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Responds to 770831 Ltr Ref GE Rept 170 F Pool Temp Limit for Safety Relief Valve Ramshead Condensation Stability, in Response to Questions 1,2,4 & 5.Submittal Date for Response to Question 3 Will Be Provided
	ML20023B720
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	09/26/1977
From:	Burke T; LONG ISLAND LIGHTING CO.
To:	Kniel K; Office of Nuclear Reactor Regulation
Shared Package
ML20023B716	List: ML20023B717; ML20023B720;
References
	FOIA-82-455, FOIA-82-471, FOIA-82-479 NUDOCS 8305060164
	Download: ML20023B720 (1)
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LONG ISLAND LIGHTING COM PANY k[ hh SHOREHAM NUCLEAR POWER STATION P.O. BOX 618, NORTH COUNTRY RO AD . WADING RIVER, N.Y.11792

September 26, 1977 - SNRC-219

    .      Mr. Karl Kniel, Chief Light Water Reactors Branch 2 Division of Project Management U. S. Nuclear Regulatory Commission Washington, D.C.          20555 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322

Dear Mr. Kniel:

In response to your letter of August 31, 1977, to Mr. A. W. Wofford, we are referencing General Electric's memorandum report "170 F Pool Temperature Limit for SRV Ramshead Condensation Stability" dated September 1, 1977 and transmitted by General Electric letter ~ MFN343-77 to your Mr. O. D. Parr, Chief, Light Water Reactors Branch 3, for our responses to Questions 1, 2, 4, and 5. With regard to Question 3, which is plant specific and will be an-swered by LILCO, we understand that the assumptions outlined in Question 3 as received by us, may undergo revision as a result of discussions between the NRC, Georgia Power Company, and General Electric on the Edwin'I. Hatch Nuclear Plant (Docket No. 50-366). We cannot provide a firm date for the completion of the required analyses until the required assumptions are finally agreed upon. However, we estimate that the types of analyses required will take four to six months for completion.

We will provide the submittal date for our response to question 3 within 7 days after the revised assumptions for this question are received by us. , Very-truly yours,

J. Turke roject Manager Shoreham Nuclear Power Station WJM/dr

8305060164 PDR FOIA 830317 HAFNER82-455 PDR h~ N

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                   ,       _,                      P.O. BOX 618. NORTH COUNTRY ROAD + WADING RIVER. N.Y.11792 December 16, 1981                                                              SNRC-645 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.                  20555 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322

Dear Mr. Denton:

Enclosed are fifteen (15) copies of Revision 5 to the Shoreham Plant Design Assessment Report (DAR) .for SRV and LOCA Loads. This report is submitted in accordance with our Mark II Containment Closure Program as outlined in our meeting of March 5, 1981. The DAR Revision 5 contains the completed design . assessment for the Shoreham Design Basis Loads, (NUREG-0487 & { Supp. 1). It also contains the Shoreham confirmatory program results based on the final generic load definitions and accept-ance criteria set forth in NUREG-0802 (DRAFT) and NUREG-0808. A proprietary supplement to Revision 5 to the Design Assessment Report is being sent under separate cover. Very truly yours, J. L. Smith l Manager, Special Projects l Shoreham Nuclear Power Station l l CC:mp ' Enclosure cc: J. Higgins i f ff .' < ',50 /pd V pj r FC.e935_

f APPENDIX J SNPS SUPPRESSION POOL TEMPERATURE TRANSIENTS r Appendix J' presents the results of analysis conducted by General Electric Company of SNPS suppression pool temperature response to plant transients involving SRV discharge. These transients have been analyzed in accordance with Ref erence 1 with one exception: loss of off site power is not assumed for SBA and isolation / SCRAM cases as described in IV.A.1. .This assumption is conservative as discussed below. Table J-1 presents the input data f or SNPS used in the analysis. Charts 1 through 6 and the accampanying Figs. J-1 through 6 present a brief description of each case and the corresponding reactor vessel pressure / suppression pool temperature transient. The following additional inf ormation is presented to facilitate review of the Shoreham-unique aspects of tnis analysis:

1. Feedwater is added non-mechanistically for all transients. In all cases except SORV at power with main condenser available, steam line isolation is assumed to occur 3.5 seconds after the start of the transient to maximize heat addition to the pool. In reality, MSIV closure would rapidly terminate steam flow to the turbine driven feed pumps and feedwater flow to the vessel. In such a sequence of events, teedwater could again begin to enter the vessel once reactor pressure f ell below the shutof f head of the condensate pumps.

This assumption is the basis for continuing to add feedwater. However, the condensate pumps require availability of of f site power to continue operation, and

therefore, availability of offsite power to continue operation, and therefore, availability of offsite power l is the conservative assumption.

l -

2. Availability of offsite power permits continued delivery of CRD return flow to the vessel. CRD 11ow was not assumed in the analysis of SBA with failure of one RHR Ex for Shoreham and provides additional conservatism for this case.

3. Availability or drywell coolers is not assumed f or SBA cases. For SBA cases, pool cooling is suspended for 10

                 . minutes    when     reactor       pressure      decreases to the permissive value for LPCI operation.               The 10 minutes suspension allows- for realignment of the RBR system to pool cooling mode.

4. For Shoreham, HPCI operation is not included in the transient analysis because of the FW assumption discussed above. In reality, HPCI would be expected to ope rate . There is no pool temperature cutoff for HPCI L

l J-1 Revision 5 - December 1981 i

2 .- .. e operation incorporated in the Shoreham design. Adequate HPCI pump NPSH has been verified f or operation at the maximum pool temperature observed in the analyses for reactor pressure greater than 150 psia.

5. For Shoreham, no single active failure can result in the permanent loss of one loop of pool cooling and the simultaneous- loss of shutdown cooling mode. A power supply failure disabling the inboard letdown valve for pool cooling and one RHR loop can be overcome by manual realignment of four valves in the "f aulted loop" (see Fig. J-7). Manual alignment of the faulted RHR HX loop in pool cooling mode is accomplished by opening two valves in the test line, opening the service water valve i for the RHR HX and closing- the HX bypass. The requirement for manual operation of tne affected RHR loop in pool cooling mode following a division electrical failure in the course of an uncontrolled heat addition to the pool (e.g., SBA) will be included in the emergency procedures for the plant. The required valves a are accessible for manual operation and realignment can be accomplished within 1 hour.

A 1 hour delay in bringing the secono loop into , 2 operation would have a negligible effect on the final pool temperature. The seal cooler on the " swing" bus pump would not be operable in this configuration, but seai cooling is not required for pump flow temperatures less than 2120F which is satisf actory for pool cooling mode. Passive failures (e.g. of the heat exchanger) are not postulated in the short term, and two heat excnanger operation is not required in the long term.

6. Section 10.4 describes pool temperature alarms at TS1 t

(90 0F) requiring operation of the RHR pool cooling mode and at TS3 (1100F) requiring SCRAM. Technical .- speci11 cation requires the node switch to be placed in -

                          " Shutdown" if TS3 is exceeded.             Shoreham has also implemented pressure switches in the SRV tailpipes to provide positive indication or an SPV lift.

7. The Shoreham main condenser is described in detail in FbAR Section 10.4.1. Operating procedures will be reviewed to ensure that the main condenser is identified I as the preferred heat sink for any transient not' leading to MSIV cicsure.

8. The Shoreham main conde"nser air removal system, steam seal system, turbine bypass system and circulating water system are described in detall in FSAR Sections 10.4.2 through 10.4.5 respectively.

l i u J-2 Revision 5 - December 1981 [ l l l

REFERENCES APPENDIX J

1. Assumptions for Use in Analyzing Mark II BWR Suppression Pool Temperature Response to blant Transients Involving Safety / Relief Valve Discharge, Mass / Energy Subconsulttee, Mark II Owners Group, March 24, 1980.

I

t ( J-3 Revision 5 - December 1981

TABLE J-1 SBOREHAM IMPORTANT SYSTEM CHARACTERISTICS - POOL TEMPERATURE TRANSIENT ANALYSIS Initial Pool Mass 4.53 x 106 lha Initial Pool Temperature 900F Initial RPV Liquid Mass 465,548 lixa Initial RPV Steam Mass 17,069 lina RPV S Internals Mass 2.209 x 106 lha Initial Vessel Pressure 1,020 psia Initial Core Power (105% rated) 2.417 x 106 Btu /sec (2550 MWt) Initial Steamilow (105% rated) 3,055 lom/sec Initial CRD Flow 6.39 lbm/sec CRD Flow After Scram (PRPV = 0 psig) 21.5 lbn/sec CRD Enthalpy (From CSD) 68 Btu /lixn HPCI On Volume 8,749 ft3 HPCI Off Volume 10,494 ita Vessel Max P For Shutdown Cooling 150 psia RHR K in Shutdown Cooling 231.7 Btu /sec 0F RBR in Pool Cooling 231.7 Btu /sec 0F RHR Flowrate in Pool Cooling 1599 lbm/sec l RBR Pump Borsepower 1000 hp/ pump RHR Flowrate in Shutdown Cooling 1599 lbm/sec Service Water Temp 800F S/RV FLOW (122.5% ASME) P psia FLOW lbm/sec 0 0 1214.7 3 15.8 ENTHALPY Feedwater MASS lbm (pipe and fluid) 136,631 402.3 Bru/lbm 193,583 353 Btu /1ha ' 315,936 287 Btu /lbm l l l l 1 l l .. 1 of 1 Revision 5 - December 1981 l t

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FIG. J- 1 POOL TEMPERATURE AND PRESSURE VS. TIME SHOREHAM POOL HEATUP- CASE I A SHOREHAM NUCLEAR POWER STATION -UNIT 1 PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 5-DECEMBER 1981

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TIME AFTER SCRAM (SECONDS) FIG. J-6 POOL TEMPERATURE AND PRESSURE VS. TIME SHOREHAM POOL HEATUP- CASE 3B SHOREHAM NUCLEAR POWER STATION-UNIT I PLANT DESIGN ASSESSMENT FOR SRV AND LOCA LOADS REVISION 5 - DECEMBER 1981

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4 l (0] C]' MANUAL OPERATION 1r 1r [O NOTE: SHOREHAM RHR SYSTEM SHOWING THE EFFECT OF INBOARD RHR VALVE FAILURE. l MANUAL OPERATION OF FOUR VALVES REQUIRED TO PLACE " FAULTED LOOP" j INTO POOL COOLING MODE- DELAY ESTIMATED TO BE-1 HOUR.

               !    i- CONT AINMENT R - RED POWER (DIVISION I)

, B - BLUE POWER (DIVISION II) ( 0 - ORANGE POWER (DIVISION III FIG. J-7 RHR SYSTEM VALVE AND PUMP ARRANGEMENT

   ,                                                                SHOREHAM NUCLEAR POWER STATION - UNIT 1 PLANT DESIGN ASSESSMENTS FOR SRV AND LOCA LOADS l

l REVISION 5 - DECEMBER 1981 L}}

ML20023B720

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Dear Mr. Kniel:

Dear Mr. Denton:

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