ML20023B717

Submits Schedule & Program for Analyzing & Resolving Effects of LOCA Re Suppression Pool Hydrodynamic Loads,Safety Relief Valve Vent Clearing & Steam Quenching Phenomena. Schedule Coordinated W/Ge & Other Utils

ML20023B717
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	06/03/1975
From:	Wofford A LONG ISLAND LIGHTING CO.
To:	Stolz J Office of Nuclear Reactor Regulation
Shared Package
ML20023B716	List: ML20023B715 ML20023B717 ML20023B718 ML20023B719 ML20023B720 ML20023B723
References
FOIA-82-455, FOIA-82-471, FOIA-82-479 SNRC-50, NUDOCS 8305060153
Download: ML20023B717 (4)

Text

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LONG~ iSLAN D LIGHTING CO M PANY w:.w

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Stolz, Chief 0

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Division of Reactor Licensing

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U. S. Nuclear Regulatory Commission Washington, D.

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20555 BWR Mark II Containment Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322

Dear Mr. Stelz:

In accordance with our letter of May 15, 1975, we hereby submit the schedule and program (Attachment A) for analyzing and resolving the effects of LOCA related suppression pool hydrodynamic loads, safety relief valve vent clearing and stcam quenching phenomena.

This schedule has been coordina-ted with the General Electric Company and other utilities building Mark II containments.

We view the specific loads identified in your letters of 3

April 18 and April 21, 1975, as possibly inter-related, and we intend to analyze them concurrently.

Much of the information presently available on these loads is related to other containment designs, and cannot be readily extra-2

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polated to the Mark II containment.

The initial assessment of the Shoreham containment must, therefore, await a deter-mination of the time history sequence of the various J

the forcing functions which result, and the phenomena, relationship of these functions to each other and to the

..i specific geometry of our containment structures and compo-nents.

Preliminary Mark II time histories, forcing functions

.1 based on available data, and the methods for relating these to the Mark II containment are being developed now and will be available in September as the " Preliminary Forcing Function 3

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This will then pomit the prelimine.:y centninment cccensment te procnnd, and wo expoct the resulto of this ccccccrr.ent to be available in November, 1975.

Certain tects and raathematical models will be helpful in verifying the preliminary assesnment.

Although some of this effort cca be identified now, the Preliminary Forcing Function Reoort and the initial containment ecscurmont may reveal nddi*;icnal testing and modoling requirements.

A complete s :hedule we'uld not, therefore, be feacible until D : c e:-b e r, 1975.

Once the test program has been defined t.nd scheduled, we will establish and submit a schedule for-the final assessment of our containcent design.

Construction of the supprension pool structures, including.

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the foundation mat, walls, liner, drywell floor, and reactor support pedestcl is essentially complete.

Although'.

some miner nodifications may be required to mitigate the conceauences of the subject phenomena, we feel that the need for major structural modifications is exceedingly remote.

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Very truly yours,

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A. W. Wofford Vice President O

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ATTACll:1ENT "A" MARK,II SCilEDULE & PROGRM1 LOCA SRV Schedule Ouestions Questions Program 1.

Submit plant unique suppression July, 1975 1

1,5*

pool and relief valve drawings reflecting current design.

2.

Submit a generic " Preliminary Forcing Sept., 1975 2,7*

1*,2,4*,6 Function Report" which will. provide the time history of pool dynamic forcing functions and the methods for relating these functions to the con-tainment structures and components.

3.

Submit a description of suppression Oct., 1975 Not Appl.

7,8,9,10 pool temperature monitoring system.

(Temperature limits and transients will be described in the Final Safety Analysis Report.)

4.

Submit a plant unique preliminary Nov., 1975 3,4,5,6,7*,E 3,4*,5 assessment of containment structures and components based on the Preliminary Forcing Function Report.

i Partial Answer l

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ATTACllMENT "A" MARK II SCllEDULE & PROGRAM LOCA SRC

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Schedule Questions _

Ouestions_

Program S.

Submit a schedule for the generic Dec., 1975 7

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test program and mathematical models which justify the forcing functions used to assess contain-ment structures and components.

First Quarter 6.

Submit a schedule for the final as-1976 sessment of containment structur,es a

J and components.

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L,.-ff-,e3 LONG EC LArg.O LgGg_; TING CC M PANY SHOREHAFA PouCLEAR POWER GTATION

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May 29, 1931 SNRC-578 b~ KY4 Ns i

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Denton, Director y

Office of Nuclear Reactor Reculation II D th

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Shoreham Nuclear Power Statien - Unit 1 Docket No. 50-322

Dear Mr. Denton:

{orsardedherewitharesixty (60) copies of L LCO's responses aluation Report (SZR) Outstanding Issues listed

]eagenotethatourresponsestooutstandingIssueNu:.be's9

,3-}'[1'0"O.gn*

Oualification" and C57, "TMI-2 Requirc=ents" h$ve

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.o you uncer separate cover via letters SSRC-576 SNRC-579 respectively, dated May 29, 1981.

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' ?roject.v.anacer Shoreham NucAear Power Station - Unit 1 CC/pd Enclosures cc:

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Higgins THIS DDCt!MfllT CONTAINS POOR QUAllty pg.g3

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AT""1 C'.92ST 1 - SER OUTSTA'; DING ISSUES 3

s Number Issue j

Appendix G - IV.A.2.a e Nil Ductility 20 Te=perature 21 Appendix G - IV.A.2.c - Pressure Temperature Limit 22

• Appendix G - Impact Testing Appendix G - IV.3 - Minimum Upper Shelf 23 Energy 35 Containment Isolation i

37 Secondary Containment Bypass Leakage 35

• Fracture Prevention of Containnent Pressure Soundry l

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Fracture Toughness of Steam Line and l

51

• Feedwater Materials i.

52 Management Organization The infor: scion provided in this response supplenants the i

nformation provided in SNRC-566, dated May 15, 1981.

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The infor=.ation contained on the following pages is provided in response to the staff concerns identified as Shorchnm SER Dutstanding Issues 20, 21, 22, 23, 38 and 52.

This data provides an adequate basis for resolution of the staff's concerns.

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S i. e.I.i. '_' : u...,\\__._L LELTLI::E PLATE 720 WELD i

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1.

Available charpy V-notch and drop-weight 1:DT data are presented i

in Tables 1, 2, and 4 f or Shoreh;e beltline plates cnd walds.

'Iable 2 gives supplementary transverse charpy results which were determined for one of the Shoreham surveillanca plctes.

l Table 3 shows a typical test certificate for a Shorch:2 beltline plate.

Table 5 verifies the location of weld materials in

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beltline weld se.-

. as well as verifys.w what data as not i

available.

I 2.

The beltline layout is shown in Pigure 1.

This gives plate heat nu., bars and locctic'

'e wall as veld sec= locations and t

2dentifications.

3.

Copper and phosphorous values, to estinste the effects of radiction on toughness, are presented in Table 6, when available.

It can be scan in T ble 6 that there are cases for icngitudinal weld setns where copper contents are not available.

This has been verified as shown in the letter from the vessel vendor 4

(Table 5).

4 Estinated starting (unirradiated) RT values are given in y;

Table 6.

They are estinated by using t..e data in Tables 1 and 4 in accordance witi GE procedure Y1000;>006 which c:::s the intent of ASME Code paragraph 10-2300.

This procedure is explained in paragraph 5.2.4.3 (1sttachrant A) of the S!.75-1 FSAR, Rev. 18.

The data base for thic pro: dure is further clarified in response to Zi:: 2r (CPS-1) 0 121.15.

5.

Estin ted end-cf-life (E3L) RT 13.;. values (for 1/4 thichness location frc: the versel inside dicn2ter, as-bui2t dia n;iens) are given in Table 6.

There estinaticns are in..ccord_c.02 with 1;RC Kagulctory Guide 1.99 Rev. 1.

C.cre Cu rontent analyscs l

are not available, tuu:inum ET 1:pT shift ( A R'. 3;y.) valuas cre 1.

l conservatively assu= d in accorcance with Reg. Guide 1.99 Rav.

This results in liciting EOL RT 33 values for the longitudinal weld seems in the beltline. Note snat the Shcrch;= vec;el sur-veillance progrtn is designed to represtat 1cngitudinal vald se s.:n 1 - 3 0 E J, although the e nct scre wald nnterials were not used.

This will be clarified in more detail in the description i

I of the Shorchcn surveillance progrcm in this response.

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Onarpy T-notch upper thcl: tcughncLa ucc not c rtquarc. nt when the Shorchcm vencel uns ranufcetured.

Thus, such dsta

• l, is not available for the Shorcham beltline ucids, but is available for the plates as shown in Tcbics 1 cnd 2.

A very conr.ervative assurption of 654 f actor en longitudinal upper chelf can be applied to the results of Table 1 in order to estinate transversa orienta'eion uppar shelf.

(Table 2 shows that an est factor may_ba more cecurate than the con-servative 651 factor of MTE35-2).

The factor of 65% uould result in a longitudinal requirc=cnt of 115 ft-lb, in order to meet the IOCTI: 50 Appendix G value of 75 ft-lb, transverse upper shelf.

This value is nat by all plates in Table 1 except C4806-2, which has an average of 107 ft-lb.

However, since the Cu content of this plate is only 0.151 (Tabic 6) 1 a reduction of upper shelf cf only 20% is concervatively predicted by Reg. Guide 1.99 nev.

1.

Cc=bining thana 2 con-1 servative factors of 65% cnd 201 results in en initial longitudinal uppar shelf value of only 96 ft-lb, to maat the goal ci 50 ft-lb. trcncve;ae upper shelf at EOL.

This value of 96 ft-lb., as calculated in the following equation, is exceeded by plate C4806-2.

(.20) [.65 (L) )

50 =.65 (L)

(where L is the longitudinal upper shelf value at start of life) 1 As seen in Table 4, upp2r shelf toughnecs values are not availchle for Shorchem valdc.

However, all chnrpy results at the test j

te=parcture of +10* cre in excess of tha 75 ft-lb, value for the E0018 ucid catericle.

Thus, th:y chould not b.* a cencarn.

The chcrpy values for the sch::rg:5 cre uald caterials in Tcbic 4 do not n20t th 75 ft-lb. loval in 4 cares.

Couaver, it is cupceted th;t further tccting at higher tt: ;eratures vould have rcvecled cn upper chalf in ene :s of 75 ft-lb.

Evid:nce in this recpact is pre :nted in Tcbics 7 through 11 which shou 4

wold procedurcs cnd upp;r chcif toughncus results for similcr matericls.

All upper shelf (^"100% chatr) re:ults in Tcbles 8 and 11 cre in enecco of 75 ft-lb.

Thace uald: cre considered to be reprecentative of the Shorchc= ucids cince the ualding (generally tendc= uire sub OrgcJ cre for the bulk of ~

procccces the velds), post veld heat treatt:nt, cnd u21d catericle are similcr.

Particulcr cttention chould ba given to tha L:Salle 1 results, since thecc walds ucro cada by th' - -' vendor cnd uith tha c:::ct scr : veld procedure 3

i (Ccmbuction Engin:cring) in n:.ny ccccc, cc for Chorchta. Tha r atarici 1P257 A/"

i (Tchic 7)

I 3950 in Tcble D is in hoth the L Calle 1 cnd Cher htm curvcillcnce progrens cnd unc prepared by the uald prc:_ dure in Tcbis 7, which was used for the Shorcht= longitudinal limiting walds.

Futher testing of these baseline curveillance spocie.cne gave an

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upper shelf of 110 ft-lb. as shown in Table 8.

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Drop-veight 107 values for the Shorch::n wcld materials were not determined by testing.

!!owsver, evidence for a con-servative cssumption of -50*r is found in Tcbic 8, based on the LaSalle 1 results.

All values of UOT cre -50*r or lover.

Futher results in thic respect are nico chown in Tcbic 11 of -50*r and lovar, c,:copt and verify NDT value)156 for L:quna Verde 2)

(CBIN ucids)

This case (lP6484/0 is for one case.

considered to be nonrepresentative of Shorcha:n, baccuss of the

. relatively low charpy test value (17 ft-lb.) at +10 and o*r for this material.

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r SHOREEAM REACTOR VESSEL NON-EELTLINE INFORF% TION 1.

Lir.iting RT NDT values which effect vessel testing and 4

operation are shown in the FSAR (paragraphs shown in Attach =2nt A).

The esti:r.ation procedurcs for thcce RT values are in accordance with GE procedure Y1006A006, NDT and are also explained in the FSAR.

As with the beltline, the data base for this procedure is further clarified in response to Zir=.er (ZPS-1) Q121.15.

2.

A sentence has been added to the FSAR (Attachment A, l

paragraph 5.2.4.3) to clarify further that these are limiting values for the vessel.

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t' ster quenchad.

(b) 1225'r + ?.5' 4 hrs.

Air cooled.

(c) 1150*P I 25' 40 hrs.

Purnc:e cooled to 600*F.

We i=p -ts ucre teken p r:11c1 to tha cojer rolling directica of the plate surfoco.

et ths 2./47 level ed no;ehed p::; 2ndicular to the picte na tene!.las were tchen in cecordance with Asm A-20-GO.

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