ML20063N435

Discusses Overpressure Events in Reactor Primary Sys.Impact of Transients on Structural Integrity of Reactor Vessel Evaluated.Related Info Encl

ML20063N435
Person / Time
Site:	Turkey Point, 05000000
Issue date:	12/16/1981
From:	Banford W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	John Miller NRC
Shared Package
ML20063C459	List: ML20063N435
References
FOIA-82-338 NUDOCS 8209210489
Download: ML20063N435 (8)

Text

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~W~ 2 wa / . / ; agJo r.m : Faterials Technology a 2 r-3157 O.2-j(,ff

%. Escember 15, 1581 s.m Turkey Point Unit 4 Overpressure Events .

Ref: 1. Chirigos J. N. and kyer, T. A., T

" Influence of l'aterial Prc~perty

'Vartations on the Assessment of Structural Integrity of Huclear Ccuperants", Journal of Testing and Iva'Juatien, September 197B.

n . J. C. Miller -

cc: 4, N. Chiriges B. J. Murray J. F. F.ndetto H. Paduano M. T. Kaisav R. L. Whitoey D. G. Faire .-

T. A. Heyer D. J. Voodruff ^

File: SH 13.7.0 (FLA) be everpressure events occurred rece.atly in the Turkey Point Unit 4 reactor prirary systea, and the purpose of this namo is to evaluate the impact of these two transients on the structural integrity of the reactor vessel. The pressure excursions which Occurred were to 1100 psig et 110'F and 750 psig at 105'F. Since the 1100 psig excursion was clearly the most limiting, it is be'.n treated in detail. .

The pressu[e temperature linit curves which were exceeded were generated accordance 5+ith Section III of the A5ME Code, Appendix 5 There are a signi-ficant number of conservatisms incorporated in the Appendix cluding (1) a safety facter of 2.0 on the pressure stress intensity G precedure, in-factor, (2) the use of XI g instead of the'more realistic K ie curve,and (3) the asstemtion of the presence of a quarter thickness flaw in all regions remote fica discontinuties. These conservatis.rs and others are discussed in detail in an article in the Journal of Testing and Evaluation [1]. Beyond these con-servatisms, the pressure temperature limit curves were Senerated using a con,'

servative analysis.

This evaluation will be carried cut to c'etermine the actual safety factor whict existed during the most governing overpressure event, while retaining all the, other conservatisms originally ecployed in the Appendix G analysis. Finite element stress analysis was employed where available. The limiting material of the vessel is the weld metal of a cirrt aferential seam in the beltline which I

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his the following material prcpertfas: (1) an initial RTNOT e N F and

'(2) of a shift.

5.ES x 108gfo/c:e.

%5*J tesed Thison surveilhacedata zurveiThnce capsule pointdata is beirs at a the ilucoce value predirtad by the Westinghouse tre::d c.urve, which is in turn mr :

coasarvatin than the recently determir,ed AST71 trtr4 curva. . As a rs~

s'. fit, the Wstinghouse trend cerve ws bred to p,redict a sMfy.in RTy of 172*F at a 1/4 thickness fleecce of -5.72 x 1923 Wcn.Z, the fluence at 5.65 effectiwt full power years. Tbes, the final RTyg3- is 175*F fer tMs dectneferential weld.

Festulatiren of a icogitudinally criented t.:arter thickness finw dn h dret:.rferential weld is the most limitfrig case, and the safety factar Wich results from comparing the stress -inteesity factor with the frac-ture tughness is 1.17. The aficneble pres.wre for this assumed flaw-is 1232 psig. Fr::xture calculatims were ude for a m.rber of other regions of the reactcr vessel, and these results are shm' in. the attached tr.ble. A shnificsnt conservatism in this analysis is that a Tc6gitudinal flaw was assta:ed to exist in the girth weld estarial. For a 1/4 thicAcess flaw, cnly a mall pcrtien of the len.gth of the flaw wculd actually be in

, the weld. 7.he cajority of the fir.w mid be in the Mgher tcughness base matal. .

A crre rulistic treet:r.ent of this .sve~ pressure event would imolve con -

sideration of a circumferetial flaw entirely witMn the g'irth weld. 7he only loc: tion fer ,-hich a lorsitudinal flaw is realistic is the base metal.

'{ 'Both these cases wre enaincd, and the results are also shcw in the table The results sf.w that the sc.verning: case new is the 1er.gitudinal flaw' in the bare rt.etal. The safety factor which results frca 'ccrparison of the stress i tmsity factor with theJteughness is r.cv 7.59, and the allcwable.9ressure is 1749 psig. Tharefore, e' realistic tre.at:ent of this ow.rpressurt e ent de rr.cstrates a rafety fattar ip the reactor tessel greator thui that' required fer a hydrotest. It is therefore clear that integrity of the,rwietar vessel was not froaired, even urider ..the assiratice of large defect sizes.

/

Furthamcre, these overprest6r,e events will Oct affect the fJ.tigue life of the vesseT even if it is asrhoed that both tesnsietts cccurred to 1100 psig.

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TURKEY PoltiT UNIT 4 STRESS INTENSITY FACTOR AT 1100 PSIG AND ASSOCIATED ALLOWA81.E PRESSllRE '

Allowable flaw Calculated Limiting K at 110*F Safety Factor Pressure R6gion Size -

K g(ksipn) RJHDTI .1), (ksiMn)

KIR/EI )- IP3I9) er Head Junction 0.25T 33.6 30e: 66,5 1.6L 1837 let Nozzle Junctt6n 0.20T 30.0 60 S2,5 1.75 1925

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tilne*** 0.25T 28.1 175* 31.6 1,12 1232 tom licad Junction 0.25T 26.5 60 52.5 1.98 2178 l c realistic beltlino cases:

hitudinalflow. l t metal 0.25T 28.1 . 83** 44.7 1.59 1749 1 .

2.24'

tumferential flaw, 0.25T 14.1 175* 31.6 2464 th wcld I0 n/cm2 at S.66 EFPY, Initial weld rnetal RTNDT

Is 3*F Itline fluence at 1/4 thickness is 6.72 x 10 L1d0.31wt% copper), ART is 172*F based on Westinghouse copper trend curve, and Final , ,

NDT JDT is 175"F ,

I0 n/cm2 at 5.66 EFPY, Iriitial base metal RT titline fluence at 1/4 thickness is 6.72 x 10 i 50'F (0.054 wt% copper), ARTliDT is 35'F based on surycillance capsule results, and Finoy0T r

NDT is 05'F. l ongitudinal flaw in girth weld. 'I 8 i - ,1

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the r a.eE short-ters cm g ascrizartan erests, tn HCO ::sig at IL . and to 750 psig at 10$"F, relactve to m- m with rmpert to the reacter p<w vessei (RFYJ. Secifically, I have re-vie.ed VestiW's Ietter to .y% os t2ris subject (MT-9E-2173) cad the l previcassTy drrteLHf ir.tJ'-E912 (ASME HE, Accusatix G Analysis cf the FTorida ihne- & Licht Co. Turkey Point Usits Ms. 3 and me. 4 Reac::ar Ve*_ set,. 3=ce 1577) and ASME 577C rewi r 3s.

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I w seith the M ccac!usion that the recmst events resulted in a factsr of srfety reir.t;ive~ta falitre, is tde pre.seace of the Wix 5 pestuisted j defects, of in -s of 1.5. Since 1.5 is ttne etquired factor of 5.2feky for bra wtatic preswee testirs, the structural intefity of the w .=T nsas acre % ired by this incident. Further, ee in the pencs of sec2s a pesta!ated f!Sii, t::e flas yu-th free these tac eve-Is eTd be w ui-metely 0.0001', aar i'nsipf ficar:t amo.est mee casaved tn the pestaTated 1.8" deep M ect.

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~itm:f, r fin:-fda MO Schjart: Terkey INrtnt Unit 4 Orcr'yessure Eves:ts Cer rw . PAeno-I h e W eses, the i <,. wt stcrt-ten, ciciwserizarf m esets to 1100 psig at ILTT arse to 750 psig at ir$*F, relattee to coa,ew-ccs ,with rw;ect to the reacter press.re vessel GWYJ. Specifically,. I have re--

vieucIf beinpc:.se*s previo,esTy cacheted Tetter to yo os tMs swhlect (NT-9E-Zi75) cnd ttst-Ficrida ber & Light Co. Turkey Foint *dsits EhM-6912 (A9fE IIE, A;cer#f x G hseT, Arc 1977) cred A9E EFYC rc=Wi rtss:ts. 3 e.r3Mc. 4JE m I r,vv u with tM W ccacIusica tut the n.unt events resgTted in a factar of safetyofretir, cterc, =_ti#ts e<w Falitee, of 1.5. in the ;:reser<e of the "gpesdix 5 postuisted for hydrostatic tsas not i P redi by perss;ra testirv3, t6 strtctts a! fnter7Since ity of time 1.5wmT is tne rup trais frcident.

pcst.4!sted f!ar, tze f f as c=cwth fr r:Festtyer , eees in the ;;resence of such a these tw evens edTd be apprcxi-me.:teTy 1.0 deep0.000L',

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R. Fehmo J, F. Er2rietto R. L. Witre D.5,F. fire U J. ! Win,ff T. A rete Fik: SN 13.7.0 (FLA) ikstfrdmse bs enluated th ' tis monssure eats stich recently ccmed darits prtsserizatica of th Terley Poir.t Unit ? Factcr Ccolat

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\v A frEtute nechnics analysis Ms be rede to eniuste t% presse ex- -

cursicns of 110) psig at 1104 ai M psig at 10?F, The welysis b5 te pertwed h accorduce with Sectica III of tk A2E Ccde, Appeedu G dich is the prc:ehre used to 9sn.srate alientle pescre-temerature .

Tirdts for rcrist tad upsat corditioss, Sirce 25 o'rerpressure event is

- e1 infrquent event, kstirfause caEs est hiien it sb1d te chssi-fied is a nonel or upset cWitica eed ceply with the Ippedh G lidts.

As a result, sone ecserntisc in th irpdix G iinits can M remad, Piene 1 frdicated th consenetisns which aist ir. tb kgdfx E lfsits ai tky it,cleit: (Da safety fetr of 2.0 on th pressm stress hte.si.ty '

.. .._._fccter.[T,b)(2) .th conserydve curra, ard 13 the 1/? thichess fke instead of ces cf tM a selkr  ! p, cu.w fra fis, This '

fracture echaics Brelysis is tased m a safety factar of 1,0 '

ea tobeA Iij)S,a stedy state tmperatete crditici,aird thre are n '

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The ituitiro caterial of t!e vessel is th vald cetal which hs the foilecing  ;

bg mterial preparties: (1) sa initial F.Tgy cf TF and (21 a shjft of 155*F hsed  !

ca stresnicr,ce cepsule f#ta at a fleen af 5,G x 10 n/e, Tirls streil-l lar.g ;lits pkt is hice tM vale 2 predicty) h th !Lstinpryse trd cers,'

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i sum is 1133 asig at a :rgerate of 110 ? '.2sg on the I&:E Agdh G j :riteia :iith a safe:y facter af 1.0 m [.Q}3 It 10FF, 65 311cabic '

! cresura is 1125 osig. An addi:icnal conseratise in this aralysis is int a lon;itgir.a; 'lah as assired :: exir, iri tre girts teld re.ttrial.

It would Mye Wert 6,grariste to asyn a cirer.ferentiel figw for this' l

! eid rate.is:, ard :r,is unic hhs bled 9.i eilcwile pressure. 32 l l attzched table lists tM stress intensity hctors ard the fr6chre tc#ess vale of the roterial at 110 F. 11u results frcs tk table sted tht tir:

iritgrity of tM reacht vessel as rot frpaired es undsf the 55 set 10ns

f a large dafe: sfu.

Asthera::re, these 0,earessurt ever.ts kill to: af's:t the fatigue lifa  ;

af the msel era if i is assrad ist Mit tritsfeats cccurred to '!

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December 28, 1961

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l Mr. James P. O'Reilly Regional Administrator, Region 11 U.S. Nuclear Regulatory Cormnission

, 101 Marietta Street, Suite 3100 l Atlanta, Georgia 30303 l

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Dear Mr. O'Reilly:

l REPORTABLE OCCURRENCE 251-81-15 TURKEY POINT UNIT 4 DATE OF OCCURRENCE: NOVEM8tR 28, 1981 TECHNICAL SPECIFICATION 6.9.2.n.1 RCS PRESSURE TRAN51ENT The attached Licensee Event Report is being submitted in occardance with Tecnnical Specification 6.9 to provide 30 day notification of the subject

occurrence, i

Very truly yours ,

.. A l N O [s.tf w . p l Lv J. W. Williams, Jr.

I Director ct Nuclear Energy PLP/cou 's .

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12.: ![ Re reactor was shutdown and the Reactor Coolant System (RCS) was in a i Ic iJ ! I water solid condf ef on with a ren.w rmenre and nremente of annrnvimar,1v f fTT71 l 1100r and 310 psig respectively. Two overpressure conditions developed i 3 , [ for which the Overpressure Mitigating System (OMS) failed to operate. his t E i is reportable nurnuant to TS 6_9.2_h_1. A af vef l ar overnremaure situation f

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. %eportable Occurrence 251-81-013 Licensee Event Report Page two Additional Event Description and Probable Consequences

• A. Conditions Prior to Occurrence The reactor was shutdown and the Reactor-Coolant Syst (RCS) was in a water solid condition with a temperature of approximately !!O and a pressure of approximately 310 psig. Pressure control was being maintained by charging into and letting down from the RCS. The letdown path was via the Residual Heat Removal System (RHRS) and the Chemical Volume Control System (CVCS) letdown Pressure Control Valve (PCV-4-145). Reactor Coolant System fimng and -

venting operations were completed and preparations were underway to start RCS heatup.

B. Description of Occurrence l

1. November 28,1981,10:35 p.m. --

l l

The 4B Reactor Coolant Pump (RCP) was started to began RCS heatup. The l Reactor Control Operator (RCO) noticed.that RCS pressure was approximately l

500 psig and increasing. He also noticed that PCV-4-145 was in the fully closed position and a'. tempted to open it in auto by lowering the setpoint.

i When this attempt failed, the valve ~was opened in manual,48 RCP,4A charging pump and the pressurizer control heaters were also secured. One Power Operated Relief Valve (PORV-4-455C) was manually opened to reduce RCS pressure. The other PORY (PORV-4-456) was isolated on a clearance. An RHR isolation valve (MOV-4-750) was found in the closed position and immediately manually opened. PCV-4-145 was returned to auto control and '

4A charging pump was restarted. The RCS pressure was maintained constant at approximately 335 psig.

The RCS peak pressure during the translent was 1100 psig. Duration of the overpressure condition was approximately two minutes.

The problem was Initially diagnosed as misoperation of PCV-4-145. When PCV-4-145 was returned to auto control with a pressure setpoint of 6.5 and RCS pressure maintained constant,4B RCP operation was again attempted by the next shif t.

l l Subsequent investigation indicates that the RCS pressure transmitter PT-4-403 had closed MOV-4-750, due to the pressure interlock at 465 psig, thus resulting l in the overpressure condition. It is common for the RCS pressure to surge during RCP startup. MOV-4-751 did not close, apparently due to PT-4-405 failing to reach the 465 psig interlock, it was later found that the instrument isolation valve associated with PT-4-405 was lef t closed following a hydrostatic test by construction personnel.

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R epor table Occurrence 251-51-015 1.w ensee Event Repor t Page three

2. November 29,1981,12:55 a.m. _

The 4B RCP was restarted. An overpressure condition reoccurred with peak pressure reaching 750 psig. Again the RCP and the charging pump in operation were secured. PORV-4-455C was manually opened to decrease RCS pressure.

Duration of the overpressure condition was approxirnately one minute.

Investigation of this condition found MOV4-750 and MOV-4-751 closed, thus isolating letdown. It was found that RCS pressure transmitter PT-4-405 was reading higher than the others. It is believed that PT-4-405 may have lost its setpoint due to hydrostatic testing of the sensing line with the block valve leaking through the seat. Since PT-4-405 was reading about 130 psig higher than the others, starting the RCP could have caused PT4-405 to prematurely close MOV-4-751 thus resulting in the overpressure condition.

During these two overpressure transients, the OMS failed to operate. Af ter the first event, the block valve to PT-4-405 was reopened. In addition,

' attempts were made to release the redundant OMS loop from clearance and restore it to operating conditions, but this was not accomplished by the time the second overpressure event occurred.

Additional Cause Description and Corrective Actions The primary cause of the overpressure conditions is considered to be the automatic closure of the RHR5 suction isolation valves coupled with the malfunctm of the OMS while operating in a water solid condition. The reason the OMS did not operate as designed was found to be a failed summator on the electronic circuitry coupled with the fact that PT-4-405 was isolated. This pressure transmitter provides input to the OMS circuitry to automatically open PORV-4-455C on high pressure conditions. The two devices which malfunctioned had previously been tested and calibrated.

During both occurrences, the operators took action to stop the charging pumps which were providing the source of rapid pressurization. However, once the letdown flow was significantly reduced or terminated by closure of the RHR5

. solation valves, timely operator action was precluded by the rapidity of the transient.

A f racture mechanics analysis based on the methods of Appendix G to Section 111 )

of the ASME Boiler and Pressure Vessel Code was performed by Westinghouse.

The analysis showed that the integrity of the reactor vessel was not impaired by these transients. It was further judged that the f atigue life of the vessel was not significantly aflected. An independent Florida Power and Light Co. consultant reviewed the analysts and concurred with its conclusions. The fact that there was no thermal stress present was a beneficial f actor in the analysis.

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ralue wittun sie Tectaka! Speedesion immts med staasang ao evalm W We ces itom tie % dear Steaso Systeso %qppber (Ev W d As part of stee locg servo conceesu=e m procedse hp wh e a udude monal equipmesa dancle as meti as se asimare proper v4me .hac w followeg any Sests prior to sdcasug the systems to gerasmoo These mesuus w1L smnnume sie W4y d cosqponent lassres sadv se ste coes sbat resulsed in ste 0625 sp-e W aromahet f

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i j ll.rf FEE O 21982 Florida Power and Light Company -

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Dr. R. E. Uhrig Vice President.

ATTN: c- -

-' .4 Advanced Systems and Technology i .

h,;.6 P. O. Box 529100 -

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Miami, FL 33152 .= .:;.

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.  ; . ,3'.-@t w3 Gentlemen: -

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Subject:

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% port Hos.QO-250/81-31)nd 50-251/81-31

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. . c.g This refers to the special safety inspection conducted'by Mr. 5.N Elrid of this U;M$

office on December 1-3, 1981, of activities authorized by RRC Operating 1.icense '.C.MR Hos. DPR-31 and DPR-41 foi the Turkey Point. facility. Our preliminary. findings l 4 4 N R .

were discussed with Mr. J. Hays, Phnt Janager, at the conclusion' ofithe N ',:'~ ' 'Tl.i "-^

e inspection. ,,

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Areas examined during the inspection 'and our findings are' discussed. '

.. g ia enclosed inspection report. - Within these areas, the inspection cons.isted.ofjRayc.+-s selective examinations of procedures and representative recordsf. interviews with .h ,

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personnel, and observations by .....--t .. .. ...,..,.

the inspectortp.

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g During the inspection, it was found. that certain ' activities:under.your.. license.4Rq. y appear to violate NRC requirements. . These. items'.and references :to pertinent $d;M.:7 requirements are listed in the Hotice of Violation' enclosed herewith'as?;;" ~i.V~- .g.T4 Q%.

Appendix A. Elements to be included in your response are,delinet.ted Jn ..

' '- ~~

- ' l -' ^ 4:i Appendix A. .

As detailed in the tiolice, the overpressure mitigating systesa was ' inoperable on  % .{

two occasions when called upon to perform its safety function.. We have a more

• than usual concern about this matter and shall evaluate your. response' to the Accordingly,  :

Notice .to detemine if further enforcement action is warranted.

please give this matter your particular attention.

One new unresolved item is identified in the enclosed inspection report. This item will be examined during subsequent inspections.

In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosures will be placed in the HRC's Public Document Room unless you nctify this office, by telephone, within ten days of the date of this letter and sutrnit written application to withhold information contained therein within thirty days of the date of this letter. Such application must be consistent with the requirements '

of 2.790(b)(1).

The responses directed by this letter and the enclosures are not subject to the clearance procedures of the Of fice of Nnagement and Budget as required by the Paperwork Reduction Act of 1980, PL 96-511.

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t Should you have any questions concerning this letter, we will be glad to dised them with you.

Sincerely.

S. .

..M-a. -

James P. O'Reilly

f. Regional Administrator id' '

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' 7' Enc 1osures:

Appendix A, Notice of Violation Ji4 1.

?- 2. Inspection Report Mos. 50--250/81-31.

IV.4 . and 50-251/81-31 n:-

cc w/encls:

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H. E. Yaeger, Site Manager-J. Hays, Plant Manager fh;-

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NRC ksident inspector -

Occument Management Branch

.. State of Florida j -.'

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idOTICC Of VIOLATI0ri ,

i 's,, i.:.s P w. r 5 L ight Cocipany Dacket ilus. 50-250 & 50-251

>.n. , W. int 3 & 4 License Hos. DPR-31 & DPR-4 As a r.. sal t of ti,e inspection conducted on December 1 - 3, 1981 and in accord 4nce e tn t5e :ntarin Enf orceswnt Policy, 45 FR 66754 (October .7,1980), the following ou14tiens w re iden ti fied.

Technical Specification 6.8.1 requires that written procedures be-.--- .

A.

establish.wi and i.:iplemented that meet or exceed the requirements and Section 5.3.6 of reconndations of Section 5.1 and 5.3 of ANSI N18.7-1972.

ATal : 118.7 requires ueasurements to keep sa fety parameters within v,wrational and sa fety limits..

Cientr.srj to the above, the Overpressure liitigating Systen (OMS) functional This 1.:s t was inadequate in that the summator circuitry was not tested.

resolted in failure to discover the. Orts was inoperable and contributed to the. reactor coolant system overpressure events of November 28 and 29,1981.

Inis is a Severity Level IV Violation.'(Supplement 1.0.3).

T .chnical Specifica tinn '6.8.1 requires that written procedures be

.:' tabli shed tha t :=tt or exceed the requirements and reconnendations of Mction 5.1 and 5.3 of ANSI il18.7-1972. ANSI 18.7-1972 Section 5.3.4.1 requires instructions for starting up including the requirement that valves M properly aligned.

f t.on*rary to the"abaye, a ligfrient of instrumentatioit root valv'es were'not i":1uded in station procedures prior to reactor coolant system fill a f ter ref uelin.i or plant startup.

i.ti i , i , .i Severity Level V Viola tion (Supplee.ent 1.E. ).-

Par .uant to the provisinns of 10 CFR 2.201, you are hereby required to sutnit to

• 'i'.

>f t ste within thirty days of the date of this Notice, a written state.wnt or

. e.>lanition in reply, including: (1) admission or denial of the alleged viold-

e.. 'd) the reasons f or the violations if aJiriitted; (3) the corrective steps

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teen taten artJ the re sults achieved; (4) carrective steps whicn will all i s- tan c to avnid iurtries viola tions; .ind (5) the da t' .< hen full conpliance

'-. J.9:esed. Comideratinn may be given to extendinj your response tim f or gc,rx

. .. shcue. 11nd.. r the m;tnority nf Sec t ion 1:1? of the Atmiic EnerJy Ac t ')f t  : . ,1. .c a v ;ded, this response shall I.e .utnitted under oath or af fir ation.

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NUCLEAR REGULATORY COMMISSICN

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V9 101 W A Alt T'T A IT, av W., sulT E 3100 AT LAN T A, G t om CLA Jc303 Report Nos. 50-250/81-31_ and 50-251/81-31 Licensee: Florida Power and Light Compa y 9250 West flagler Street Miami, FL 33152 Facility Hane: Turkey Point 1 and 2 Docket'Nos. 50-250 and 50-251 License Nos. DPR-31 and DPR-41 Inspection at Turkey Point site nea Ho .estead, Florida inspector: 'C 5,e [M / / 72.

S .' A. E r p, // Aate Si ned Approved by: '%  ! b H. C. Dance, Section Chief, Division of Oa te' Signed Resident and Reactor Project Inspection i SttriARY l Inspection on December 1-3, 1981 1

Areas inspected This special unannounced inspection involved 20 inspector-hours onsite in the area of exceeding technical specification pressure limits during cold shutdown conditions. -

Results Of the one area inspected, two apparent violations were found. (Inadequate l procedures for alignment of instrumentation root valves prior to filling the I

reactor coolant system - paragraph 5.f., failure to conduct adequate surveillance of the overpressure mitiga ting system - paragraph 5.f.).

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D.ETAILS Persons. Contacted

1. .

Licensee Employees H. Yeager, Site Manager-J. Hays, Nuclear Plant lianager -

C. Baker, Acting Operations Superintendent . _ .

V. Wager, Operations Supervisor J. Labarraque, Technical Staff Supervisor J. Kappas Instrumentation and Controls Supervis'or t.. Huenniger, Nuclear Plant Supervisor D. Jones, Quality Control Supervisor .

E. Hayes, Instrumentation and Control Technician

- Other licensee emplo'yees' contacted included technicians, operators, and

. .of fice personnel. .

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• Attended exit interview- * -

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The inspection scope and findings were sumarized The Nuclear enPlant December Manager 3; 198 those persons indicated in paragraph 1 above.

acknowledged the two violations discussed in paragraph 5.f. He agreed to resolve the iten discussed in paragraph 6 and to ccanunicate that resolution to the Senior Resident Inspector. He also agreed to develop a method of identifying on each record, such as clearances, that are corrected after completion because of Quality Control or other reviews.

3.. Licensee Action on Previous Inspection Findings Not inspet.:ed.

4 Unresolved items ,

Unresolved items are matters about which more infornation is required to determine whether they are acceptable or may involve violations or devia-tions. New unres31ved items identified during this inspection are discussed in paragraph 6.

5. Reactor Pressure Excursions Above Allowed Limits

a. The inspector inquired into the circumstances concerning two Unit 4 reactor system pressure excursions above technical specification limits that occurred nn November 28 and 29,1981. This inquiry included interviews, review of the fiuclear Plant %pervisor and reactor control opera tor logs, review of oper.s ting procedut t . review of interval notes hirdsmy eMursions, revie.' of certain im irJ5e"*

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2 calibrations affecting the overpressure mitigating system and review of '

.- plant drawings used by the.operations staff.

Procedures reviewed included: operating procedure .1001.1 <tated October 30, 1980 - Filling and Venting the Reactor Coolant System;-

operating procedure 0202.1 dated August 20, 1981 - Reactor Startup - -.

Cold Condition to Hot Shutdown Conditions; and operating' procedure' . .

0202.1 dated August 20, 1981 - Reactor Startup - Cold Condition to Hot? # 4#

Shutdown Conditions; and operating Procedure .1004.4 dated May 7,1981 - /

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'c Overpressure Mitigating' System Functional Test of. Nitrogen Backup '%: ..

system. - -'

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q Though some preliminary information concerning ,the. sequence of. events was available, the licensee had not yet" completed a' rigorous. internal %:' "V investigation. Some persons involved were not' available for. interview[ /.D i

during the time of this inspection.-- Though' severalifactors 'which .. #

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contributed directly to the result wers identified 7the irish ~ctor%s". ., ? 9{

unable to identify the ' initiating events with any'certaintyi .J-( .

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The plant operators were performing OP '0202.1% Reactor Startuf- Cold 's.g b.

' Condition to llot -Shutdown Conditionf.?JThe ReactodCoola'nt' System' (RCS)l.:.: .!i.

had been filled solid.. The let down ;Gth3as' viit'the7 Residual Heatl.~ . i":E .

Removal (RHR) system suction" valves N0Y4-750'ari751'Ewhich'.'close atMM ~

465 psig to protect the RHR pump' suction' linesi .The:RHR'.' system was . ~

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cross-connected to the' letdown portion' of..the;OienicaliandJolmeT "- .- $.

7 Control System (CYCS) downstream of.the'RHR heat'ex'chan90 s at' valve.: '.

' .. +* - 4 HCV-4-142. Letdown flow control to the.'volmel control.? tank $and

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consequently, RCS pressure was controlled by ~ pressure control valve PCY-4-145 in the letdown portion of the CYCS. One charging pump was in operation providing both makeup into the RCS and Reactor Coolant Pump seal in.jection flow. RCS Temperature was aboct 110'.F and pressure about 340 psig.

c. With the plant alignment described above, any flow blockage in the letdown path would cause an imediate increase in RCS pressure because Designa ted the charging pump would be charging into a solid system.

mitigating devices include an alarm at 400 psig warning of impending overpressure mitigating system (0ftS) protective action and two indeper. dent Otis channels designed to both alam and operate at 415 psig (at low temperature) and prevent an unacceptable pressure excursion,

d. The following conditions were identified which could contribute to pressure excursions.

The primary OMS valve (PORV) train was known to be inoperable, wit.. the Maintenance was being conducted on the high PORV block valve shut.

pressere controls for the PORV. Apparently unknown at the time, a blown fuse in the O'15 comparator output rendered inoperable the alarm that signals decand for prinary On5 protective action at 415 psig.

3 Unknown at the time of the first pressure excursion, the backup OMS train pressure transmitter PT-4-405, root isolation valve was found shut but was reported to " leak through". This condition would at least' preclude response -to a rapi,d pr' essure change - rendering the backup Orts also inoperable.

-No procedure was found that aligns RCS instrumentation root valves prior to RCS fill.

-The backup Of tS temperature sumator, which generates the " pressure set point", which loop pressure is compared with to generate the OMS aClua tion signal, had failed high - about 2335 psig - also rendering the backup ONS inoperable. This condition was unknown because of an inadequate surveillance procedure used to satisfy technical specification 4.26-1 and 4. The procedure is OP 1004.4 - Overpre~ssure Mitigating System Functional Test of Nitrogen backup Systesa dated -

May 7, 1981. This procedure did not test the sumator.

-When unisolated. PT-4-405 (backup OMS input) was reading about 110 psig higher than actual RCS pressure. This was determined fran past event testing. The transmitter had been relocated and the sensing ._

. line had been hydrostatically tested. This transmitter also controls ,

RHR suction line isolation valve.t0V-4-751 which isolates at 465 ,

psig, thereby isolating letdown when in this. plant aligrenent. ,

-Pressure Transmitter PT-4-401 does not actuate the OMS, but provides at 400 psig an alarm called "0MS High Pressum Alert" to wam of impending OMS protection action. PT-4-402 had not been aligned after maintenance, though.the alarm bistable setpoirt had been checked. The accuracy of this signal loop was therefore suspect. Opera tions personnel indicated that this alarm had not. functioned,.

-Operations personnel Indicated that the control action of PCV-4-145, which was controlling RCS pressure, was thought to be erratic. This could initiate a pressure excursion. The hand-auto controller was replaced af ter the pressure excursions and subsequent control action appeared smooth, however erratic behavior of the removed hand-auto controller could not be demonstrated during shop tests af ter repl a cemen t. The other components in this control system, including PCV-4-145, had not been investigated for erratic action as of the conclusion of this inspection.

e. Resultant Pressure Transients

-At 1053 p.n on November 28, tne opera tor observeo RCS ' pressure increasing above 500 psig, attempted to control pressure w.ith ,

NY 4-145, noticed that MOV-4-750 had shut due to the pressure inter-l lock '465 psig), stopped 48 RCP, de-energized pressurizer heaters,

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' 4 and opened PORV-4-455C manually. Pressure peated at 1100 psig. The overpressure condition lasted about two minutes. M0Y-4-751 did not shut apparent-ly due tp the pressure transraitter root valve being isolated.

At 12:55 a.m. on November 29 apparently after PT-4-405 was unisolated, a second excursion to 750 psig occurred. This time both N0Y-4-750 and y

10V-4-751 closed. Since PT-4-405 was indicating 110 psig higher than other detectors N0Y-4-751 shutting at about 355 psig actual RCS g

g. pressure could have been the initiating event in this case. Final '

c determination is yet to be made. . .

k E f. The inspector identM'M two apparent violations of technical l' specifications in ti..s area. .

i j -Failure to include in. procedures the alignment 'of instrumentation ,

. root valves appears to violate Technical Specification 6.8.1 which requires that written procedures be established, implemented and

rdintained that meet or eKCeed the requirements of sections 5.1 and

5.3 of ANSI N18.7-1972 and Appendix "A" of USNRC Regulatory Guide * ~

1.33, (Violation 250/.81-31-01,251/81-31-01)r ,

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-Deficient test procedure resulted in the failure to conduct an ..

i adequate functional test of the overpressure mitigating system.

Technical specification 4.16 requires such testing. The test performed did not meet the definition of functional test. The ONS summator was not included in the test. (Violation 250/81-31-02, 251/81-31-02).

! 6. Documentation of Reactor Coolant Prer.Sunt l

During post inspection review of the Safety Evaluation for operating license amendments 55 for license DPR-31 and 47 for license OPR-41, the inspector encountered a requirement f,or " pressure and temperature instruoentation to provide a permanent record of the transient" and sone requirements affecting the type of recorders to be used. During the inspection, the inspector had sought to determine if such records existed and understood that they did not. This matter is unresolved pending determination by the licensee and resident inspector whether or not such records actually exist.

(250/81-31-03, 251/81-31-03).

7. Quality Control Review of Documents Prior to Storage During the course of this inspection, the inspector observed Quality Control personnel obtainiig corrections to coupleted equipment clearance forms prior to permanent storage in the QC Records Yault. The corrections would have added stamps indicating that the equipnent was safety-related or nad independent verification of actions perfort ed, .etc. Since this correctiun t he t ine ,

would modify a record to show considerations not actually made at the inspector requested the licensee explain.

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The licensee had recently initiated this program in good faith to respond to 7 a HRC Notice of Violation concerning lack of corrective action for  :' ~

iaproperli executed clearances, etc. The , intent was to improve plant staff performance through education and feedback of eriors. The licensee stated ,

that nonconformance reports did exist documenting corrected documents, but "

.. m that a technique of also marking corrected .docu6ents themselves would be '

devel oped. This was discussed with NRC Region II management and found to be

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an acceptable course of action. This item will be followed up by the -M _

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resident inspector. (IFI 250, 251/81-31-04). :fyi I-1 e.

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. -9 n o.ls March 4,1982 L-82-80 ,

Mr. James P. O'Reilly Regional Administrator, Region 11 -

U.S. Nuclear Regulatory Commission 101 Karietta Street, Suite 3100 Atlanta, Georgie 30303

Dear Mr. O'Reilly:

Re: Turkey raint Units 3 and 4 Docket Nos. 50-250 and 50-251 IE Inspection Report 81-31 Florida Power and Light Company has reviewed the subject inspection report and a response is attached. -

There is no proprietary information in the report.

. We have reviewed the performance of the overpressure mitigating systes during the overpressure transient. We feel that the corrective actions taken provide a high degree of assurance that the overpressure mitigating system will in the future properly perform its safety function.

.Very truly yours, Robert E. Uhrig -

Vice President Advanced Systems and Technology .

REU/DWJ/mbd cc: Harold F. Reis Esquire

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-n ws= m mr-u m m ATTACHMENT RE: TURKEY POINT UNITS 3 AND 4 00CKET r 05, 50-250, 50-261 IE INSPECTION REPORT 81-31_ ,

. 1 FINDING A:

Technical" Specification 6.8.1 requires-that written procedures be established j

?

and implemented that meet or exceed the require: tents and reccrmendations of 'I Section 5.1 and 5.3 of ANSI N18.7-1972. Section 5.3.6 of AMSI M18.7 requires measurements to keep safety parameters within operational and safety limits. ,

i Contrary to the above, the Overpressure Mitigating Systes (OMS) functional' -

test was inadequate in that the suncator circuitry was not tested. Tnts- . . ,

]

resulted in failure to discover the OMS was inoperable and contributed to the 4

reactor coolant system overpressure events of Noverber,28 and 29,1981. ,

. IJ

RESPONSE

}

(A-1) FPL concurs vith the finding. ,

. .-l

( A-2) Tne Overpresure Mitigating System (OMS) functional test requires the introduction of a simulated o'v erpressure signal. . During the test,;the 1

~

stmnator was bypassed because the summator is ~used to provide ~ a . .g constant D.C. signal as a reference or setpoint for. the OMS:

_ ~ "r circuitry. In this operating region the stmator output' does not.' 5

,p change 'with the input Reactor Coolant systan tempe,rature change. ,

Because of this it was determined that the oest testing method would be to introduce the test signal at the summator output. V- .;

(A-3) As corrective action and in order to prevent recurrence, we have (A-4) evaluated the OMS performance during the overpressure incident and cade .

the follo,<ing revisions:

a) Operating Procedure 1001.1, Filling and Venting the Reactor Coolant System, has been changed to include verification that instrtnent block valves are correctly aligned. The procedu're has been updated to include testing of OMS at two different steps in the procedure, and addition of transmitter and sumator checks to tne tests.

l' b) Operating Procedure 1004.4, Overpressure Hitigating Systen Functional Test of Nitrogen Back-up Systen has been changed .to include checks on applicable pressure transmitters, suuator output, and recording of actual test data.

.2 Cold c) Operating Procedure 0205.2, Reactor Shutdown, Hot Shutdown Shutdusn Conditions, has also been revised to include additional .

checks on OM5 sunnators.

(A-5) Full cunpliant.e was achieved un March 1,1982. ,

M

I ! *.:ll .:. t I.f r.nn te.al ',peci f ica tion 6.8.1 requi res that written procedures be establisned th.st v..:t or e4r.eed the requireinents and reconmendations of Section 5.1 and,

. r,. 3 <>f A!is! N18. 7-19 /2; AN5I N18.L-1,972 Section 5.3.4.1 requires instruction; for storting up incl.; ding the requirement that valves ce properly aligned.

Cui.trary tu tne above, alignment of -instrumentation root valves were not incl ,ded in station procedures ' prior to reactor coolant system fill af ter r-ef uelin.j or plant startup.

HLSP03SE:

(B-1) FPL concurs with the finding. - -

(B-2) in.rse valves were inadvertently omitted fron the plant startup -

procedures.

(B-3) As corrective action and in order to prevent recurrence, Operating d n.1 Procedure 0202.1, Reactor Startup, Cold Conditions to Hot Shutdown (B-4) Conditions, will be changed to include root valve alignment checks on inst ru:aents af fecting alann f unctions, automatic action, and transient contro!.

(11-5) full conpliance will be by fspril 15, 1982.

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STATE OF FLORICA ) '

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Cct:: Y cr crac 7 .

J. W. Williams, Jr. , being first duly sworn, deposes and says:

That he is Vice President of Florida Power T.

Light Company, the . herein; 4

• That he has executed the foregoing document; that the stith-ments .nade in this said document are true and correct to the best of his 'r.ncvledge, information, and belief, and that he is authorized to execute the document on behalf of said

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J. W. Williams, Jr.

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Subscribed and sworn to before me this Y day of , 19 b d-

';oT.5Y P6aLIC() in and for the County of Dade, State of Florida My commission expires: '

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florida PMr and Light Company ATTN: Dr. R. E. Uhrig, Vice President Advanced System and Technology -

P. O. Box 529100 Mi a ai , FL 33152 Gentlecen:

Subject:

Report Nos.QO-250/81-31 and 50-251/81-31

~

Thank you for your letter of March 4,1982, informing us of steps you have taken to correct the violations conceming activities under MRC Operating License Nos. DPR-31 and DPR-41 brought to your attention in our letter of February 2,1982. We will examine your corrective actions and plans during subsequent inspections.

We apprettate your cooperation with us.

Sincerely.

<*9 R C. Lewis. Director Division of Project and Resident Programs cc: H. E. Yaeger, Site Manager t,t c : NRC Resident inspector Document Management Branch _,

N State of FL g ! 3 * ' ' ' s 4 N. ' '

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