ML20008E681

Forwards Addl Info Re Seismic Qualification Review,Per NRC 810228 Request.Info Suppls 810130 Info

ML20008E681
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	03/05/1981
From:	Novarro J LONG ISLAND LIGHTING CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
SNRC-535, NUDOCS 8103090431
Download: ML20008E681 (20)

Text

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g LONG ISLAND LIGHTING COM PANY

~~/M N#N SHOREHAM NUCLEAR POWER STATION P.O. BOX 604, NORTH COUNTRY ROAD e WAC*NG RIVER. N.Y.11792 SNRc-535 March 5, 1981

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w' ,* 44, 50 Mr. Harold R. Denton, Director ',

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Office of Nuclear Reactor Regulation p U.S. Nuclear Regulatory Commission k / -Il-Washington, D.C. 20555

,g Seismic Qualification Review "

Response to Request for Additional Information '" / s Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322

Dear Mr. Denton:

Forwarded herein are fifteen (15) copies of our responses to information requests 271.1 thru 9, inclusive. These information requests were transmitted to us via Mr. R. L. Tedesco's letter dated January 28, 1981.

This information supplements the information provided in SNRC-531, dated January 30, 1981. If additional information is required, please contact this office.

Very t ly yours,

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[5YZUM i

l J. P. No arro i

RAH /pd Enclosures cc: J. Higgins Dr. Morris Reich l

s 810309048 1

SNP3-1 FSAR Request 271.1 Pursuant to General Design Criterion 2, safety-related structures, systems and components are to be designed for appropriate load co'linations arising from accidents and severe natural phenomena.

With regard to the vibratory loads attributed to the feedback of hydrodynamic loads from the pressure suppression pool of the containment, the staff requires that safety-related mechanical, electrical, instrumentation and control equioment be designed and qualified to withstand effects of hydrodynamic vibratory loads associated with either safety relief valve (SRV) discharge or LOCA blowdown into the pressure suppression containment combined with the effects of dynamic loads arising from earthquakes.

The criteria to be used by the staff to determine the acceptability of your equipment qualification program for seismic and dynamic loads are IEEE STD 344-1975 as supplemented by Regulatory Guides 1.100 and 1.92, and Standard Review Plan Sections 3.9.2 and 3.10.

! State the extent to which the equipment in your plant meets these requirements and the above requirements to combine seismic and hydrodynamic vibratory loads. For equipment that does not meet these requirements, provide justification for the use of other cri-teria.

Response

All Shoreham safety-related mechanical, electrical, instrumentation and control equipment will be designed and qualified to withstand l the effects of hydrodynamic vibratory loads combined with the ef-fects of dynamic loads arising from site OBE and SSE seismic ef-fects.

l In your request for additional information, dated January 28, 1981, i the staff has indicated their intention to determine the ac-l ceptability of the Shoreham equipment qualification program for seismic and dynamic loads based on criteria contained in IEEE STD 344-1975 as supplemented by Regulatory Guides 1.100 and 1.92, l and Standard Review Plan Section 3.9.2 and 3.10. As we desc~;ibed

! in the November G, 1979 and January 13, 1981 neetings with your j staff, we defined our program as being in conformance with Branch i

Technical Position B.1 of EICSB 10. That position states that for plants docketed before October 27, 1972, and for which operat-ing licensing reviews are not completed, acceptable methods for ~

qualifying electrical and mechanical equipment are contained in IEEE STD 344-1971. The Shoreham equipment qualification. program meets and exceeds NRC criteria applicable to plant design.

271.1 February 1981

SNPS-1 FSAI; Reevaluation of Shoreham equipment has included consideration of criteria contained in IEEE 344-1975 and the referenced Regulatory Guides and Standard Review Plans. The objective of reevaluation of Shoreham equipment is to demonstrate compliance to these criteria ,

as well as consideration of the effects of hydrodynamic loads.

271.la February 1981 mh

SNPS-1 FSAR Request 271.2 Provide the following information:

(i) Two summary equipment lists (one for NSSS supplied equipment and one for BOP supplied equipment). These lists should include all safety related mechanical components, electrical, instrumentation, and control equipment, including valve actuators and other appurtenances of active pumps and valves. In the lists, the following information should ba specified for each item of equipment:

(1) Method of qualification used:

a) Analysis or test (indicate the company that prepared the report, the reference report number and date of the publication).

b) If by test, describe whether it was a single or multifrequency test and whether input was single axis or multi-axis.

c) If by analysis, describe whether static or dynamic, single or multiple-axis analysis was used.

Provide natural frequency (or frequencies) of equipment.

(2) Indicate whether the equipment has met the qualification

. requirements.

(3) Indicate the systaa in which the equipment is located and whether the er uipment is required for:

a) hot stand-by b) cold shutdown c) both d) neither (4) Location of equipment, i.e., building, elevation.

(5) Availability for inspection (Is the equipment already installed at the plant site?)

(ii) An Acceptable scenario of how to maintain hot stand-by and cold shutdown based on the following assumptions:

(1) SSE or OBE (2) Loss of offsite' power

-(3) Any single failure 271.2 February 1981

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SNPS-1 FSAR (iii) A compilation of the required response spectra (RRS) for all applicable vibratory loads (individual and combined if required) for each floor of the nuclear station under construction.

Response

271.2(i) A preliminary version of these equipment lists was submitted to the NRC during our January 13, 1981 meeting. To facilitate the selection of specific equipment for review by the Seismic Qualification Review Team (SQRT), and in response to your letter dated January 28, 1981, a revised version of these lists was submitted via SNRC-531, dated January 30, 1981. That submittal contained complete report numbers and equipment frequencies, sorted on report numbers rather than on the equipment mark number.

Where possible, information requested in information request 271. 2 (i) (1) thru (5) has been provided. As agreed at the January 13, 1981 meeting, an updated form of these lists will be sent to the NRC prior to the March 30, 1981 review.

271.2(ii) This information has been provided in SNRC-368, dated March 20, 1979 (copy attached).

271.2(iii) A compilation of the typical amplified response spectra for all applicable vibratory loads (seismic plus hydrodynamic) for each building and elevation was submitted via SNRC-376 dated April 9, 1979.

Revision 1 to this compilation will be submitted in March of 1981. Changes are listed below.

1. Spectra for the shield wall, pedestal and reactor vessel have been added.

2. The seismic spectra at each floor elevation have been incorporated into the reactor building spectra rather than the worst-case seismic spectra used in the original issue.

3. The hydrodynamic spectra have been revised in the upper frequency range based on improvec building dynamic analyses.

4. Screenwell spectra have been revised and are pro-vided for two elevations.

These spectra are the required response spe tra for the BOP equipment. They envelope the hydrodynamic sub-events and combine the individual dynamic events spectra by SRSS, to arrive at a single horizontal and a single vertical spectrum for each elevation in the reactor build-ing. NSSS equipment qualification is to individual event spectra as well as seismic SSI (soil- structure interaction) spectra.

271.2a February 1981

SNPS-1 FSAR Request 271.3 Identify those items of nuclear steam supply system and balance-of-plant equipment requiring reevaluation and specify why reevaluation is necessary (i.e. because the original qualification used the single frequency, single axis methodology, because equipment is affected by hydrodynamic loads, or because both of the above conditions were present) for each item of equipment.

Response

All Shoreham safety-related equipment was reevaluated to determine its ability to withstand the effects of combined seismic and hydrodynamic loads in conformance with applicable referenced criteria.

271.3 February 1981 f

SNPS-1 FSAR Reguest 271.4 Describe the methods and criteria used to determine the acceptability of the original equipment qualification to meet the required response spectra of item 2. (iii).

Response

The methods and criteria used to determine the acceptability of the original Shoreham equipment qualification to meet the required response spectra were first presented and reviewed with the NRC at November 8, 1979 (16 issues ) and January 13, 1981 (3 additional issues) meetings with your staff. These methods and criteria are summarized in Attachment I. Detailed criteria are provided in the FSAR.

271.4 February 1981

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- SNPS-1 FSAR t

i Request 271.5 ,

Describe the methods and criteria used to address the vibration fatigue cycle effects on the affected equipment due to required loading conditions.

Response

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Fatigue is not significant for equipment su'ojected solely to a seismic environment because of the limited number of stress cycles (less than 1000). For equipment in the reactor building which is l subjected P.o hydrodynamic loading, fatigue is not significant be-cause the nlmber of high stress cycles is still numerically small (less than 2.0,000). The response levels associated with the larger number of cycles due to a single SRV cycling are only fractional j values of the design basis multi-valve events. Furthermore, the equipment response levels are limited due to filtering by the piping, structural framing, and by the local compliance of the support attachment.

It is also noted that the stresses are maintained within elastic limits which are well below the fatigue limit at 10,000 cycles of maximum stress.

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271.5 February 1981

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1. i SNPS-1 FSAR l

_ Request 271.6 i Base? on the methods and criteria described in items 4 and 5, provide the results of the review of the original equipment qualification with identification of (1) equipment which has failed i to meet the required response spectra and required requalification,

] and (2) equipment which was found acceptable, together with the necessary information to justify the adequacy of the original l ' qualification.  ;

i i Response:

The results of the review of the original equipment qualification presented in Attachment II, shows that 82 percent of Shoreham equipment was found acceptable. The qualification status of each item is indicated on the equipment lists submitted via SNRC 531, dated Jr.nuary 30, 1981. l

Adequricy of the original qualification is justified in the qualification reporbs also identified in the equipment lists.

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271.6 February 1981 4

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SNPS-1 FSAR Request 271.7 Describe procedures and schedule for completion of each item identified in item 6. (1) that requires requalification,

Response

The schedule and procedure (method of qualification) for requalifying the remaining 18 percent of Shoreham equipment is presented in Attachment III. A more detailed schedule for equipment requalification will be submitted to the NRC prior to the March 30, 1981 audit.

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SNPS-1 FSAR Request 271.8 Describe plans for a confirmatory in-situ impedance test and an in-plant SRV test program or other alternatives to characterize the ability of equipment to accommodate hydrodynamic loading.

Response

No specific comfirmatory in-situ impedance testing or SRV test programs are planned for Shoreham since our current program adeque'aiy qualifies affected safety-related equipment.

271.8 February 1981

SNPS-1 FSAR Request 271.9 To confirm the extent to which the safety related equipment meets the requirements of General Design Criterion 2, the Seismic Qualification Review Team (SQRT) will conduct a plant site review. For selected equipment, SQRT w.ll review the combined required response spectra (RRS) or the coabined dynamic response, examine the equipment configuration and mo'nting, and then detcrmine whether the test or analysis which has been conducted demonstrates compliance with the RRS if the equ pment was qualified i

by test, or the acceptable analytical criteria it qualified by analysis.

The staff requires that a " Qualification "ummary of Equipment" as shown on the attached pages be prepared for each selected piece of equipment and submitted to the staff two weeks prior to the plant site visit. The applicant should make available at the plant site for SQRT review all pertinent documents and reports of the qualification for the selected equipment. After the visit, the applicant should be prepared to submit certain selected documents and reports for further staff review.

Response

A " Qualification Summary of Equipment" form will be prepared for each piece of equipment selected by the NRC for review and submitted to the NRC two weeks prior to the plant site visit. The presently scheduled visit of March 30, 1981, will require that the NRC advise the applicant by March 2, 1981 of those equipment items selected for detailed review. Completed forms for each selected equipment item will be forwarded to the NRC by March 16, 1981.

All pertinent documents and reports of the qualification for the selected equipment will be available at the site for SQRT review.

After the visit, other selected documents and reports will be made available t.o the NRC as required.

271.9 February 1981 J

PAGE 1 of 3 i

l.

i ATTAOi.'G T I

! I:RC/S03T E0UIP!!E?ti OUAL!?ICAT10:1 POE!!:C I SHO'EMI'l -

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OUAL.

!!P.C l.:Eli10D i ISSUE CPlTEP.lA AFFECTED Si!0REHl.!1 FOS!T10:1

! 1 IlYDP.0DYi!A!.i!C S SEPT. 78 1.LL REACT 03 EUILD!::'i ACS 1.':CLUDES fiX !!

! LOADS & LOADS. SU::lliTED 10 fi..C U::CER LETTER I 23 FES. 79 10 f.13. H. R. DENIO:1 DATED 9 A??.!L.1979

LETTERS SNRC 376.

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,1 2 00P/liSSS S02T REYlEYi ALL BASIS IS 1:iDi'!! DUAL EVE?iT SPECTRA PER J CO:.itl*:ED AT GE - DAR. DOC".ETED 221l.:1.1979, 0 SPECTRA 8 JU: E 79 SNRC 355.

LETTER

} 3 TO VS RH SORT REVIEYl ALL LETTER TO !!R. H. R. DENTON. S!!P.C.433, I IN LO7f AT GE - 23 OCT. 79. SHO?!S FP.EQUEl;CY S JU:iE 79 RAMSHEAD IS COU::Dl::G.

l PJ.!;GE LETTER SPATIAL 344 1975 fflALYSIS 20 RULES L'.P.GER OF 2 HOF.!ZO:!TM.

C0:.1CINATIO!! 1.92 RESP 0."SES CO:/C;::ED 'lilTH THE YEF;i: CAL RESP 0 iSE SY /.BSOLUTE SU:J.

) 5 C10SELY S?/.CED - 1.92 /JIALYSIS CLOSELY SPACED !.!CDES ARE USED F02 L!0 DES -

DYNA .'IC A :Al.YS:S CF .$E!C!03 CU:!.C:::G EQUIP aENT FO.i .U' 11 LOADS.

6 Lis 11 LOAD 23 FEB. 79 /JIALYSIS & SRSS PER THE DESIG:( ASSESS?.iE?iT i:EFOT.T C0!.13:l!AT10N TEST SUD.'.tlTTED TO THE ?!2C 221A:1. 79 LIET110D SNRC-355 AND !!ED0 21C313 JU:1.197S (DFFR) 7 OPEP. ABILITY 344'1975 ALL ANALYSIS DEFLECTION CALCULAT10:1 SY 3.9.2 DYllA?!!C A.'iALYSIS OF !.!ECHAhtCAL 3.10 - EQUIP!.!EliT.

TEST !.!O?ilTOR SAFETY-RELATED FU:!CT 0NS BEFORE A::D AFTEi: E';E:iT. */03l TOR DURI?;G EVENT IF SAFE.SHUIDG7/N CAPA-BILITY IS TH.1EATE!iED BY INTERRUPT!0N OF FUNCTION.

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' . PAGE 2 or 3 1 ATTACHMEi;T I I l

? fiP.C/50'iT EQU:P*tEllT QUAT.IFICATiG:i FOSITION SHC':EHA!.t i

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. QUAL. '

l  !!RC l.tETHOD lSSUE CTITERIA AFFECTED SHOP.EHA?.I FOSITIO!!

l l 8 SI'.!!U,ElIY SQ3T REYlEVI ALL YAllD IF ECU!'"E:li ITE.'.!S AP.E CEO.'.tET21-

- AT GE- CALLY SI:.;iLAR.

i S Alli. 79

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} 9 GEl:EE!C !!SSS 23 FEB. 79 ALL PUJIT SPECIFIC.

QUAUFICAT10:1

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h 10 STATIC 1.100 AllALYSIS A l.ti:!!!.tU .t FACTC3 0F 1.3 HAS EEE!!

COEFFiCIEI:T 244 1975 YEEiF1ED SY 07:".'.t:01.::'. LYE!E. ALL SIGlilFICArii !',0CES 1::CLUDED.

11 STATIC 3(4 1975 A!!ALYS!S ALYlAYS ACCEPi!. LE TO '.'E.1171ED ECU!'!A-

/JiA!.YSIS 3.9.2 LEliT STAlic LC.'.3 <STl.Tll CCEFFIC27 T)

FOR S!!.lPLE l.'EC;-!A? iCAL EQU ??.:E:li.

2 DY!!.'.!.11C 344 1975 A!!ALYSIS ACCEPT l.0LE 1:1 CESC::.'.::T R.'.':'JE BY

/Ji!. LYSIS 3.9.2 RESPO!!SE SPECi:'A. ALYiAYS ACCE?TASLE BY TI.'.lE 1:!ST00Y.

13 Sl!!GLE 344 1975 TEST ACCEPTACLE FEA:(ED SPECTP.A. Y!!DELY FREQUE? ICY 3.10. 1.100 SPACED !.'0 DES. C:iLY 0 :E 03 TY D

$1:GLE AXIS 23 FEB. 79 $!G.:! FICA:IT "0ES. I::E2EFC':E. "ULTI.

TEST MODE EXCITAT!C:i 0I Et?02illiT. TEST I LEVEL IS ZPA X COU?L!:iG FACIC2.

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g 14 SINGLE 344 1975 TEST RECilll ! EAR f 45" TEST FlXTUE) !T.EATED

., FREQUE!:CY 3.10 SA!.tE AS SI:iGLE FTIQUE::CY - S!!!GLE "-

51 AX1AL TEST A..X.IS

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15 f.!Util- 3441975 TEST RAllD0!.! CIAXIAL 1:iC0!!EF.E!iT F2EFET.EED.

IT.EQUEf CY 3.10 TRS 1.tUST EriVE'.0P RCS EETWEEN EOU!?-

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TEST LtEIII FUi:DA .!EliTAL FREQUEliCY &

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2P'A ASY!.!PIOIE.

RRS DOES IIO! E.X!ST IT THE FL':!D's.'.tENTAL FREQUE!:CY > THE ZPA ASY.'.!PIOTE.

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I 16 DYllAMIC 344 1975 TEST SlflGLE FREQUEi:CY TESTl!!G PREFEF.2ED A.'!ALYSis & 3.10 Slf!CE fic R3S IS lJ/A!LAULE AT DEY!CE i

DEYlCE TESil!G 3.9.2 f. IOU?iTI:iG EASE. '!Ulil.FREOUE ICY IEST!:iG OK IF RACT 15 RELAll','ELY RIGID.

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ATTACHMENT II

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QUALIFICATION STATUS OF SHOREHAM EQUIPMENT TO SQRT REQUIREfENTS

. E0T. QUAL. QUAL. TOTAL  %

. DESCR. COMPL. INCOMP. NO. EDT. - QUAL.

NSSS .

27 15 42 64 MECH ELEc 107 5 112 96 154 B0P Misc 64 0 64 100 CB 151 6 157 96 -

, RB-FM 177 28 '205 86 RB-PM 197 98 295 67 721

ATTACif1ENT III REVISED REQUALIFICATION SCHEDULE TO NRC REQUIREMENTS EQUIPMENT NUMBER METHOD COMPLETE MECHANICAL 27 ANALYSIS 6/81 3 ANALYSIS / TEST 12/81 ELECTRICAL 23 TEST 6/81 1 TEST 12/81 MOTOR OPERATED VALVES 98 ANALYSIS (YOKES) 6/81 TEST (OPERATORS) 12/81

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f ATTACllCE.lT 4 SilOREHAM NUCLZAR PC**ER STA? !ON - UNIT 1 PAGE I 0F ;,

SORT SCENARIO POP ACHIEVING COLD FiiUTDOWN (PREVIOUSLY PROVIDED AS AM ATTAC11 MENT TO SNRC-368, DATED MARCH 20, 1979)

The following is a chronological listing of the anticipated actions which would be required to safely shut doun the Shoreham reactor under the scenario that the following events were to occur simultaneously:

1. Safe shutdown earthquake

2. Loss of all off-site power

3. Loss of one RHR outboard injection valve lEll*MOV-036A or B In the preparation of this listing, the following events were presumed to occur automatically on the loss of all off-site power:

1. ' Reactor scram -

2. NSSSS isolations

3. Main turbine trip

4. Diesel generators start and supply emergency switchgear

5. RBSVS and CRACS startup

6. RCIC and HPCI injection if low level reached The following procedure could be used to achieve cold shutdown from normal operating conditions given the above:

1. Start *RCIC (if not automatically initiated) to maintain reactor level as follows:

. a. Arm and depress RCIC manual initiation switch on 1Hll*PNL-602.

b. Control the RCIC pump flow with lE51*FlC003 on 1Hll*PNL-602 as necessary to return and to maintain vessel level.

2. Control reactor pressure by relieving ADS relief valves to the suppression pool from 1Hll*PNL-601.

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, . A T T A C H M E i1 T L: PAGE 2 0F 3

3. If operation of RCIC is inadecuate to restore reactor vessel level, start HPCI (if not automatically initiated) for vessel injection as follows:

a. Arm and depress the Manual Initiation Switch on lHll*PNL-601.

b. Control HPCI pump flow with lE41*FIC-003 on lHll*PNL-601 in auto or manual as necessary to restore vessel level.

4. Establish suppression pool cooling using the remaining RHR loop as follows:

a. Open the RHR heat exchanger service water system discharge valve 1P41*MOV-034 A/B at 1Hll*PNL-MCB and verify service water flow on lEll*FI-006 A/B.

b. Close LPCI injection valve lEll*MOV-036 A/B.

c. Close RHR heat exchanger bypass valve lEll*MOV-034 A/B.

d. Open the suppression pool cooling valve lEll*MOV-040 A/B.

e. Start Lhe applicable RHR pumps by placing the appropriate gontrol switches in the AUTO-AFTER-START position on

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lHll*PNL-601.

f.. Open suppression pool inlet valve lEll*MOV-042 A/B as desired to establish the desired flow rate on lEll*FI-001 A/B on lHil*PNL-601.

g; Operate the RHR system to maintain suppression pool temperature.

5. Cool down the reactor by RCIC and HPCI operations. Do not exceed a cooldown rate of 100 F/hr.

6. Restart each RPS MG set by depressing and holding its start pushbutton until speed and voltage are established, then reclose its output breaker. ~

7. Reset the NSSSS isolations with switch S-32 on lHll*PNL-602 and switch S-33 on 1Hll*PNL-601.

8. Secure the HPCI pump when the low steam pressure isolation is received.

9. Fill the RPV using the RCIC purn %dtil RPV pressure falls below 60 psig.

ATTAClinE!lT I: PAGE 3 0F 3

10. Discontinue suppression pool cooling and commence a reactor cooldown using the shutdown cooling mode of the RHR system as follows:

a. Close suppression pool cooling valve lEll*MOV-040 A/B and lEll*MOV-042 A/B.

b. Stop the operating RHR pump (s)

c. Open RHR heat exchanger bypass valve lEll*MOV-034 A/B.

d. Close pump suction valves lEll*MOV-031 A and C/B and L.

e. Open shutdown cooling valve lEll*MOV-032 A or C/B or D

f. Open shutdown cooling suction valve lEll*MOV-048.

g. Open shutdown cooling suction valve lEll*MOV-047.

h. Open injection valve lEll*MOV-037 A/B.

i. Start 1 RHR pump by placing control switch in AUTO-AFTER-START on lHll*PNL-601.

j. Open 1E41*MOV-036 A/B and throttle to achieve the desired flow rate.

k. Establish cooldown rate by throttling closed the RHR heat exchange bypass valve lEll*MOV-034 A/B.

11. Maintain RPV temperature between 180 F and 200 F using shutdown cooling.

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