The purpose of this presentation was to review the conditions analyzed by the base case team and summarize the calculated results to date. This talk served as a prelude for the next four presentations by the base case team members.

The base case results will be used to anchor elicitation responses by the elicitation panel members as part of their individual elicitations. The elicitation members can use one or all of the base case results directly for their anchoring, or provide their own base case analysis if they choose. There were a total of five (5) base cases defined at the first elicitation meeting in February, see Table B.2.1.

As part of the base case effort, the base case members were to evaluate the LOCA frequencies at 25 years (current-day), 40 years (end-of-plant-license), and 60 years (end-of-plant-license-renewal). These results will then be used by the individual elicitation panel members to anchor their respective responses so that they can estimate the various LOCA frequencies at these same time periods.

The goal for the base case members is to calculate results for the set of conditions listed in Table B.2.1. The base case members also shared their results and presentations prior to this meeting so that there was a common format for the presentations. At this time the base case results comparison charts in the handouts should be viewed as works in progress. Furthermore, some results from Vic Chapman (OJV Consultancy) are still forthcoming. Vic and Chris Bell need to provide additional information to the panel members on how they conducted their base case analyses.

The current base case results are summarized in slides 16 through 20 of this presentation. For David Harris’s calculations, the frequency results at 60 years were averaged over the 20 year time period from 40 to 60 years while the 25 year estimates were averaged over the first 25 years of operation. It is important that results are consistent (with consistent assumptions and conditions) among each base case team member.

The hot leg results (PWR-1 on page 18 of the handout) indicates large initial uncertainty. Bruce Bishop questioned if the results are for individual welds or the overall system. The response was that the intent was that these results should reflect the frequencies for the overall system. It was pointed out that the hot

leg results should reflect the LOCA frequencies for the hot leg only. The results should not consider all the other lines associated with the RCS, i. e., the cold leg, cross over, surge line, etc.

The slides 21 and 22 (Remaining Work and Differences Among Methodologies) of Rob’s presentation (Base Case Conditions and Summary Results) were not included in the handout and were to be filled out later by the team members. These updates will be posted on the ftp site once available. It was indicated that the LOCA frequencies for the lower leak rate categories (> 100 gpm [380 lpm]) include all the incidences of LOCAs in the higher leak rate categories, e. g., the 100,000 gpm (380,.000 lpm) bin should include all incidences of LOCAs in the 500,000 gpm (1,900,000 lpm) bin.

During a meeting in Rockville (MD) in February 2003 [D.1], the Expert Elicitation Panel members defined five Base Cases that are denoted as BWR-1, BWR-2, PWR-1, PWR-2 and PWR-3, respectively. The five Base Cases are:

BWR Base Case (Plant ‘B’)

• BWR-1; Reactor Recirculation (RR) System. This reference case includes one-of-two RR System loops. Each loop consists of one NPS28 recirculation pump loop with a NPS22 manifold with five NPS12 risers; NPS is nominal pipe size in inch. The reference case excludes any small-diameter piping or tubing attached to the main RR piping. With a few exceptions, the selected piping system layout is representative of a BWR/4 reference plant as described in NUREG/CR-6224 [D.2]. The Base Case RR System does not include the NPS4 bypass line, however. The RR piping is fabricated from austenitic Cr-Ni stainless steel of Type A-304 (> 0.035% carbon).

• BWR-2; Feedwater (FW) System. As defined by isometric drawings, this reference case includes Loop B of the Class 1 portion of the FW System (i. e., the part of the FW System that is located in the drywell containment structure). This system of two loops includes NPS12, NPS14 and NPS20 piping. The FW piping is fabricated from carbon steel of Type A-333 Gr. 6.

• Section D.1.3 includes additional information on the BWR Base Case system definitions.

PWR Base Case (Plant ‘A. a/b’)

• PWR-1; Reactor Coolant (RC) System. As defined by an isometric drawing, this reference case includes one of the NPS30 hot leg (HL) in the RCS.

• PWR-2; Pressurizer Surge Line. As defined by an isometric drawing, this reference case includes the NPS14 piping, which connects the pressurizer to the cold leg (CL).

• PWR-3; High Pressure Injection/Normal Makeup (HPI/NMU) System. As defined by an isometric drawing, this reference case includes the 2-/2 inch schedule 160 line between the containment isolation valve and the RCS cold leg (CL).

• The PWR base cases associated with the RC hot leg and pressurizer surge line are typical of a 3-loop Westinghouse PWR (Plant A. a). The PWR base case associated with the HPI/NMU line is typical of a Babcock & Wilcox PWR (Plant A. b).

• Section D.1.4 includes additional information on the PWR Base Case system definitions.

Dr. Ranganath has spent 30 years working with BWRs. He spent over 28 years working on BWRs at General Electric (GE) before moving to set up a consulting company — XGEN Engineering that provides fracture mechanics, materials and stress analysis services to the power industry. His last position at GE Nuclear Energy was Engineering Fellow and Manager, Hardware Design. He has also taught graduate courses in structural mechanics and materials at Santa Clara University and San Jose State University for over 10 years

Dr. Ranganath has a Ph. D in Engineering from Brown University and a Masters degree in Business Administration from Santa Clara University. He is a Fellow of the ASME. He has also been an Engineering Fellow at GE Nuclear Energy and was elected to the Engineering Hall of Fame at GE Nuclear Energy.

Dr. Ranganath has been active in the development of the ASME Code for over 20 years. He led the effort on developing flaw acceptance rules for austenitic piping in the ASME Code. He was also played a major role in developing improved rules for seismic design of nuclear power plant piping. He has also been the principal investigator on several materials research programs at the Electric Power Research Institute. He has also been active in the BWR Vessel and Internals Program (BWRVIP) and has been the lead author of several Inspection and Evaluation documents for BWR internal components. His expertise in BWR issues such as IGSCC, corrosion fatigue, fracture mechanics, ASME Section XI and Section III Codes, repair hardware design and BWR design is important in assuring that the LOCA frequency conclusions reflect BWR field experience.

Day 1 — February 10, 2004

Rob Tregoning from the USNRC welcomed everyone and explained logistics for the meeting. Rob had everyone introduce themselves. Next, Rob reviewed the agenda for the three-day meeting. Day 1 would focus mainly on piping; Day 2 on non-piping, and Day 3 on emergency and faulted loads plus soliciting feedback on the process. There are no results to present on the topic of emergency and faulted loads.

Only the basic approach will be shown.

Presentation #1 — Elicitation Project Plan, Schedule, and Milestones By Rob Tregoning

The NRC has initiated some ongoing work looking at active mechanisms, e. g., stuck open valves. Bill Galyean is doing this.

There is a SECY paper due to Commissioners on March 31, 2004 with LOCA frequencies for normal operating loads.

Rob will distribute a draft NUREG documenting expert elicitation results so the panel can provide feedback on the NUREG. Rob expects that the panelists will only a have short time (~2 weeks) to provide feedback.

For the April-June public meetings, Bruce Bishop from Westinghouse suggested meeting with the WOG risk-based group.

Most critical future milestone is finalizing individual expert responses for normal operating loading frequencies by February 25th.

Presentation #2 — Elicitation Results (Box and Whisker Plots By Rob Tregoning

Rob made a presentation on the details of the box and whisker plots that will be shown over the next 3 days. Many different methods of calculating percentiles; we used Standard method; fundamental message is that doesn’t make much difference in final analysis.

Shown in Figure D.15 is a general four-state Markov model of piping reliability. All failure processes of this model can be evaluated using service data, assuming that such a data collection is of sufficient technical detail and completeness. This model is used in Section D.6 to develop time-dependent LOCA frequencies.

Figure D. 15 Four-State Markov Model of Piping Reliability[4]

In this and subsequent report sections, a pipe failure (F) is defined as a through-wall defect resulting in a non­active leak or small, active leak. The frequency of a large leak (L) in excess of Technical Specification limits is estimated using the following simple model:

Fl = Л X Plf (d.1)

Frequency of a large leak [1/Reactor-year].

Failure frequency [1/Reactor-year. Extension]; where ‘extension’ refers to the piping component boundary definition. Depending on the intended application and type(s) of degradation mechanism, the extension could be formed by counts of bends, pipes, tees, welds or length of piping. In Equation (4.2), the exposure term reflects the total component population in the data survey.

Exposure time (or reactor operating years)

Conditional probability of a large leak given a through-wall defect. Section D.5 includes a technical basis for estimating conditional failure probabilities.

The parameter estimation uses a Bayes’ update process that begins with the development of prior distributions for each of the terms in equation (D.1). These prior distributions are shaped by our knowledge about the susceptibility of different piping systems to degradation. The input to the Bayes’ update process comes from a small subset of PIPExp after it has been subjected to screening for pipe failures that meet certain selection criteria. A software tool (Bayesian Analysis Reliability Tool — BART™) is used to perform the updates.[5]

Piping reliability is a function of pipe size (diameter and wall thickness), and metallurgy, process medium, environment and design requirements; or attributes and influences, respectively. The purpose of data processing and data reduction is to extract from the total PIPExp database those subsets of service data that correspond to the attributes and influences of the Base Case definitions.

D.4.1.1 Informative versus Noninformative Prior Distributions — The type, extent and quality of applicable service data will determine the actual implementation of the Bayesian update process. Where sufficient service data is available an empirical Bayes approach is used. In this case classical estimation techniques are used to fit a prior distribution to the available data. When no or sparse service data is available a non-informative prior is defined. Relative to the five Base Cases the following approaches are used to determine the prior failure rates distributions:

• BWR Base Case — RR Loop B. There is ample service data on IGSCC. Our prior state-of-knowledge consists of service data before implementation of IGSCC mitigation strategies (mid-1980s). A prior failure rate is derived through classical statistical estimation.

• BWR Base Case — FW Loop B. Given the scarce service data, a lognormal distribution with a mean value of 1.0E-06 per weld-year and range factor (RF) equal to 100 is used. This is a noninfomative prior distribution.

• PWR Base Case — RC Hot Leg. The only available service data involves axial cracks in RPV nozzle-to-safe-end welds at three PWR units. A point estimate for the failure rates is calculated for

the period 1970 through 2000. This point estimate is approximated by a lognormal distribution with range factor of 100; i. e., essentially a noninformative prior.

• PWR Base Case — RC Surge Line. For the pressurizer surge line there is no service data including non-through wall or through-wall cracking. Again, a lognormal distribution with a mean value of 1.0E-06 per weld-year and RF = 100 is used.

• PWR Base Case — HPI/NMU Line. Service data exists, which is directly applicable to this base case. To account for design changes that have been implemented post-1997, a non-informative prior is combined with B&W-specific failure data and exposure data through end of calendar year 1997. The resulting failure rate represents a prior distribution, which is applicable to this Base Case.