ML13175A347

Request for Additional Information Regarding Request for Diesel Generator Completion Time Extension

ML13175A347
Person / Time
Site:	Brunswick
Issue date:	07/10/2013
From:	Gratton C Plant Licensing Branch II
To:	Hamrick G Carolina Power & Light Co
Gratton C, NRR/DORL/LPLII-2, 415-1055
References
TAC ME8893, TAC ME8894
Download: ML13175A347 (6)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 10, 2013 Mr. George T. Hamrick, Vice President Brunswick Steam Electric Plant Carolina Power & Light Company Post Office Box 10429 Southport, North Carolina 28461

SUBJECT:

BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION REGARDING REQUEST FOR DIESEL GENERATOR COMPLETION TIME EXTENSION (TAC NOS. ME8893 AND ME8894)

Dear Mr. Hamrick:

On June 19, 2012, Carolina Power & Light, doing business as Progress Energy Carolinas, Inc.

(Progress Energy), requested a license amendment that would revise the technical specifications for the Brunswick Steam Electric Plant, Units 1 and 2 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12173A112). Specifically, the amendment would extend the completion time for an inoperable diesel generator from 7 days to 14 days. On November 21, 2012, the Nuclear Regulatory Commission (NRC) staff provided a request for additional information (RAI) regarding the proposed revision (ADAMS Accession No. ML 1231 OA305). Progress Energy responded to the RAI by letter dated January 21, 2013 (ADAMS Accession No. ML13022A572).

The NRC staff continues to review 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 conference call with your staff on July 3, 2013, it was agreed that you would provide a response 60 days from the date of the call.

The NRC considers that timely responses to RAls help ensure sufficient time is available for the NRC staff review and contribute toward the NRC's goal of efficient and effective use of staff resources.

G. Hamrick - 2 If you have any questions regarding this letter, please feel free to contact me at (301) 415-1055.

Christopher Gratton, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-325 and 50-324

Enclosure:

Request for Additional Information cc w/encl: Distribution via ListServ

OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ADDITIONAL INFORMATION BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 CAROLINA POWER & LIGHT REQUEST FOR DIESEL GENERATOR COMPLETION TIME EXTENSION DOCKET NUMBERS 50-325 AND 50-324 TAC NOS. ME8893 AND ME8894 On June 19, 2012, Progress Energy Carolinas, Inc (Progress Energy) requested a license amendment that would revise the technical specifications (TSs) for the Brunswick Steam Electric Plant (BSEP) Units 1 and 2 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12173A112). Specifically, the amendment would extend the completion time for an inoperable diesel generator from 7 days to 14 days. On November 21, 2012, the Nuclear Regulatory Commission (NRC) staff provided a request for additional information (RAJ) regarding the proposed revision (ADAMS Accession No. ML12310A305). Progress Energy responded to the RAI via a letter dated January 21,2013 (ADAMS Accession No. ML13022A572). The NRC staff has reviewed risk information related to the proposed amendment and has identified the following areas where additional information is needed to complete its review:

1. Regulatory Guide (RG) 1.200, Rev. 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," states that results used to support a risk-informed application should be derived from a probabilistic risk assessment (PRA) model that represents the as-designed, as-built, as-operated plant to the extent needed to support the application. Clarify whether the PRA model of record accurately reflects plant changes associated with the BSEP extended power uprate, including but not limited to modifications to the standby liquid control system, updates to sequence timing, and human error probabilities.

2. A heat release rate (HRR) associated with motor fires (69 kilowatt (kW>> was used for pump electrical fires rather than the pump electrical HRR of 211 kW that is specified by Table G-1 in NUREG/CR-6850, "Fire PRA Methodology for Nuclear Power Facilities."

Given that a 211 kW fire would have a larger zone of influence (and thus, potentially new targets), provide a justification as to why the choice of HRR would not be expected to impact the change in risk associated with the proposed diesel generator (DG) completion time extension. Alternatively, provide a sensitivity study that shows the impact on change in core damage frequency (L'.CDF), change in large early release frequency (L'.LERF), incremental conditional core damage probability (ICCDP), and incremental conditional large early release probability (ICLERP) of using the NUREG/CR-6850-recommended HRR of 211 kW.

3. Section 2.3 of RG 1.177, Rev. 1, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," provides guidance on calculating ICCDP and ICLERP. Please clarify whether the method described by the RG was applied consistently to all hazard groups. For example, Enclosure 4, Section 4.6.2 of the License Amendment Request (LAR) quantifies the risk due to seismic hazards as follows:

Enclosure

- 2 ICCDP =flCDF x d Where:

=

flCDF [Seismic core damage frequency with completion time extension] - [Seismic core damage frequency without completion time extension]

d = Duration of exposure time = completion extension time = 14 days Bounding values are provided in Enclosure 4, Table A4-6 of the LAR:

Hazard Base CDF ICDF (y(') flCDF (y(') Duration ICCDP (y(1)

Seismic 3.4E-07 2.4E-06 2.1E-06 14 days = 8E-08

.038 years (Unit 1,

DG2, Operating Basis Earthquake)

This does not appear to be consistent with the approach used for internal events, internal flood, high winds, external flood, and internal fires as shown in Enclosure 4, Table A4-3 of the LAR:

Hazard ICDF (y( ) flCDF (yr' ) Duration Internal S.7E-07 (not Events rovided)

Calculating ICCDP in the same manner as for seismic would yield:

5.7 E-07 y(1 x .038 yrs =2.2E-8 This does not equal the ICCDP value shown in Table A4-3 of the LAR.

Hazard Base CDF ICDF (yr-l) flCDF (y(l) Duration ICCDP (y(1)

Fire 2.7E-OS (not 2.8E-07 (not 2.8E-07 provided) provided)

Calculating ICCDP in the same manner as for seismic would yield:

2.8 E-07 y(1 x .038 yrs = 1.1 E-8 This does not equal the ICCDP value shown in Table A4-16 ofthe LAR.

- 3 If ilCDF, ilLERF, ICCDP, and ICLERP were calculated in different ways for different hazard groups, please clarify whether this was done in accordance with the aforementioned RG 1.177 guidance. If not, provide a justification for why comparison with the acceptance guidelines in RG 1.177 and RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," would not be impacted for this application.

4. The Tier 2 guidance in RG 1.177, Section 2.3 states that the licensee should provide reasonable assurance that risk-significant plant equipment outage configurations will not occur when specific plant equipment is out of service consistent with the proposed change to the TSs.

Enclosure 4, Table A4-10 of the LAR provides risk achievement worth values for several such configurations but does not include an assessment as to whether "certain enhancements to the TS or procedures are needed to avoid risk-significant plant configurations," as called for by RG 1 177. Furthermore, the configurations that are identified were based only on the internal events model, even though the stated fire CDF values are higher than the internal event CDF values for both units.

The statement in Enclosure 1 of the LAR, Section 4.4.2.2 that "out-of-service combinations can be evaluated for their risk significance to determine if additional measures may be required," provides no definition of how risk-significant combinations are selected and does not appear to satisfy RG 1.177 guidance to provide reasonable assurance that risk-significant configurations will not occur. Please explain how your proposal complies with all of the Tier 2 guidance in Section 2.3 of RG 1.177. If the guidance is not clearly satisfied, provide justification for why the intent of the RG is met.

Provide a justification if any hazard groups are excluded from this analysis. If only one unit is analyzed, address how known asymmetries in fire PRA results were addressed.

G. Hamrick -2 If you have any questions regarding this letter, please feel free to contact me at (301) 415-1055.

Sincerely, IRA!

Christopher Gratton, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation Docket Nos. 50-325 and 50-324

Enclosure:

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