the role of statistical analysis of service experience data and Markov modeling in a “bottom — up” approach to piping system reliability analysis.

D. 1.1 Objectives

Using primary coolant piping design information for three reference plants (one BWR plant and two PWR plants), the overall objective is to determine LOCA frequency distributions that are representative of currently operating U. S. nuclear power plants, including current in-service inspection (ISI) practices and degradation mitigation strategies. This determination is done analytically using a parametric model of piping reliability. The LOCA frequency distributions are determined for three time periods. To address today’s piping reliability state-of-knowledge the LOCA frequency is determined at T = 25 years. Next the LOCA frequency is extrapolated to T = 40 years to represent the primary system piping reliability status at the end of a 40-year operating license. Finally an extrapolation is made to T = 60 years to account for a possible license renewal. Analytically, this extrapolation is concerned with the potential impact on the structural integrity of the piping by material aging as well as by reliability improvement efforts.

As implied by the report title, the objective is to develop LOCA frequency distributions. The report addresses two aspects of LOCA frequency distributions. It develops LOCA frequencies associated with a distribution of flow rate threshold values ranging from 380 lpm (100 gpm) at the low end to beyond 380,000 lpm (100,000 gpm) at the high end. Additionally the study develops statistical uncertainty distributions for each set of LOCA frequencies to account for the uncertainty in the input parameters to this piping reliability analysis.

Mr. Lydell has 30 years of risk and reliability analysis experience. Prior to joining ERIN®, he held positions with the Swedish Nuclear Power Inspectorate (SKI), Pickard, Lowe and Garrick, Inc., and NUS Corporation. Mr. Lydell has extensive, practical experience with applied quantitative risk assessment. In various capacities (systems analyst, human reliability analyst, independent reviewer), he has supported numerous domestic and foreign PSA projects (Level 1 and 2, and internal flooding). As an independent contractor, during the period 1993-99 he performed R&D in piping reliability analysis for the oil and gas and nuclear industries. This work explored field experience data and its role in quantitative piping reliability analysis, including the interfaces between PSA requirements and PFM. The SKI pipe failure database resulted from this work. Under contract to SKI and BKAB (a Swedish utility), during 1998-99 he performed a pilot LOCA-frequency study; a summary report is published as SKI Report 98:30 (May 1999). This particular study was commissioned to address the feasibility of applying BWR pipe operational experience data to the estimation of plant-specific LOCA frequencies. The SKI pipe failure database formed the basis for the OECD Nuclear Energy Agency’s “OECD Pipe Failure Data Exchange” Project (OPDE), an international forum for the exchange of pipe failure information. Managed by Mr. Lydell, a clearinghouse is operating the OPDE database and provides the quality assurance function.

Rob Tregoning and Lee Abramson want to do 2 elicitations early, possibly the week of July 14th. Then take a month off and restart the elicitations in mid August with the goal of completing all of the elicitations by the end of September. That would leave about a month to analyze the results. Rob is looking for volunteers to be the first two individuals to go through the elicitation process. All panel members should give their schedule to Rob Tregoning so that the individual elicitations can be scheduled.

Once the elicitations are complete and the results analyzed, a wrap-up meeting will be held in the October/November timeframe. As part of this meeting the calculated LOCA frequencies will be presented and any interesting and surprising results from individual questions will also be presented. The panel will also be solicited for feedback on the process. We will try to identify strengths and weaknesses with the process. Any necessary follow-on work will be defined during this meeting. We are also open to suggestions for conducting a reanalysis of these results in ten years. It is anticipated that the wrap-up meeting will take two days.

Again, it was noted that the panel members can change the results after their individual elicitations and after the wrap-up meeting. However, the final report must be submitted by the end of December. It is too premature to speculate on the form of the final report. However, confidentiality of the individual elicitation opinions will be maintained.

Limited to the PWR Base Case systems, this section summarizes the service experience with Code Class 1 piping. The results of this review are input to the pipe failure rate estimation.

D.3.3.1 RC & HPI/NMU Piping Service Experience — There have only been a limited number of events involving through-wall cracks in the large-diameter RC piping and the Class 1 portion of SI/CV piping. Evidence of axial primary water stress corrosion cracking (PWSCC) in the bimetallic safe-end to RPV nozzle welds of the RC-HL piping has been reported at Ringhals [D.20] and V. C. Summer [D.21].

During an eight-year period, the now decommissioned Trojan nuclear power plant experienced pressurizer surge line movement, which was attributed to thermal stratification [D.22]. In response, the NRC issued Bulletin 88-11 in December of 1988 [D.23] requesting that licensees perform visual inspections of the pressurizer surge line at the first available cold shutdown. Purpose of the inspections was to determine presence of any “gross discernible distress or structural damage in the entire pressurizer surge line, including piping, pipe supports, pipe whip restraints, and anchor bolts.”

The current version (June 2004) of the PIPExp database includes four records associated with degradation of pressurizer surge lines:

• Record # 19849; during the Three Mile Island-1 2003 Refueling Outage (18-Oct-2003 to 3-Dec — 2003), a UT examination found an axial flaw about 13 mm (0.5-inch) deep in the surge line nozzle — to-safe end interface in dissimilar metal weld No. SR0010BM. This weld connects a 10-inch Schedule 140, carbon steel nozzle to stainless steel safe end.

• Record # 19736; in November 2002 during UT examination of RC piping in the Belgian plant Tihange-2 (a 900 MWe series plant designed by Framatome), code rejectable indications were

discovered in the 14-inch Inconel safe-end to nozzle weld. The flaw is believed to be an original construction defect.

• Record # 1119; while in hot shutdown condition, a non-isolable weld leak developed in a 1 — inch drain line off the pressurizer surge line of Oconee-1 (LER 50-269/1998-002-01). The through-wall crack had initiated by TGSCC and propagated through-wall by vibratory fatigue. Small-diameter piping connecting to a pressurizer surge line is not part of the PWR-2 Base Case definition.

• Record # 420; during the 1988 annual refueling outage a pinhole leak was discovered in a 10-inch pressurizer surge line bi-metallic weld of Loviisa-1 (a Soviet designed WWER-440/213 plant located in Finland). The weld degradation was attributed to poor weld penetration and high residual stresses. This event was screened out from the data analysis.

Figures D.11 and D.12 summarize relevant service experience with medium — and large-diameter RC and safety injection (SI) and normal makeup (CV) piping. For comparison, Figure D.13 shows the service experience with small-diameter RC and SI/CV piping (< NPS2). Figure D.14 is a summary of the worldwide, PWR-specific data pipe failures that are attributed to thermal fatigue. In addition to RC-, SI — and CV-piping this figure includes failures in FW — and RHR-piping.

Figure D.11 Weld Failures in PWR RC-, CV — and Si-Piping (1970-2002)

N <ъ <o л o> NN <ь Ф <v ч<ь ф <p <p $ Ф>

Year(s) of Operation

Figure D.14 Pipe Failures Attributed to Thermal Fatigue in PWRs Worldwide

Figure D.14 includes four significant (v > 38 lpm [10 gpm]) events, three in foreign plants (Civaux-1 in France, Tsuruga-2 in Japan and Biblis-B in Germany) and one in a domestic plant (Oconee-2). The latter event involved a failure of a weld between the HPI/NMU and the RCS cold leg (= PWR Base Case Plant A. b). The plant operators correctly diagnosed the leak and brought the plant to safe shutdown. Subsequent to the weld failure in Oconee-2, limited to small-diameter piping the Electric Power Research Institute issued the “Interim Thermal Fatigue Guideline” [D.9] for evaluating and inspecting regions where there might be high potential for thermal fatigue cracking. Additional perspectives on thermal fatigue mitigation are included in an OECD-NEA report [D.24]. The Babcock &Wilcox-designed plants now include a new design thermal sleeve to mitigate or prevent thermal fatigue cracking of welds.

Prior to these ‘four significant events’, thermal fatigue damage occurred at Farley-2 and Tihange-1 (a Belgian plant) during 1988. At these plants, thermal fatigue initiated from cold water leaking through closed check or globe valves in safety injection lines. At Farley-2, the damage occurred in piping connected to the RCS cold leg, and at Tihange-1 in piping connected to the RCS hot leg. In these events the leak rates were 2.6 lpm (0.7 gpm) and 23 lpm (6 gpm), respectively. The U. S. NRC issued Bulletin 88-08 in response to these events.

D.4 Data Processing and Data Reduction

The objective of data processing is to extract from a pipe failure data collection relevant case histories that reflect specific combinations of reliability attributes and influence factors. Next, the data reduction prepares the input to the statistical parameter estimation in the form of event counts and exposure terms to develop Bayesian prior and posterior distributions.

This presentation by Bengt Lydell discussed the SKI-PIPE database evolution and background; the database structure and content; current database status; and LOCA frequency estimate conducted with the data base. Some of the specific key points from his presentation and subsequent discussion are outlined below:

• Background: The database was motivated to create a tool that would serve both PRA and the PFM/material science practitioners. It’s structured to provide information to completely define the piping systems attributes (design characteristics) and the influence functions (operating history) which govern system failure probability. By thoroughly assessing these features it is possible to determine plant specific estimates of piping system reliability.

• Structure and Content: The database covers pipe failures in commercial nuclear power plants from 1970 to the present.

• It should be stressed that SKI-PIPE only includes failures in piping systems, external to the RPV. Non-piping system failures are not included. Also, SKI-PIPE contains only passive piping failures of metallic piping.

• A pipe failure is defined in the database as any degradation that results in piping repair or replacement.

• Each record in the database is indexed. References to the original data source (e. g., LER report) and supporting information are provided. All the supporting documented is stored electronically.

• The database is organized by reliability attributes (i. e. design features such as material, dimensions) and influence factors (i. e. unique service conditions, including degradation susceptibility).

• When the original record is incomplete (such as an LER), a best effort is made to fill in database gaps by directly contacting the plant operators.

• It is noted in the database when each record consists of multiple flaws at a single component location. However, subsequent data entries are typically associated with only the largest flaw at that location.

• The database includes both surface penetrating flaws and non surface penetrating flaws (i. e., embedded flaws).

• Current Database Status:

• The database is continually being updated.

• The current OECD-sponsored OPDE project has participants from 12 nations. The first year of the three year effort is concerned with adding and validating database entries for each of the member countries from 1998 through 2001.

• Raw data is currently obtained from over 40 different sources

• Applications: Two relevant studies are the determination of LOCA frequencies for the

Barseback-1 plant and examination of IGSCC in Russian graphite moderated reactors (RBMK).

• The Barseback-1 study employed plant-specific attribute and influence functions which were comprehensively developed for all “known and credible” degradation mechanisms.

• The Beliczey and Schulz conditional rupture probability was not used in the Barseback-1 analysis. Instead a Bayesian update of a Jeffrey’s modified non-informative prior was employed.

• The database results have been compared with PFM predictions for welds in certain systems with some success.

• The RBMK studied indicated that the experience today with IGSCC in Russia is similar to US BWR IGSCC cracking experience in the late 70’s to early 80’s, before wide-spread mitigation was adopted

Discussion: The panel asked if they could get copies of the SKI-PIPE database. Karen Gott of SKI indicated that it is possible to distribute a non-proprietary version of the database. This non-proprietary version contains piping failures thru 1998.

Almost everyone agreed that for PWRs the dominant failure scenario for Category 1 LOCAs was steam generator tube failures.

Used non-piping database which was augmented back to 1987 to capture 2 major events in ‘87 and ’89. There have been 15 leaks since 1990 with 4 events over 100 gpm (380 lpm) since 1987.

On VG2 the reference to Nine Mile Point should be to Indian Point.

Fred Simonen stated that usual scenario assumed for higher category LOCAs is common cause failures such as losing pressure on secondary side causing pressure differential across tube and failure of multiple already degraded tubes.

Rob’s analysis of independence (ignoring common cause) was viewed with a great deal of skepticism;

Bill Galyean commented that if look at LERs always see multiple tube degradation but only see single tube rupture in the LERs.

This analysis is for 25 years, panel members left to their own devices for later years.