ML083250719

Request for Additional Information Regarding Severe Accident Mitigation Alternatives for Kewaunee Power Station

ML083250719
Person / Time
Site:	Kewaunee
Issue date:	01/08/2009
From:	Sarah Lopas License Renewal Projects Branch 1
To:	Christian D Dominion Energy Kewaunee
Lopas, S, DLR/REBA - 415-1147
References
TAC MD9409
Download: ML083250719 (11)

Text

January 8, 2009 Mr. David A. Christian Dominion Energy Kewaunee, Inc.

President and Chief Nuclear Officer Innsbrook Technical Center - 2SW 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING SEVERE ACCIDENT MITIGATION ALTERNATIVES FOR KEWAUNEE POWER STATION (TAC NO. MD9409)

Dear Mr. Christian:

The U.S. Nuclear Regulatory Commission staff has reviewed the Severe Accident Mitigation Alternatives analysis submitted by Dominion Energy Kewaunee, Inc., regarding its application for license renewal for Kewaunee Power Station, and has identified areas where additional information is needed to complete its review. Enclosed is the staff=s request for additional information.

We request that you provide your responses to these questions within 60 days of the date of this letter, in order to maintain the environmental review schedule. If you have any questions, please contact me at 301-415-1147 or by e-mail at sarah.lopas@nrc.gov.

Sincerely,

/RA/

Sarah Lopas, Project Manager Projects Branch 1 Division of License Renewal Office of Nuclear Reactor Regulation Docket Nos. 50-305

Enclosure:

As stated cc w/encl: See next page

ML083250719 OFFICE LA:DLR PM:DLR:RPB1 BC:DLR:RPB1 NAME IKing SLopas DPelton DATE 11/21/08 11/21/08 01/08/09 Letter to D. Christian from S. Lopas, dated January 08, 2009 DISTRIBUTION:

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING SEVERE ACCIDENT MITIGATION ALTERNATIVES FOR KEWAUNEE POWER STATION (TAC NO. MD9409)

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REQUEST FOR ADDITIONAL INFORMATION REGARDING THE ANALYSIS OF SEVERE ACCIDENT MITIGATION ALTERNATIVES FOR KEWAUNEE POWER STATION (KPS)

1. Provide the following information regarding the Probabilistic Risk Assessment (PRA) models used for the Severe Accident Mitigation Alternative (SAMA) analysis:

a. The first paragraph of Section F.2.1 states that logic changes were made to the Level 1 model to address internal flooding related design changes planned for completion prior to the license renewal period. Describe these design and logic changes.

b. The last paragraph on page F-8 indicates that a proposed change to elevate supply breakers would be scheduled in the future. This design change was apparently credited in the current PRA. Another change, re-routing a wire to the turbine building fan coil unit, has apparently been made but not included in the PRA used for the SAMA analysis.

However, a related discussion in Section F.7.6 implies that at least a portion of the planned breaker modification has been made. Provide additional details regarding design changes, the associated PRA models, and the estimated date for the breaker modification, if it is still planned.

c. On page F-9, it is stated that station blackout (SBO) contributes 13.6% of the core damage frequency (CDF), while in Items 16 and 29 (and others) of Table F-3 it is stated that SBO contributes 4.3% of the CDF. Confirm which value is correct.

d. The CDF increased by a factor of 24 from the 8/2003 model to the 12/2004 model and then decreased by a factor of almost 10 in the K101AASAMA model, all subsequent to the Westinghouse Owners Group (WOG) peer review. Discuss the major reasons for the large increase and subsequent decrease in CDF, with particular attention to the evolution of the internal flooding model.

e. One of the unresolved WOG Peer Review Fact and Observations (F&Os) is related to not treating loss of ventilation as a unique initiating event. The discussion of this F&O (IE-1) in Table F-5 indicates that manual shutdown may be required for loss of certain ventilation systems and that these events are subsumed in the reactor trip with main feedwater initiating event. This latter initiating event will presumably have all HVAC initially operating normally rather than having a failure that caused the manual shutdown, and the likelihood of a random HVAC failure during this event would be small. Justify that treatment of a loss of ventilation initiating event in this manner is appropriately bounding, and would not adversely impact the identification of HVAC-related SAMAs.

f. KPS License Amendment Request 242 of September 11, 2008 provides information on the K107Aa PRA model of July 15, 2008, which post-dates the PRA version used for the SAMA analysis. The CDF and large early release frequency (LERF) reported therein are approximately half of the values in the SAMA PRA. An independent assessment of the K107Aa PRA against the supporting requirements of the ASME PRA standard was also briefly described.

i. Provide the principal reasons for the reduction in CDF from the SAMA PRA to the K107Aa PRA, and address the impact of these changes on the SAMA analysis.

ii. Identify who performed the independent assessment and discuss the impact that any unmet supporting requirements might have on the SAMA analysis.

iii. Confirm whether a review of the importance analysis for the K107Aa model leads to the identification of any additional potentially cost-beneficial SAMAs.

g. In a June 17, 2005 submittal on risk-informed in-service inspection, NMCs response to RAI 3.7 indicated that 6 weaknesses were identified in the IPE. Confirm that none of these items remain applicable to the PRA used for the SAMA assessment.

h. Provide a more detailed description of the Level 1 and 2 PRA update process, the quality control of PRA model changes, and the independent review and approval of the PRA model update documentation mentioned at the end of Section F.2.5 (including scope of review, independence of reviewers, and documentation of review comments).

i. The contributions to CDF by initiating event given in Table F-1 total only 77% of the CDF.

Characterize the remaining 23% as to initiator or initiator type and any noteworthy attributes.

2. Provide the following information relative to the Level 2 PRA analysis:

a. Section F.2.4 states that the Level 2 model was developed for the Individual Plant Examination (IPE) and updated in 2004 and 2007, and describes the changes made in the 2007 update. Describe the nature of the changes made in the 2004 update beyond those described in Section F.2.4.1, if any.

b. Section F.2.4 mentions the use of bridge trees. Describe the bridge trees. Confirm whether they are separate event trees that link to the Level 1 trees or are bridge events incorporated directly into the Level 1 trees. Indicate whether they are quantified by direct linking or by binning.

c. Describe any changes made to the definition and development of plant damage states subsequent to the IPE.

d. Section F.2.4 states that, with one exception, the Modular Accident Analysis Program (MAAP) case selected to be representative for each release category was the same as for the IPE. The risk profile is much different now than in the IPE, for example, LOCCW -

IPE < 1%, now 8%; SLOCAs - IPE 21%, now 2%; SBO - IPE 40%, now 14%. Provide further discussion and justification for the selection of the representative MAAP case for each release category.

e. The release fractions for several nuclides for source term categories (STCs) 11 and 12 are reversed between Tables F-6 and F-10. Confirm which values are correct.

f. Tables F-6 and F-10 indicate a zero release fraction for STCs 1 and 8. Even though these STCs may involve an intact containment, there will be some release to the environment due to normal leakage. Justify that omitting this contribution to total risk does not impact the results of the SAMA evaluation.

3. Provide the following information regarding the treatment of external events in the SAMA analysis:

a. Section F.2.3.1 summarizes several conservatisms in the fire PRA model. Indicate the fire zone(s) to which each conservatism is applicable.

b. Section F.2.3.1 states that an assessment of the effects of plant procedure changes shows that the CDF would be reduced by a factor of 5 and that a more appropriate fire CDF would be 3.6 E-5. Discuss in more detail the assessment of procedure changes and the impact of the changes on the CDF for each of the fire zones listed in Table F-22.

c. The individual plant examination of external events (IPEEE) safety evaluation report (SER) indicates that the protection of the underground diesel oil storage tank vents against tornado missiles is an open item. Confirm that this has been resolved, or address the implications for the SAMA analysis.

d. Table 2.12 of NUREG-1742 indicates that Kewaunee had the potential for adverse seismic-fire interactions due to the presence of mercoid switches in the fire jockey pump and the Cardox system. Confirm that this has been resolved, or address the implications for the SAMA analysis.

e. Although Table F-17 includes SAMAs for external events based on generic insights, the plant-specific fire and seismic risk results do not appear to have been systematically reviewed for the purpose of identifying potential external event SAMAs.

i. For each of the major fire risk contributors at KPS, provide an evaluation demonstrating that there are no viable SAMA candidates that would further reduce the fire risk. Address the impact of the weaknesses in the fire analysis (as identified in the IPEEE SER/technical evaluation report (TER)) on this evaluation.

ii. For each of the major seismic risk contributors at KPS, provide an evaluation demonstrating that there are no viable SAMA candidates that would further reduce the seismic risk. Address the impact of the weaknesses in the seismic analysis (as identified in the IPEEE SER/TER) on this evaluation.

4. Provide the following information relative to the Level 3 PRA analysis:

a. Provide additional information on how the population growth rates and the transient population data were developed, including the source of the county growth rates, how the growth rate estimates were applied, and how growth was estimated for the transient population.

b. The base case analysis assumes all releases occur at the top of the containment with an ambient thermal content. Demonstrate that the resulting population doses bound those expected for a steam generator tube rupture (SGTR) with failure of secondary side isolation, which is the dominant contributor to population dose at KPS.

c. The core radionuclide inventory is stated as being based on an end-of-cycle ORIGEN2 analysis for KPS. Confirm that this core inventory reflects the expected fuel management/burnup during the license renewal period.

d. Describe the methodology and data sources used to fill in any gaps in the onsite meteorology data.

5. Provide the following information with regard to the selection and screening of Phase I SAMA candidates:

a. For Item 2 in Table F-3 (LOSP-24, Loss of all power from the grid during 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />), it is stated that the ability to isolate flooding events without requiring power would greatly lower the importance of this event and that SAMA 168 (Provide the ability to manually close electrically operated valves needed to isolate flooding events) is applicable. The Fussell-Vesely value for LOSP-24 is 0.1793. However, the evaluation of SAMA 168 in Section F.6.31 resulted in only a 1% reduction in CDF. Explain why the impact of this SAMA is so small. Identify and discuss alternative SAMAs that might be more effective in addressing this important risk contributor.

b. For items 22, 23 and 35 (and others) in Table F-3, adding a refueling water storage tank (RWST) low level alarm and/or an automatic refilling system for the RWST could potentially reduce dependency on prior action or eliminate the need for the operator to refill the RWST. Provide an evaluation of these alternative SAMAs.

c. In several places in Table F-17 (SAMAs 7 and 30, for example), the SAMA is stated to be already implemented, but the basis for this statement (e.g., citation of a specific procedure change) is not cited. Provide the basis for the statement that the SAMA is already implemented for all SAMAs where no citation is currently provided.

d. SAMA 10 (Revise procedure to allow bypass of diesel generator trips) is stated in Table F-17 to be of very low benefit based on review of only 8 months of EDG failure data (January 2001 through August 2001). Justify that this is enough data to exclude trip circuitry as a cause of EDG unavailability.

e. The potential enhancement for SAMA 64 involves either implementing procedure and hardware modifications to allow manual alignment of the fire water system to the component cooling water (CCW) system or installing a cooling water header cross-tie.

Table F-17 indicates that this SAMA is already implemented, apparently on the basis that the system is normally cross-tied. Confirm that the CCW system can be manually cross-tied to the fire water system or, if this capability does not exist, evaluate its addition as a potential SAMA.

f. It is stated in Table F-17 (for SAMA 75) that KPS does not have a diesel-driven fire pump. Discuss the potential benefits (in both internal events and fire events) of adding a diesel-driven fire pump at KPS.

g. For SAMA 144 (Install additional transfer and isolation switches), it is noted in Table F-17 that spurious actuations do not contribute to fire CDF since no credit is taken for equipment that is not specifically analyzed to survive a fire. It is not clear how not taking credit for this equipment reduces the importance of spurious actuations. Provide further justification for screening out this SAMA or consider appropriate plant improvements.

h. SAMA 151 (Increase training and operating experience feedback to improve operator response) is dispositioned in Table F-17 as needing further evaluation. However, it is not included among the SAMAs that were further evaluated (as listed in Table F-19). Also, the comments in the column Results of Potential Enhancement for this item refer to Tables 5 and 6, but no such tables are provided in the ER. Clarify the disposition of SAMA 151.

6. Provide the following information with regard to the Phase 2 cost-benefit evaluations:

a. Table F-19 indicates that implementation of SAMA 19 (Use fire water as a backup source for diesel cooling) would result in an increase in CDF. Explain why this occurs.

b. The discussion in Section F.6.15 of SAMA 76 (Change failure position of condenser makeup valve so that the valve fails closed on loss of power or air) indicates that this SAMA was modeled by removing the power dependencies from the valve. Clarify whether this included removing its dependence on air. If not, incorporate the removal of this dependency or justify why it would not impact the results.

c. As indicated in Sections F.6.17 (Diesel Room Cooling Improvements) and F.6.18 (Switchgear Room Ventilation Response), the evaluations of SAMA 81 and SAMA 82 assume implementation of a number of other SAMAs, including SAMAs 170 and 171.

Based on Table F-17, the latter two SAMAs are plant-specific improvements that pertain to improving room cooling for the safeguards alley. Explain why SAMAs 170 and 171 have been combined with SAMAs 81 and 82 and why a SAMA involving implementation of SAMAs 170 and 171 for just the AFW rooms was not evaluated.

d. Section F.6.30 indicates that the benefit of SAMA 150 (Improved maintenance procedures) was determined by setting maintenance unavailability of Maintenance Rule (a)(1) equipment to zero. This approach appears to reduce the risk due to maintenance unavailability rather than the risk due to any improvement in equipment reliability.

Provide additional information supporting this evaluation.

7. Provide the following information with regard to the sensitivity and uncertainty analyses:

a. On page F-93 it is stated that 12 additional analyses representing 5 SAMA items would show potentially positive cost-benefit if a discount rate of 3% was used. It appears that

use of the 3% discount rate resulted in identification of 12 rather than 5 additional cost-beneficial SAMAs. Clarify this reference to representing 5 SAMA items.

b. The discussion in Section 7.1 of SAMA 58 (Replacement of existing reactor coolant pump (RCP) seals with seals that do not require any seal cooling) describes added costs for changing the seal cooling system for the new seals. This cost should be minimal since the new seals would not require cooling. The discussion of this SAMA in Section F.7.1 states that the added cost would be over $750,000 whereas the discussion in Section F.7.2 states that the added cost would be over $500,000. Clarify the cost estimates for this SAMA.

c. The listing of SAMAs on page F-100 does not include SAMA 58, which had a negative net value in Table F-19 but a positive net value in Table F-20. Provide the results of the evaluation of this SAMA in the listing and in the subsequent discussion.

d. Section F.7.7 discusses the simultaneous implementation of SAMAs 81, 82, 83, 166, 167, 170 and 171. SAMA 160 is not included in the Section F.7.7 discussion but is included in the individual discussion in Sections F.6.17. Clarify which changes in the diesel generator room and switchgear room are included in the combined package.

8. For certain SAMAs considered in the Environmental Report, there may be lower-cost alternatives that could achieve much of the risk reduction at a lower cost. In this regard, provide an evaluation of the following SAMAs:

a. Automate the cross-tie of the existing condensate storage tank (CST) to other water sources rather than installing a new CST.

b. Modify procedures to direct primary system cooldown to further reduce the probability of RCP seal failures.

c. Modify procedures and equipment for using a portable diesel-driven or AC-powered pump to provide feedwater to the steam generators with suction from the intake canal.

d. Develop a procedure to cross-connect the chemical and volume control system (CVCS) holdup tanks to the volume control tank (VCT) through the CVCS holdup transfer pump.

9. Section F.8.2, indicates that SAMAs 81, 160, 166 and 167 may also be cost beneficial if implemented concurrently with other SAMAs. This would bring the total number of SAMA candidates for further evaluation to 18. Confirm that these additional four SAMA candidates will be further reviewed as part of Dominions ongoing performance improvement program.

Kewaunee Power Station cc:

Resident Inspectors Office Mr. Paul C. Aitken U.S. Nuclear Regulatory Commission Supervisor - License Renewal Project N490 Highway 42 Innsbrook Technical Center - 3NE Kewaunee, WI 54216-9510 5000 Dominion Boulevard Glen Allen, VA 23060-6711 Mr. Chris L. Funderburk Director, Nuclear Licensing and Mr. David A. Sommers Operations Support Supervisor - Nuclear Engineering Dominion Resources Services, Inc. Innsbrook Technical Center - 2SE Innsbrook Technical Center - 2SE 5000 Dominion Boulevard 5000 Dominion Boulevard Glen Allen, VA 23060-6711 Glen Allen, VA 23060-6711 Ms. Lillian M. Cuoco, Esq.

Mr. Thomas L. Breene Senior Counsel Dominon Energy Kewaunee, Inc. Dominion Resources Services, Inc.

Kewaunee Power Station 120 Tredegar Street N490 Highway 42 Riverside 2 Kewaunee, WI 54216 Richmond, VA 23219 Mr. Michael J. Wilson, Director Mr. Stephen E. Scace Nuclear Safety & Licensing Site Vice President Dominion Energy Kewaunee, Inc. Dominion Energy Kewaunee, Inc.

Kewaunee Power Station Kewaunee Power Station N490 Highway 42 N490 Highway 42 Kewaunee, WI 54216 Kewaunee, WI 54216 Mr. William R. Matthews Mr. David R. Lewis Senior Vice President - Nuclear Operations Pillsbury Winthrop Shaw Pittman, LLP Innsbrook Technical Center - 2SE 2300 N Street, N.W.

5000 Dominion Boulevard Washington, DC 20037-1122 Glen Allen, VA 23060-6711 Mr. Ken Paplham Mr. Alan J. Price E 4095 Sandy Bay Road Vice President - Nuclear Engineering Kewaunee, WI 54216 Innsbrook Technical Center - 2SE 5000 Dominion Boulevard Mr. Richard Gallagher Glen Allen, VA 23060-6711 Senior Scientist, License Renewal Dominion Resources Services, Inc.

Mr. William D. Corbin Route 156, Rope Ferry Road Director - Nuclear Engineering Waterford, CT 06385 Innsbrook Technical Center - 3NE 5000 Dominion Boulevard Glen Allen, VA 23060-6711

Kewaunee Power Station cc:

Mr. Ronald Kazmierczak Regional Director Wisconsin Department of Natural Resources Northeast Region Headquarters 2984 Shawano Avenue P.O. Box 10448 Green Bay, WI 54307-0448 Ms. Kathleen Angel Federal Consistency and Coastal Hazards Coordinator Wisconsin Coastal Management Program P.O. Box 8944 Madison, WI 53708-8944