ML073170039

Request for Additional Information Related to the Containment Integrated Leak Rate Testing (ILRT) Interval Extension

ML073170039
Person / Time
Site:	Point Beach
Issue date:	11/15/2007
From:	Jack Cushing NRC/NRR/ADRO/DORL/LPLIII-1
To:	Mccarthy J Florida Power & Light Energy Point Beach
Cushing, Jack /NRR/DORL/LPL3-1, 415-1424
References
TAC MD7013, TAC MD7014
Download: ML073170039 (6)

Text

November 15, 2007 Mr. James McCarthy Site Vice President FPLE Point Beach 6610 Nuclear Road Two Rivers, WI 54241-9516

SUBJECT:

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE CONTAINMENT INTEGRATED LEAK RATE TESTING (ILRT) INTERVAL EXTENSION FOR POINT BEACH, UNITS 1 AND 2 (MD7013 AND MD7014)

Dear Mr. McCarthy:

The purpose of this letter is two-fold, (1) to request additional information so that the staff can complete its review, and (2) to request your assistance relating to planning of your future licensing requests submitted to the Nuclear Regulatory Commission (NRC).

FPL Energy Point Beach, LLC (FPLE-PB), submitted a license amendment request dated October 12, 2007, to revise Point Beach Nuclear Plant, Units 1 and 2, Technical Specification (TS) for containment integrated leak rate testing (ILRT) program in TS 5.5.15, Containment Leak Rate Testing program. This requested change would allow a one-time interval extension of no more than 5 years for the ILRT.

FPLE-PB requested a completion date by March 1, 2008. This requested completion date allows approximately 4.5 months for us to complete our review. The NRC has established performance goals for licensing actions to complete 96 percent in less than 1 year and 100 percent in less than 2 years. Consistent with these performance goals, we will endeavor to complete our review of your October 12, 2007, license amendment request as expeditiously as possible. However, based on the technical issues that need to be resolved, we may not complete our review by March 1, 2008. The reason it is unlikely that the review can be completed in a shorter than normal review schedule, is due to the technical issues associated with this review. The technical issues include the need to perform a sensitivity study on steam generator tube rupture, evaluation of your efforts to reduce the large early release frequency (LERF) below the guidance in Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis.

These issues and others are discussed in more detail in the enclosure.

We request your assistance to ensure that your licensing requests are well planned, taking into consideration your needs as well as the NRCs performance goals.

The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on November 13, 2007, it was agreed that you would provide a response by December 14, 2007.

J. McCarthy The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-1424.

Sincerely,

/RA/

Jack Cushing, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301

Enclosure:

Request for Additional Information cc w/encl: See next page

J. McCarthy The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-1424.

Sincerely,

/RA/

Jack Cushing, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266 and 50-301

Enclosure:

Request for Additional Information cc w/encl: See next page DISTRIBUTION:

PUBLIC LPL3-1 R/F RidsNrrDorlLpl3-1 RidsNRRPMJCushing RidsNrrLATHarris RidsAcrsAcnwMailCenter RidsOgcRp RidsRgn3MailCenter RidsNrrDorlDpr MRubin RPalla ADAMS Accession Number: ML073170039 OFFICE LPL3-1/PM LPL3-1/LA DRA/APLA/BC LPL3-1/(A)BC NAME JCushing THarris MRubin CMunson DATE 11/14/07 11/13/07 11/14/07 11/15/07 OFFICIAL RECORD COPY

Point Beach Nuclear Plant, Units 1 and 2 cc:

Licensing Manager Mr. R. S. Kundalkar FPL Energy Point Beach, LLC Vice President 6610 Nuclear Road Nuclear Technical Services Two Rivers, WI 54241 FPL Energy, LLC P. O. Box 14000 Mr. Ken Duveneck Juno Beach, FL 33408-0420 Town Chairman Town of Two Creeks J. Kitsembel 13017 State Highway 42 Electric Division Mishicot, WI 54228 Public Service Commission of Wisconsin P. O. Box 7854 Resident Inspector's Office Madison, WI 53707-7854 U.S. Nuclear Regulatory Commission 6612 Nuclear Road Mr. M. S. Ross Two Rivers, WI 54241 Managing Attorney FPL Energy, LLC Chairman P. O. Box 14000 Public Service Commission of Wisconsin Juno Beach, FL 33408-0420 P.O. Box 7854 Madison, WI 53707-7854 Mr. J. A. Stall Senior Vice President and Chief Nuclear Officer FPL Group P. O. Box 14000 Juno Beach, FL 33408-0420 Mr. Antonio Fernandez Senior Attorney FPL Energy, LLC P. O. Box 14000 Juno Beach, FL 33408-0420 Mr. Mark Warner Vice President Nuclear Operations, North Region FPL Energy, LLC P. O. Box 14000 Juno Beach, FL 33408-0420

REQUEST FOR ADDITIONAL INFORMATION POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 REGARDING THE INTEGRATED LEAK RATE TESTING (ILRT) INTERVAL EXTENSION DOCKET NOS. 50-266 AND 50-301 In reviewing the FPLE Point Beach, LLC=s submittal dated October 12, 2007, the Nuclear Regulatory Commission (NRC) staff has determined that the following information is needed in order to complete its review:

1. The requested ILRT extension is estimated to result in an increase in large early release frequency (LERF) (for internal events) of approximately 3E-7 per year. In accordance with Regulatory Guide 1.174, such applications will be considered only if it can be reasonably shown that the total LERF is less than 1E-5 per year. As indicated in the license amendment request (LAR), the combined internal and external events LERF for the baseline probabilistic risk assessment PRA is approximately 2E-5 per year and exceeds the risk acceptance guideline. A further assessment of induced steam generator tube rupture (ISGTR) events in Attachment C-1 of the LAR indicates that the total LERF could be reduced below the acceptance guideline if the conditional probability of an ISGTR of 0.25 used in the baseline PRA (obtained from NUREG-1570 and applied generically to all sequences with high reactor coolant system pressure and dry steam generators to arrive at the frequency of an ISGTR), is replaced by a plant-specific assessment of the frequencies of the various sequences leading to high/dry conditions at Point Beach (particularly high/dry sequences with concurrent reactor coolant pump seal loss-of-coolant accident (LOCA)) and their respective conditional failure probabilities.

However, this analysis continues to rely on conditional failure probabilities and steam generator flaw distributions from NUREG-1570, which is nearly 10 years old, and fails to reflect additional information developed subsequent to NUREG-1570. More recent NRC-sponsored thermal-hydraulic studies to evaluate steam generator tube integrity suggest that the conditional probability on an ISGTR could be higher than considered in NUREG-1570, i.e., ISL draft reports ASCDAP/RELAP Base Case Calculation for the Station Blackout Uncertainty Study,@ August 2006 (ML070220062), and AEvaluation of Uncertainties in SCDAP/RELAP5 Station Blackout Simulations,@ August 2006 (ML070220056). Discuss the applicability and implications of the aforementioned studies on the Point Beach LERF assessment. Provide additional assessments, as appropriate, to illustrate the sensitivity of the total LERF estimate to higher conditional probabilities of ISGTR as might be inferred from the more recent studies.

2. Provide a description of the reactor coolant pump seal LOCA model used in the Point Beach PRA. The NRC staff has approved the use of the AWOG 2000 RCP Seal Leakage Model for Westinghouse PWRs@ described in WCAP-15603, Revision 1 (ML032040132). If a different model has been used in the Point Beach PRA, provide an assessment of the impact on core damage frequency and total LERF of using the approved seal leakage model.

ENCLOSURE

3. Confirm that the treatment of the turbine-driven auxiliary feedwater pump in the PRA on which the LAR is based, is consistent with plant-specific operating experience to-date with this system. Discuss the recent operating experience and how it has been reflected in the PRA.

4. Provide a breakdown of the key contributors to the total LERF estimate for internal and external events. Discuss the degree of realism, conservatism, or non-conservatism associated with each key contributor and the rationale for this characterization, including the underlying models and assumptions that may contribute to this view. Provide an estimate of the extent to which the total LERF would be impacted if other areas of the analysis (besides ISGTR) were treated more realistically.

5. Describe the internal and external industry/peer reviews of the Point Beach Level 2/LERF model (or its predecessor), the findings from these reviews, and the actions taken to address the peer review findings. Specifically address any comments that relate to the conservatism or non-conservatism of key models or assumptions that impact the total LERF, and how these comments have been addressed within the PRA version used to support the LAR.

6. In the discussion regarding very small increases in LERF, Regulatory Guide 1.174 (Section 2.2.4) notes that if there is indication that total LERF may be considerably higher than 1E-5 per year, the focus should be on finding ways to decrease LERF. Discuss any activities planned or in progress at Point Beach to reduce the total LERF for the plant.