ML080560615

Response to Request for Additional Information Regarding Stretch Power Uprate License Amendment Request Revised and Supplemental Responses to Questions AFPB-07-0007 & AFPB-07-0008

ML080560615
Person / Time
Site:	Millstone
Issue date:	02/25/2008
From:	Gerald Bichof Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
07-0799A
Download: ML080560615 (7)

Text

Dominion Nuclear Connecticut, Inc.

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February 25. 2008 U. S. Nuclear Regulatory Commission Serial No.: 07-0799A Attention: Document Control Desk NLOS/MAE: R1 One White Flint North Docket No.: 50-423 11555 Rockville Pike License No.: NPF-49 Rockville, MD 20852-2378 DOMINION NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING STRETCH POWER UPRATE LICENSE AMENDMENT REQUEST REVISED AND SUPPLEMENTAL RESPONSES TO QUESTIONS AFPB-07-0007 AND AFPB-07-0008 Dominion Nuclear Connecticut, Inc. (DNC) submitted a stretch power uprate license amendment request (LAR) for Millstone Power Station Unit 3 (MPS3) in letters dated July 13, 2007 (Serial Nos. 07-0450 and 07-0450A), and supplemented the submittal by letters dated September 12, 2007 (Serial No. 07-0450B) and December 13, 2007 (Serial No. 07-0450C).

The NRC staff forwarded requests for additional information (RAls) in October 29, 2007, November 26, 2007, December 14, 2007, and December 20, 2007 letters. DNC responded to the RAls in letters dated November 19, 2007 (Serial No. 07-0751), December 17, 2007 (Serial No. 07-0799), January 10, 2008 (Serial Nos. 07-0834, 07-0834A, 07-0834C, and 07-0834F), January 11, 2008 (Serial Nos. 07-0834B, 07-0834E, 07-0834G, and 07-0834H),

January 14, 2008 (Serial No. 07-0834D), January 18, 2008 (Serial Nos. 07-0846, 07-0846A, 07-0846B, 07-0846C, and 07-0846D) and January 31, 2008 (Serial No.

07-08341).

In a conference call with the NRC staff on January 24, 2008, the NRC staff requested additional clarification to the response to RAI Question AFPB-07-0007 provided in DNC's December 17, 2007 letter. The response to RAI Question AFPB-07 -0007 has been revised and is provided in the attachment to this letter. This revised response supersedes the December 17, 2007 response to RAI Question 07-AFPB-0007 in its entirety.

ONe's December 17, 2007 letter also contained a response to RAI Question 07-AFPB-0008.

That response committed to providing the results of the updated control room fire analysis by February 29,2008. A supplemental response to RAI Question 07-AFPB-0008 which contains the updated control room fire analysis is also provided in the attachment to this letter.

The information provided by this letter does not affect the conclusions of the significant hazards consideration discussion in the December 13, 2007 DNC letter (Serial No.

07-0450C).

Serial No. 07-0799A Docket No. 50-423 Supp. Resp, SPU Page 2 of 3 Should you have any questions in regard to this submittal, please contact Ms. Margaret Earle at 804-273-2768.

Sincerely, Vice President - Nuclear Engineering COMMONWEALTH OF VIRGINIA COUNTY OF HENRICO The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Gerald T. Bischof, who is Vice President - Nuclear Engineering of Dominion Nuclear Connecticut, Inc.

He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this J 5 ~day or-JldJ/JLril.j ,2006 My Commission Expires: J~ 3~ doLo . ff~ ./ /

!Lk1:L ~ . . :MtlL Notary Public VICKI L. HULL Commitments made in this letter: None Attachment COIMIIlllon NoIaIY PuDIC COINftOn* . . - of . . . . .

al.lO,e cc: U.S. Nuclear Regulatory Commission Region I Regional Administrator 475 Allendale Road King of Prussia, PA 19406-1415 Mr. J. G. Lamb U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 0-8B1A Rockville, MD 20852-2738

Serial No. 07-0799A Docket No. 50-423 Supp. Resp, SPU Page 3 of 3 Ms. C. J. Sanders Project Manager U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Mail Stop 0-8B3 Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127

Serial No. 07-0799A Docket No. 50-423 ATTACHMENT LICENSE AMENDMENT REQUEST STRETCH POWER UPRATE LICENSE AMENDMENT REQUEST RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REVISED AND SUPPLEMENTAL RESPONSES TO QUESTIONS AFPB-07-0007 AND AFPB-07-0008 MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC.

Serial No. 07-0799A Docket No. 50-423 Attachment, Page 1 of 3 NRC Question AFPB-07-0007 LAR Attachment 5, Section 2.5.1.4.2.3.4, "Safe Shutdown Evaluations," states that

"... the safe-shutdown analysis identifies fire-induced failures that affect the plant and the operator actions that can be used to compensate for these failures ..."

Discuss the response time, including any assumptions, especially those of a potentially non-conservative nature, which may have been made in determining that the operator manual actions can confidently be accomplished within the available time.

Revised ONC Response As described in LAR Section 2.5.1.4.2.3.4, the revised analyses performed at SPU conditions for both Control Room Fire Transient and Charging Cubicle Fire Transient confirms that SPU does not impact the required operator action times.

The assumptions for critical operator actions used in pre-SPU and post-SPU analyses remain the same for the Control Room Fire Transient. The critical operator actions assumed following the reactor trip from the control room and initiation of main steam isolation (MSI) signal from the control room are letdown isolation and charging flow restoration. Specifically, the analysis assumes 15 minutes for letdown isolation and 30 minutes for the restoration of charging flow from the event initiation. The response times for these assumed action times have been validated for current power levels and are well within the assumed operator action times. These response times are not impacted for post SPU conditions.

Note that the response time for the manual initiation of auxiliary feedwater is discussed in AFPB-07-0008. Operator initiation time of auxiliary feedwater flow is not a parameter used in the above stated analysis.

The current analysis for the Charging Cubicle Fire Transient was performed using the TREAT model. There are several operator actions that occur at specific times throughout the transient that are critical in maintaining pressurizer level on scale.

Specifically, the current analysis assumes letdown isolation in 5 minutes, reactor trip in 10 minutes, closure of the MSI valves (MSIVs) in 11 minutes and securing the pressurizer heaters in 15 minutes. The revised analysis for SPU conditions was performed using the NOTRUMP model. The assumed critical operator action times remain the same as the pre-SPU conditions. It is noted that the SPU analysis assumed 695 seconds (11.6 minutes) for MSIV closure. This provides a small increase in margin for operator action that is unrelated to SPU. These analysis assumptions are bounding with respect to the fire shutdown procedure of record.

Serial No. 07-0799A Docket No. 50-423 Attachment, Page 2 of 3 NRC Question AFPB-07-0008 LAR Attachment 5, Section 2.5.1.4.2.3.7, "Operator Actions Required Following a Fire" states that" ... [an] analysis was performed to determine the steam generator dryout time at the support stretch power uprate (SPU) power level; the results showed a dryout time of approximately 37 minutes. Therefore, there continues to be adequate time for the operator to manually initiate auxiliary feedwater to the steam generators (SGs) at SPU conditions ..."

Discuss the response time, including any assumptions that may have been made in determining that the operator manual actions can confidently be accomplished before SG dryout.

ONC Initial Response The following response was provided in DNC letter dated December 17, 2007, (Serial No. 07-0799):

"The MP3 BTP 9.5-1 Compliance Report contains information on prioritization of operator actions. One of the manual actions given high priority is auxiliary feedwater (AFW) initiation to a minimum of two Steam Generators (SGs) for the fire shutdown scenario requiring control room evacuation (i.e., CB-8, 9, 11A1B fires).

In Attachment 5, Section 2.5.1.4.2.3.7 of the LAR, the steam generator dry-out time is stated as approximately 37 minutes. During validation and verification of the response to this RAI, it was identified that there was an error in the SPU license submittal. The correct steam generator dry-out time for SPU conditions is 34.67 minutes based on the cited analysis. This value is obtained from an analysis that used a very conservative reactor trip time.

MP3 BTP 9.5-1 Compliance Report (Section 6.1.1) states that AFW flow can be initiated within 1O-minutes to a minimum of two steam generators. It was recently identified that the initiation of AFW flow would occur at approximately 27 minutes for the limiting fire scenario.

Even with a decrease in steam generator dry-out time and an increase in the AFW flow initiation time, based on the availability of the large capacity turbine driven AFW pump, preliminary analysis has shown that the margin between 27 minutes and 34.67 minutes is still sufficient to assure that all BTP 9.5-1 criteria will be met. In order to provide a better estimate of the steam generator dry-out time as well as provide more complete documentation that all BTP 9.5-1 criteria can be met, an update to the control room fire analysis is in progress. Results of the updated control room fire analysis will be provided to you by February 29, 2008."

Serial No. 07-0799A Docket No. 50-423 Attachment, Page 3 of 3 ONe Supplemental Response:

As committed to in the DNC response to RAI AFPB-07-0008 (DNC letter 07-0799 dated December 17, 2007), a thermal hydraulic analysis has been performed for the fire shutdown scenario requiring control room evacuation (i.e., CB-8, CB-9, CB-11A1B fires),

which assumes a delayed auxiliary feedwater (AFW) initiation time of 31.5 minutes (30 minutes plus 90 second turbine driven AFW pump start time) to a minimum of two steam generators (SGs). As stated in DNC initial response, the operator action to initiate AFW flow from the auxiliary shutdown panel to at least two SGs has been estimated to occur in approximately 27 minutes for the limiting fire scenario based upon our current fire shutdown procedures. In order to provide margin, this AFW initiation time will be reduced to less than 21.5 minutes (20 minutes plus 90 second turbine driven AFW pump start time) prior to implementation of the SPU, by making changes to the fire shutdown procedure and design compliance report.

The analysis shows that the level in the two steam generators begins to recover immediately after receiving AFW flow at 31.5 minutes. Secondary steaming shifts preferentially to the two steam generators not receiving AFW flow. Inventory in the two steam generators not receiving AFW flow continues to decrease until dryout in the range of 45 minutes.

Even though the analysis conservatively assumes no AFW flow to two of the SGs because of the closure of two of the AFW control valves due to potential hot shorts, the current fire shutdown design/procedure calls for denergizing the AFW solenoid operated control valves causing them to fail open. This action will occur before the 45 minute dryout time. Thus, completing this action within 45 minutes will assure none of the steam generators will dryout. The minimum AFW flow rate and operator action time assumed in the analysis are sufficient to remove decay heat and prevent SG dryout.