ML113180196

Summary of Meeting with STP Nuclear Operating Company Discuss Risk-Informed GSI-191, Assessment of Debris Accumulation on Pressurized-Water Reactor (PWR) Sump Performance, Resolution Approach for South Texas Units 1 and 2

ML113180196
Person / Time
Site:	South Texas
Issue date:	12/05/2011
From:	Balwant Singal Plant Licensing Branch IV
To:	Office of Nuclear Reactor Regulation
Singal, Balwant, 415-3016, NRR/DORL/LPL4
References
TAC ME5358, TAC ME5359
Download: ML113180196 (12)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 5, 2011 LICENSEE: STP Nuclear Operating Company FACILITY: South Texas Project, Units 1 and 2

SUMMARY

OF NOVEMBER 2, 2011, PRE-LICENSING PUBLIC MEETING WITH STP NUCLEAR OPERATING COMPANY HELD VIA CONFERENCE CALL TO DISCUSS THE PROPOSED RISK-INFORMED APPROACH TO THE RESOLUTION OF GSI-191, "ASSESSMENT OF DEBRIS ACCUMULATION ON PWR SUMP PERFORMANCE" (TAC NOS. ME5358 AND ME5359)

On November 2, 2011, a public meeting was held via conference call between the U.S. Nuclear Regulatory Commission (NRC), and representatives of STP Nuclear Operating Company (STPNOC, the licensee), at NRC Headquarters, Rockville, Maryland. The meeting notice and agenda, dated October 11, 2011, is located in the Agencywide Documents Access and Management System (ADAMS) under Accession No. ML112790099. The purpose of the meeting was to discuss the proposed risk-informed approach to the resolution of Generic Safety Issue (GSI)-191, "Assessment of Debris Accumulation on PWR [Pressurized-Water Reactor]

Sump Performance." South Texas Project (STP) is the lead plant and STPNOC plans to submit a license amendment request (LAR) in mid-2012. The licensee previously provided an overview of its proposed approach during the public meetings held on June 2, July 6, July 26, August 22, October 3, and November 1, 2011 1

• The purpose of this conference call was to discuss STPNOC's proposed response to NRC staff comments forwarded to the licensee via e-mail on September 20, 2011 (ADAMS Accession No. ML112630671 and included as Enclosure 1),

associated with developing Loss-of-Coolant Accident (LOCA) initiating event frequencies as part of its assessment as requested by Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors," dated September 13, 2004 (ADAMS Accession No, ML042360586), and uncertainties analysis of the LOCA event frequencies.

The licensee's presentation slides and additional materials provided for the meeting are located at ADAMS Accession No. ML113060594. A list of meeting attendees is provided in Enclosure 2 to this meeting summary.

Summaries of the meetings held on June 2, July 7, July 26, August 22. October 3, and November 1.

2011, are available in ADAMS Accession Nos. ML111640160, ML111950094, ML112130165.

ML112411419, ML112840114, and ML113120129, respectively.

-2 Meeting Summary The licensee provided copies of the following documents prior to the meeting:

1. Technical Review of STP LOCA Frequency Estimation Methodology, dated October 20, 2011 (ADAMS Accession No. ML113060623);

2. Development of LOCA Initiating Event Frequencies for South Texas Project GSI-191, Final Report for 2011 Work Scope, Revision 1, October 2011 (ADAMS Accession No. ML113060615); and

3. STPNOC Proposed Responses to NRC Staff Comments Associated with LOCA Frequency Analysis (ADAMS Accession No. ML113060606).

STPNOC did not prepare any formal presentation slides for this meeting. Items 1 and 2 were provided for reference only and were not specifically discussed during the meeting. However, the licensee made several references to information contained in item 2 during its discussion.

STPNOC provided an overview of the proposed approach to address the NRC staff comments transmitted to the licensee via e-mail on September 20, 2011. The licensee also discussed these comments in Revision 0 and Revision 1 of the reports (ADAMS Accession Nos.

ML112770237 and ML113060615, respectively). The NRC staff expressed the view that STPNOC did not address all the staff's comments sufficiently. However, since the revised report was made available to the NRC staff only a few days before the meeting, the NRC staff was not prepared to comment during the meeting. The majority of the discussion was based on Revision 0 of the report (ADAMS Accession No. ML112770237). The NRC staff indicated that Revision 1 of the report (item 2) will be reviewed at a later date and any additional comments will be addressed in separate correspondence.

The licensee discussed the proposed resolution to the NRC staff comments transmitted on September 20, 2011, during the meeting (ADAMS Accession No. ML112630671 and included as Enclosure 1).

Results of Discussions NRC staff expressed the view that not all its comments have been addressed by the responses received from STPNOC under item 3. The NRC staff offered the following additional comments to STPNOC.

Comment 1:

The response proposed by STPNOC appears to address the NRC staff comments.

Comment 2:

This proposed response, along with the modified report and presentation, appears to provide an acceptable description of the differences in the method described during the June 2 and October 3,2011, meetings.

- 3 Comment 3:

The NRC staff expressed the view that it understands the rationale provided by STPNOC.

However, the NRC staff suggested that an additional scenario be considered and addressed.

That is the potential for a failure of a degraded Class 1 passive system component. Because of the degradation, these components may have less margin than an undegraded component.

This scenario appears to offer the potential for single failures to occur. The NRC staff also suggested another scenario that should also be considered is the failure of one or a limited number of supports which would cause failure within a single system. The support failures could be a result of degradation within the support or higher seismic loads due to the response characteristics of the plant. The NRC staff also suggested that more analysis is needed besides only small break LOCAs and that the entire range of LOCA sizes needs to be considered. The NRC staff suggested that seismic-induced LOCAs be considered in its 2012 report.

STPNOC plans to address this comment in its seismic report scheduled to be prepared in year 2012.

Comment 4:

• The licensee's assertion that an approach accepted previously by the NRC staff for risk informed inservice inspection (RI-ISI) should be a sufficient basis for considering the new application. The NRC expressed the view that these applications are totally different. RI-ISI is used to prioritize and characterize relative risk to determine inspection locations, while this application is attempting to determine frequencies which are then analyzed to determine risk to determine if the emergency core cooling system is acceptable. Because absolute frequencies are developed, this necessarily requires more stringent criteria to provide a reasonable assurance of safety. Also, the response does not provide justification on how the screening criteria "conservatively rule out the potential for damage mechanisms."

STPNOC plans to address this comment later in 2012 evaluations.

• All degradation mechanisms (OMs) identified in the Electric Power Research Institute's (EPRI's) materials degradation matrix (MOM) and NUREG/CR-6923 2 should be considered for each weld joint and/or stress-corrosion cracking (SSC). These studies are more recent and were developed using a broad range of expert opinion. These studies identify possible OMs that have yet to be experienced in service, and hence, are not contained in SCAP-SCC, OPOE, and PIPExp knowledge bases. The NRC staff expressed the view that it is critical to ensure that the OMs identified in these reports are addressed for each specific location.

• The response does not sufficiently address the question related to thermal fatigue (TF) or vibration fatigue (VF). The NRC staff expressed the view that TF or VF (or any other OM) could occur at a specific location that has not previously been experienced in 2 Andresen, P. L., et ai, "Expert Panel Report on Proactive Materials Degradation Assessment,"

NUREG/CR-6923, February 2007, U.S. Nuclear Regulatory Commission, Washington (DC) (ADAMS Accession No. ML070710255).

-4 service and that all OMs identified in EPRl's MOM and NUREG/CR-6923 should be considered. The NRC staff requested STPNOC to justify the uncertainty analysis regarding the TF or VF.

Comment 5:

The NRC expressed the view that this is an issue where the conditional rupture probability (CRP) for weld overlays should be different than for a weld joint containing. for example, TF or just SCC. Currently, none of the CRP models that have been developed include the effect of a weld overlay. Assuming construction defects in the overlay may not be either conservative or realistic, especially if the overlays are not full structural weld overlays. Those overlays do not retain the original code margins for design basis events. It is possible a crack in the original weld could initiate under loading and propagate into and through the overlay. This scenario should be addressed.

STPNOC agreed to address the NRC staff comment.

Comment 6:

A sensitivity analysis to address this comment appears to be reasonable. The NRC staff requested STPNOC to provide more information on how these sensitivity analyses will be conducted.

STPNOC agreed to address this comment in 2012.

Comment 7:

The NRC staff questioned whether the use of Lydell's results as one of the experts' inputs, and then again as an update to the aggregated frequency estimates was not "double accounting" the Lydell results. Licensee representatives explained that it is not clear why Lydell's results should be so heavily weighted and the response does not appear to provide proper justification.

The NRC staff requested the licensee to justify why it is appropriate to combine the system based LOCA frequencies from NUREG-1829 expert judgments with LOCA frequencies from one of Lydell's base case analyses because the base case analyses did not consider all possible LOCA-contributing factors (Le., OMs, mitigation measures, stresses) while the 1829 results did consider them. The scope of the Lydell base case and the 1829 results are different.

STPNOC stated that the comment was addressed in Revision 1 of the report. The NRC staff indicated that it will review Revision 1 of the report and update its comment, as needed.

STPNOC agreed to address the comment after receiving revised input.

Comment 8:

The revised report and the response provided by STPNOC clarify the analysis and address the original question. However, the NRC staff expressed the view that the justification for assuming that the CRP models that are developed, are appropriate for the systems that they are developed for, given that other OMs and LOCA-contributing factors are active in these systems and have not been considered in the base case analyses. More importantly, the applicability of

- 5 these CRP models to other systems that were not included in the base case analyses needs to be justified.

STPNOC agreed to provide additional justifications in support of the NRC comment.

Comment 9:

• The first part of the question seems clarified appropriately in the revised report.

• The NRC staff comment in second part of the question still needs to be addressed. The NRC staff believes that the experts did not provide results for every system. The experts were not required to consider systems that they did not believe contribute to greater than 20 percent of the total LOCA frequencies. However, just because an expert did not provide an answer does not mean that his results contain no information and it is not appropriate to combine the results from the remaining experts to get a new distribution.

For example, the given system might contribute only 19 percent of the expert's estimate so that he did not quantify the results for that system. However, that percentage could be a significant percentage of the estimates provided by the other experts for that system. By ignoring the testimony of other experts, it may potentially skew or bias the community distributions developed for all the other systems that did have explicit results from all nine experts.

STPNOC expressed the view that the comment has been addressed in Revision 1 of the report.

The NRC staff agreed to review Revision 1 of the report and provide additional feedback to STPNOC. STPNOC agreed to review the revised input and address NRC staff comments.

Comment 10:

This response appears to address the NRC staff comment but the approach used for considering these types of ruptures in 2012 will be of interest to the NRC staff.

Comment 11:

Please refer to the earlier comment (Comment 4) about the acceptability of the RI-ISI review for this particular application in Comment 4. STPNOC should summarize the technical basis for the range factors that is contained in EPRI'S topical report (TR)-111880, "Piping System Failure Rates and Rupture Frequencies for Using Risk Informed In-Service Inspection Applications,"

September 1999. The development of the means using global industry service experience data was clarified in the revised report.

STPNOC agreed to summarize the technical basis for the range factors contained in the EPRI report in the next revision of the STPNOC report.

Comment 12:

The presentation, revised STPNOC report, and this response clarified the differences between these two aggregation methods. However, the NRC staff expressed the view that the results of

.6 a sensitivity analysis using both methods should be provided to demonstrate the sensitivity of the results to these two, very different, aggregation methods.

STPNOC agreed to include the results of a sensitivity analysis using both methods and discussion on differences in the revised report.

In summary, STPNOC agreed to address the following resulting from discussion during the meeting:

1. STPNOC will provide additional detail on the following items related to the DMs:

• Address any potential impacts due to changes in the operating experience database.

• Discuss how the EPRI screening criteria rule out DMs.

• Discuss how more recent work on DMs was utilized.

• Address how STPNOC plans to address unknown DMs.

2. Since the failure rates for the piping with weld overlay would be different than an original defect failure rate, STPNOC agreed to discuss how piping with overlay failure rates was taken into consideration.

3. STPNOC to provide additional justification explaining why the approach to combining the results of the community distribution and the Lydell base case results is appropriate.

Also, STPNOC will address the question of whether the base case results addressed all the relevant DMs. STPNOC also agreed to discuss how the possible dependence of CRP on damage mechanisms was taken into account.

4. STPNOC to provide more information on the quantitative difference of using the worst case percentile method vs. the mixture distribution method on the CRP results.

The NRC staff also agreed to provide its comments to STPNOC on the factors important to chemical effects that were discussed in the white paper during the meeting. The staff agreed to forward its comments to the licensee prior to the conference call public meeting scheduled for December 1 2011.

J General Comments

• The majority of the discussion during the conference call was focused on the comments and related issues discussed above. STPNOC plans to address the open issues in next revision to the report.

• It may be noted that this is a pre-licensing action and there is no formal application submitted to the NRC staff for review. The results of the discussions reflect the staffs views and in no way constitute approval of the proposed methodology. A formal and

-7 detailed review will be performed by the NRC staff after the formal application by STPNOC and may result in additional NRC staff comments and questions.

No Public Meeting Feedback Forms were received for this meeting.

Please direct any inquiries to me at (301) 415-3016, or balwant.singal@nrc.gov.

Sincerely, bc._\~~~~

Balwant K. Singal, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499

Enclosures:

1. September 20, 2011 E-Mail

2. List of Attendees cc w/encl: Distribution via Listserv

From: Singal, Balwant Sent: Tuesday, September 20, 2011 4:05 PM To: Paul, Jamie Cc: Grantom, Carl; Bailey, Stewart; Tregoning, Robert; Tsao, John; Thadani, Mohan

Subject:

U. S. Nuclear Regulatory Commission (NRC) Staff Questions - Topic of Loss-of Coolant Accident (LOCA) Frequency Analysis - Risk-Informed Approach to GSAI-191, TACs ME5358 and ME5359 Attachments: Questions Associated with LOCA Frequency Analysis.docx Jamie:

Based on our discussions this morning attached are the NRC staff questions on the topic of LOCA Frequency Analysis for the proposed Risk-Informed Approach for resolution to GSI-191 issues The NRC staff expects that STP Nuclear Operating Company will address these questions during the forthcoming presentation during the conference call public meeting on October 3, 2011. Please let me know if you have questions.

I will be out-of-office next week. Please call my backup Project Manager, Mohan Thadani at 301-415-1476 during my absence.

Thanks.

Balwant K. Singal Senior Project Manager (Comanche Peak and STP)

Nuclear Regulatory Commission Division of Operating Reactor Licensing Balwant.Si ngal@nrc.gov Tel: (301) 415-3016 Fax: (301) 415-1222 Enclosure 1

Questions Associated with Loss-of-Coolant Accident (LOCA) Frequency Analysis

1. Is there an updated document since presentation in July that fully describes the LOCA frequency estimation and provides an example with the latest approach for calculating these frequencies?

Please explain.

2. Can the differences between the method for calculating frequencies in the July meeting be compared to what's currently being done be summarized and rationale provided for the reason for the change? Please explain.

3. Please explain how are LOCAs due to seismic effects considered?

4. Please explain how at South Texas Project (STP, Units 1 and 2 the modeled degradation mechanisms (OMs) at each joint are comprehensive if there is no prior history of that mechanism at a piping location, or the particular piping location has never been inspected? Is this rectified with the Risk-Informed Inservice Inspection (RI-ISI) expert judgment process at each plant? Have the OMs modeled for each joint been compared with more expansive studies such as Electric Power Research Institute's (EPRls) materials degradation matrix (MOM) and the NRC's proactive materials degradation assessment (PMOA) (Le., NUREG!CR-6923) assessments to ensure assessment of all possible mechanisms? How do you know that certain mechanisms (i.e., thermal fatigue (TF) and vibration fatigue (VF)) can't occur at a specific location?

S. Please explain how mitigation of primary water stress corrosion cracking (PWscq is considered in the analysis to determine LOCA frequencies at STP, Units 1 and 2, especially for hot legs and cold legs as no mitigation methods have been identified for these systems in the information provided to date?

6. Currently, each OM at a specific piping location is combined independently to determine the LOCA frequency distribution for that weld joint location. Please explain how are the synergistic effects of multiple OMs on either the failure frequency or the conditional probability of rupture considered? For example, TF could initiate at a design and construction (O&C) flaw and possibly combine with PWSCC to accelerate the degradation rate compared to each OM rate considered independently so that the failure frequency is increased. Also, the degradation may evolve in a manner that affects the likelihood of rupture.

7. The approach for determining LOCA frequencies combines Lydell's base case results from NUREG-1829 with the distributions from all the experts. However, Lydell's estimates are part of the community of results. Doesn't this approach double-count Lyden's estimates? Please explain why is it appropriate to combine LydeU's estimates for specific OMs with the total LOCA frequency results (i.e., from all OMs) in NUREG-1829?

8. Please explain how were Lydell's estimates for multiple base cases added to form a single distribution? How were systems and OM that weren't part of Lydell's base case estimates analyzed?

9. Please more fully explain how composite LOCA frequency distributions are developed from the individual expert LOCA frequency distributions from NUREG-1829? How situations where the expert didn't provide any information for a particular system in NUREG-1829 are addressed, since the experts weren't required to assess all systems (only those that they thought were most important were assessed)?

10. Please verify that debris from vessel rupture is not considered in the GSI-191 analysis to determine plant debris sources, but is part of the general probabilistic risk assessment (PRA) modeling that is unaffected by plant changes to address GSI-191 resolution? If so, why is not considering vessel rupture as a debris source term appropriate?

11. Please more fully explain how the prior distributions in Step 4 (Le., slide 6 in the presentation dated 8/22/11 for pre-licensing meeting) are determined?

12. Please more fully explain the differences between the worst-case percentile method and the mixture distribution method (Le., slides 11- 13 in presentation dated 8/22/11 for pre-licensing meeting)?

LIST OF ATTENDEES FOR MEETING WITH STP NUCLEAR OPERATING COMPANY REGARDING RISK-INFORMED APPROACH TO RESOLUTION OF GSI-191 ISSUE SOUTH TEXAS PROJECT, UNITS 1 AND 2 OCTOBER 3, 2011 NAME TITLE ORGANIZATION Paul Jamie* Licensing STPNOC Steve Blossom* Project Manager, Special Projects STPNOC Ernie Kee* Risk Management STPNOC i Tim Sande* Principal Engineer Alion Science and Technology, I Represented STPNOC Karl Fleming* Consultant KNF Consulting, Represented STPNOC Phillip Grissom* - Southern Nuclear Company Mike MacFarlane* - Southern Nuclear Company John Tsao Materials Engineer NRC Michael Snodderly Senior Reliability and Risk Engineer NRC Balwant K. Singal Senior Project Manager NRC I Robert Tregoning* Senior Technical Advisor for NRC Materials Engineering Issues Stephen Dinsmore Senior Reliability and Risk Engineer NRC Stewart Bailey Branch Chief NRC Steve Smith Reactor Systems Engineer NRC

• Participated via phone NRC - U.S. Nuclear Regulatory Commission STPNOC - STP Nuclear Operating Company Enclosure 2

ML113180196 *Via E-mail OFFICE NRR/LPL4/PM NRR/LPL4/LA NRR/DSS/SSIB/BC NAME BSingal JBurkhardt SBailey*

DATE 11/17/11 11/15/11 11/18/11 OFFICE NRR/DRAIAPLAlBC NRR/LPL4/BC NRR/LPL4/PM NAME SDinsmore(A)* MMarkley BSingal DATE 11/17/11 12/5111 12/5/11