B17212, Forwards Responses to NRC RAI Re Proposed License Amend Request for Recirculation Spray Sys Direct Injection Change

Forwards Responses to NRC RAI Re Proposed License Amend Request for Recirculation Spray Sys Direct Injection Change

ML20247J076
Person / Time
Site:	Millstone
Issue date:	05/07/1998
From:	Brothers M NORTHEAST NUCLEAR ENERGY CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
B17212, NUDOCS 9805210333
Download: ML20247J076 (7)

Text

Northeast ""I" M "dd"""'" '56)' ***"'"'a. w a8s Nuclear Energy win, tone Noacar Pom station Northeast Nudear Energy Company P.O. Box 128

. Waterford, CT 06385 '1128 (860) 447-1791 Fax (860) 444-4277 The Northeast Utilities System MAY - 7 1998 Docket No. 50-423 B17212 Re: 10CFR50.90 10CFR50.59 (a)(2)

U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 3 Proposed License Amendment Request Recirculation Spray System Direct injection Change (PLAR 3-98-1)

Resoonse to Reauest for Additional Information Northeast Nuclear Energy Company (NNECO), in a letter dated March 3,1998, proposed an amendment to Chapter 6 of the Millstone Unit No. 3 Final Safety Analysis Report. The NRC in a letter dated May 7,1998, requested additional information to support their review of the submittal. Attachment 2 contains NNECO's responses to the NRC questions. 1 i

Attachment 1 identifies that no commitments are contained within this letter. If the NRC .

Staff should have any questions or comments regarding this submittal, please contact ]

Mr. D. Smith at (860) 437-5840. i Very truly yours, ,

NORTHEAST NU EAR ENERGY COMPANY N0 M. H. Brothers N Vice President - Operations cc: H. J. Miller, Region i Administrator  !

W. D. Travers, Ph.D., Director, Special Projects Office l l J. W. Andersen, NRC Project Manager, Millstone Unit No. 3 A. C. Cerne, Senior Resident inspector, Millstone Unit No. 3

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Director )

Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street g)

Hartford, CT 06106-5127 9905210333 990507 I PDR ADOCK 05000423 p PDft

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!' Docket No. 50- 423 l

B17212 i  !

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l Attachment 1 Millstone Nuclear Power Station, Unit No. 3 Proposed License Amendment Request l Recirculation Spray System Direct injection Change (PLAR 3-98-1)

Response to Request for Additional Information NNECO's Commitments i

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May 1998

U.S. Nuclear Regul: tory Commission B17212 Attachment 1\Page 1 Enclosure List of Regulatory Commitments The following table identifies actions committed to by NNECO in this document. Please notify the Manager - Regulatory Compliance at the Millstone Nuclear Power Station Unit No. 3 of any questions regarding this document or any associated regulatory commitments.

Commitment Committed Date or Outage NONE I

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Docket No. 50-423 B17212 l

l Attachment 2 Millstone Nuclear Power Station, Unit No. 3 Proposed License Amendment Request Recirculation Spray System Direct injection Change (PLAR 3-98-1)

Response to Request for Additional Information 1

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U.S. Nuclear Regulitory Commission B17212 Attachment 2\Page 1 Question T -

In your reanalysis, the assumed pump degradation was reduced from 10 percent to 5 percent. Discuss the procedures you have in place for the monitoring of pump l degradation and how these procedures support this change.  !

NNECO's Response to Question 1 The revised degraded pump curves (flow versus pressure differential) were utilized during the performance of Surveillance Procedure EN - 31121. This surveillance procedure was utilized to verify that the pump did not degrade beyond the flow assumed in the accident analysis. _;

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Question 2 in your submittal, the proposed change will still require direct cold leg injection in case of a long-term passive failure. In the 1986 Final Safety Analysis Report change, you completely deleted references to the Recirculation Spray System (RSS) direct injection path. Please include a discussion of this configuration for completeness or justify its exclusion.

NNE.CO's Resoonse to Question 2 NNECO indicated in the February 16,1998 submittal, that the 1986. modification did not /

i eliminate direct cold leg injection from the Emergency Operating Procedures (EOP). It was recognized at that time, that cold leg direct injection would be needed in case both Safety injection (SlH) and the Charging (CHS) pumps failed during long-term recovey actions after a Loss of Coolant Accident (LOCA), in the beyond design basis situation.

Review of the 1986 Final Safety Analysis Report (FSAR) change made to support the modification indicates that it was thought that sufficient redundancy existed in the Emergency Core Cooling System (ECCS) components and associated flow paths to accommodate a limiting long tena passive failure without reliance on the direct injection flow path. The safety evaluations, written to support the 1986 change, did not specifically address the limited passive failure scenario.

A subsequent 1998 evaluation of the 1986 modification concluded that the EOP guidance would have supported the core cooling function for the mitigation of a limited passive failure, i

FSAR Section 6.3.2.5, ECCS System Reliability, addresses limited passive failure in some detail. The structural failure of a static component that limits the component's '

effectiveness in carrying out its long term design function is considered a passive failure. Examples include cracks in pipes, sprung flanges, valve packing leaks or pump i

U.S. Nucl:tr Regulitory Commission B17212 Attachment 2\Page 2 seat failures. A single failure analysis is presented in Table 6.3 6. FSAR Table 6.3-6 has been enhanced by a 1997 FSAR change (MP3-97-0569) to addresses the Limiting Passive Failure and the use of the direct injection flow path. Accordingly, use of the direct injection flow path for a Limited Passive Failure is included in the current FSAR.

Question 3 On page 6.2-50s, you deleted the discussion on evaluation of Net Positive Suction Head (NPSH) for the recirculation mode of the RSS. Confirm that the recirculation spray mode is the limiting case for required NPSH.

NNECO's Response to Question 3 An evaluation of the RSS pump NPSHm., was performed for the RSS spray, cold leg recirculation and hot leg recirculation modes. It was confirmed that the limiting case was the spray mode.

It should be noted that during the preparation of the response to this question, review of PLAR 3-98-001 (i.e. the existing 1986 FSAR change request) determined that the NPSH data provided was inaccurate. In the 1986 time frame, additional analyses were completed to evaluate the impact of LOCA induced containment sump screen blockage on RSS pump NPSH. However, the results of these analyses were not incorporated into the FSAR prior to the elimination of the direct injection mode of ECCS cold leg recirculation. The correct values for the Spray Phase of operation are documented in a February 26,1986 calculation revision and are provided below:

1st Stage imp El(ft) -47.33 Floor Water Level (ft) -23.1 Z (ft), El Head 24.23 H (ft), Pipe Loss 10.96 Z-H (ft), NPSHavail 13.27 Q (gpm) 4475 NPSHrea ja Margin (ft) 2.27 1

Question 4 in your submittal you stated that the results of the design basis analyses were verified to be acceptable without the direct injection in that the modified alignment delivered sufficient flow to meet the long-term cooling requirement after Loss of Coolant Accident, and the results of the containment analysis show that the design basis of maintaining l

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U.S. Nuclear Reguiltory Commission B17212 Attachment 2\Page 3 subatmos@eric containment pressure was unchanged. Discuss in more detail the evaluations / analyses performed to support these conclusions.

NNECO's Response to Question 4 Lona Term Coolina Requirements The minimum ECCS flow requirement for long-term core cooling following a LOCA is the alignment in the recirculation mode without direct injection on minimum Engineered

' Safety Features (ESF). Minimum ESF is defined as one charging pump and one ECCS pump. Enough ECCS flow must be provided to exceed RCS boil-off in accordance with 10CFR50 Appendix K requirements and with Westinghouse internal criteria. To

onsider whether these requirements / criteria were met, three basic cases were considered:

1) cold leg recirculation mode following a cold leg break,

2) cold leg recirculation mode following a hot leg break, and

3) hot leg recirculation mode following a hot leg break .

The most limiting break was determined to be the cold leg break while the plant is in the cold leg recirculation mode. For this case, at the time when recirculation would be initiated, Reactor Coolant System (RCS) boil-off was calculated to be 68.9 lbm / see while total recirculation flow was 148.8 lbm / sec. Of this total recirculation flow,111.6 lbm / sec is calculated to enter the core while 37.2 lbm / see was calculated to be lost out of the break. Since the total recirculation flow to the core is greater than the RCS boil off with margin, the proposed ECCS alignment is acceptable.

The RCS boil-off is calculated based on ANS decay heat in accordance with 10CFR 50 Appendix K at 20 minutes. The calculation of boil-off is conservative since the minimum time to recirculation is greater than 30 minutes and the boil-off requirements are reduced as time elapses.

The long term cooling analysis assumes no direct mixing of the recirculation flow with the core until after the switchover time to hot leg recirculation. This assumption allows the maximum amount of boron to build-up in the core. The proposed alignment has no effect on the time when hot leg switchover should occur and has no effect on the amount of boron calculated to build-up in the core.

Based on the above analysis work, all requirements for long term cooling were

! determined to have been met.

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U.S. Nucicer R:gul: tory Commission B17212 Attachment 2\Page 4 Containment Analysis A calculation was prepared to address containment integrity relative to changing the direct injection flow path during the recirculation phase from an automatic action to a contingency action. The calculation evaluated the effect of the reduced RSS flow on containment response. Prior to the reduction in RSS recirculation phase flow, the total flow to the core for the minimum ESF case was approximately 4000 gpm. The RSS recirculation phase flow to the core was reduced to approximately 1200 gpm by closing

/ maintaining closed the direct injection isolation valves. The flow provided by the remaining RSS pump (assuming minimum ESF) to the spray header was unchanged.

The pump suction (PS) double ended rupture (DER) at 100% power was selected as the limiting case for evaluating the effect of reduced RSS flow through the RSS heat exchangers. The limiting case for peak containment pressure is a hot leg (HL) DER.

The peak pressure occurs relatively early in the accident (HLDER), before Quench Spray System (OSS) or RSS become effective. Main Steam Line Break (MSLB) does not govern for containment pressure. Therefore, based on existing plant specific analysis results the PSDER at 100% was the case which could be most adversely affected by the reduction in containment cooling in the long term (i.e. the limiting case for demonstrating that the unit will return to subatmospheric conditions in one hour is the PSDER).

The LOCTIC Computer Code Version 23, Level 01 was utilized to perform the reanalysis of the containment response for the reduction of RSS flow. The overall RSS heat exchanger heat transfer coefficient was adjusted accordingly for the reduction in RSS flow. The LOCTIC Code was used to evaluate the minimum ESF, PSDER at 100% power with the following initial conditions:

Initial Containment Pressure: 9.8 psia Initial Containment Temperature: 80 *F Service Water Temperature: 75 *F The analysis results with 10% degraded pumps resulted in a return to atmospheric conditions (i.e. 0.02 psig) at 16,900 seconds.

Since unacceptable results were obtained with the 10% degraded pumps, the pumps were only degraded by 5%. The same initial conditions specified above were utilized with the 5% degraded pump flow. The results for this analysis were acceptable. The subatmospheric peak pressure was -0.09 psig for this case.

Based on the above analysis work, all requirements for containment heat removal including return to subatmospheric conditions within hour were determined to have been met.

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