ML20216J931

Forwards Response to NRC 981119 Suppl RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions

ML20216J931
Person / Time
Site:	Nine Mile Point
Issue date:	09/30/1999
From:	Abbott R NIAGARA MOHAWK POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-96-06, NMP1L-1473, NUDOCS 9910070041
Download: ML20216J931 (10)

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September 30,1999 Pune siunau i Mce Pasident NMPIL 1473 rex:sina44n NuclearEngineering  ;

, U. S. Nuclear ' Regulatory Commission Attn: Document Control Desk Washington, DC 20555 RC: Nine Mile Point Unit 1 Docket No. 50-220 DPR-63

Subject:

Supplemental Requestfor Additional Irtfonnation Regarding Generic Letter 96-06 ;

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. Gentlemen:

The NRC issued Generic Letter (GL) 96-06, " Assurance of Equipment Operability and Containment Integrity During Design-Basis Accident Conditions," on September 30,1996.

GL 96-06 identified certain safety-significant issues, including thermally induced overpressurization of isolated water-filled piping sections inside the containment, that could affect containment integrity and equipment ' operability under accident conditions. The GL requested licensees to make appropriate determinations and report the results, along with any corrective actions implemented or planned to be implemented.

In its February 7,1997 letter, Niagaa Mohawk Power Corporation (NMPC) identified 10 penetrations at Nine Mile Point Unit .1 (NMP1) with associated piping segments susceptible to thermally induced overpressurization. Later that year, during refueling outage number 14

' (RF014), NMPC modified three of the 10 penetrations to provide protection against piping overpressurization. By letter dated November 19,1998, the NRC sent a supplemental request for additional information regarding the seven penetrations that remained to be add.?ssed, and requested that this additional information be provided by September 30,1999. The Enclosure and Attachment to this letter provide the requested information.

Very truly yours,

- L-9910070041 990930 PDR ADOCK 05000220 .M[N

• P PDR Richard B. Abbott gsin08 vice President nuclear Engineering RBA/IAA/ kap q4 Enclosure and Attachment ,oI xc: Mr. H. J. Miller, NRC Regional Administrator Mr. S. S. Bajwa, Section Chief PD-I, Section 1, NRR Mr. G. K. Hunegs, NRC Senior Resident Inspector Mr. D. S. Hood, Senior Project Manager, NRR Records Management Nine Mde Point Nuclear Station PD. Box 63. Lyconung, New York 13093-0063

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ENCLOSURE

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ENCLOSURE SUPPIIMENTAL REOUEST FOR ADDITIONAL INFORMATION l REGARDING RFRPONSE TO GL 96-06 NIAGARA MOIIAWK POWER CORPORATION NINE MIII POINT NUCIRAR STATION UNIT NM DOCKET NO. 50-220 ReauestfarInformarian Mh Provide summaries ofNMPC's planned corrective actionsfor the 31 pipe segments determined i to be susceptible to thennally-induced pressurization. i l

Resnonce #1:

The response below addresses the ten penetrr.uons at Nine Mile Point Unit 1 (NMP1) with associated piping segments that were previcusly iden ified as potentially susceptible to i l

thermally-induced overpressurization during an accid mt condition. The response for the 21 Nine Mile Point Unit 2 piping segments was previsus.'y provided by Niagara Mohawk Power Corporation (NMPC) letter dated January 20,1999.

Three of the ten NMP1 penetrations (X-7, X-8, and X-139) were modified during refueling 1 I

outage number 14 (RFO14) to resclve the pressurization issue. Responses to the NRC's initial Request for AdditionalInformation concerning these three penetrations were submitted by the January 20,1999 letter. The remaining seven penetrations are:

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• Penetration X-9 (Reactor Water Cleanupi - Piping between internal isolation valve 33-02R and external isolation valve 33-04. _ Se 6 inch diameter piping includes both

!- carbon steel and stainless steel segments.

• Penetration X-13A (Core Spray) - Piping between external isolation valve 40-02 and internal isolation valves 40-01 and 40-09. The piping is 12 inch diameter stainless steel.

• Penetration X-14 (Core Spray) - Piping between external isolation valve 40-12 and internal isolation valves 40-10 and 40-11. The piping is 12 inch diameter stainless steel.

• Penetration X-25 (Drywell Floor Drain) - Piping between internal isolation valve 83.1-11 and external isolation valve 83.1-12. The piping is 4 inch diameter carbon steel.  !

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I 'e Penetration X-26 (Drywell Equipment Drain)'- Piping between internal isolation vsive

-83.1-09 and external isolation valve 83.1-10. The piping is 3 inch diameter carbon  ;

steel.

e Penetration X-154 (Reactor Water' Cleanup) - Piping between internal isolation valve 33-01R and external isolation check valve 33413. The 6 inch diameter piping includes  ;

both carbon steel and ' stainless steel segments. I e Penetration X 238 (Core Spray High Point Vent) - Two piping segments: the first is

. between external isolation valve 40-32 and internal isolation valve 40-30. The second :i 5

piping segment is between external isolatioa valve 40-33 and internal isolation yhlve

' 40-31. The piping is 1 inch diameter stainless steel.

Detailed analyses have been performed on these seven piping segments to evaluate the capability of the piping to withstand pressurization induced by containment heatup under design basis accident conditions. These analyses were performed in accordance with the

~ ASME Code,Section III, Appendix F (1986 edition) criteria for inelastic analysis, with results compared to Level D service limits to determine acceptability. In all'seven cases, the piping l segment in~ question was found to be within Level D service limits under design basis accident conditions.

' The core spray isolation valves associated with penetrations X-13A and X-14 have post-l accident operability requirements to open but may not be able to open against the maximum internal pressures calculated if one assumes no leakage past the valves. However, these isolation valves currently remain operable due to known leakage past them. . As previously ,

l stated in NMPC letter dated December 16,1997, penetrations X-13A and X-14 will be )

modified during refueling outage number 16 (RFOl6) to preclude thermal pressure buildup under Generic letter 96-06 postulated conditions. The proposed modification is similar to the modification implemented for the shutdown cooling system penetrations X-7 and X 8 during RFO14. This modification will add a bypass line around one of the inboard isolation valves for each penetration. The bypass line will contain a check valve, which will prevent reactor coolant leakage through the bypass line during normal operating conditions. These check valves will also serve as reactor coolant isolation valves.

Note that use of ASME Code,Section III, Appendix F criteria is currently not addressed in the  ;

NMP1 Updated Final Safety Analysis Report (UFSAR). The original design code for the l piping segments was ASA B31.1-1955, including Nuclear Interpretations. A 10CFR50.59 safety evaluation for using Appendix F criteria is in preparation. If this evaluation indicates that an unreviewed safety question is involved in the use of Appendix F as a permanent resolution method, a license amendment request will be prepared and submitted.

Reauestforinfon=ntion #2:

For any ofNMPC's modifications that involve heat transfer and/or structural analyses of the pipe segments, provide thefollowing informationfor these segments:

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l b t f b2.1& Specify the applicable design criteriafor the piping and tulves. Thh should include the

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, requiredloa, combinedonsq 2.2 Submit a drawing (or identify a previously docketed drawing) of the piping run betsun p the isolation valves. This should include the lengths and thicknesses ofthe piping segments and the type and thickness of the insu!ation;

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[ 2.3  : Provids the maximum-ca!culated temperature and pressurefor the ylping run.

[ Describe, in detail, the method used to calculate these pressare and temperature I values. This should include a discussion of the heat transfer model and the basisfor l+, ,  : the heat transfer coeficients used in the analysis.

i 7 ' bename #22: -

The following response is specific to the two core spray penetrations X-13A and X-14, as .

these are the only remaining penetrations that need to be modified to resolve GL 96-06  !

( overpressurization concerns.  ;

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The original design code for the core spray piping segments was ASA B31.1-1955, in:luding

Nuclear Interpretations and certain requirements of the ASME Code, Section III-B-1965, L Articles N324 and N460 to N469. The applicable design conditions for the two piping L segmints are listed below.

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PENETRATION DESIGN PRESSURE DESIGN TEMPERATURE l

L X-INA ' 1200 psig 575*F i

%-14 1200 psig 575"F The load combinations used in the design basis piping analysis are as follows:

l I. Nonal and Seismic I

II. Normal, Seismic, and Accident (excluding waterhammer)

III. N'ormal and Waterhammer (excluding external stresses, as allowed by ASA B31.1-1955)'

Rmenance #2.2: .

Sketches showing the piping run between the isolation valves for the two core spray

penetrations (X-13A and X-14) are included in the Attachment.

Res=aa** #2 3:

The calculated maximum sustained temperature for the core spray piping segments is approximately 261*F under small break loss of coolant accident (SBLOCA) conditions. The 3

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. j. small break accident scenari results in higher drywell temperatures for a sustained time f*'1'y @riod.3 The value of 261,*F is the highest saturation temperature that could be reached under

'SBLOCA conditions befox containment sprays would be initiated per Emergency Operating

- Procedur> (EOP) required actions. This temperature is based on suppression chamber spray (initiadon pressure (13 psig), a pressme allowance for instrument uncertainty (5.15 psig), and

.tSe hydrostatic pressure due.to maximum downcomer submergence (2.81 psig). However,

, ' calculations to determine the adequacy of the bypass lines with check' valves installed to t

, provide pressure relief for the coke spray penetrations (X-13A and X-14) will be I;. conservatively based on the design basis accident containment temperature profile (see UFSAR Figure XV-60 in the Attachment). The maximum pressure in the penetration will be kept at or below 1320 psig _(1.1' times design pressure). The pipe outer surface convective heat transfer coefficitnt will be based on Uchida approximation. The inside heat transfer coefficient will be conservatively calculated on the basis of free convection over a vertical wall. No credit will ,

be'taken for leakage through the inboard or outboard isolation valves.

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