ML20085H056

Final Deficiency Rept Re Design of Safety Relief Valves. Caused by Condensation Collecting Behind Main Valve Piston. Design Modified to Move Condensate to Valve Inlet & New Delay & Response Times Justified

ML20085H056
Person / Time
Site:	Browns Ferry
Issue date:	08/29/1973
From:	Gilleland J TENNESSEE VALLEY AUTHORITY
To:	Kruesi F US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML20085H038	List: ML20085H056
References
10CFR-050.55E, 10CFR-50.55E, NUDOCS 8308300121
Download: ML20085H056 (6)

Text

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TENNESSEE VALLEY AUTHOAITY CHATTANOOGA, TENNESSEE 37401 August 29, 1973

$YN55?$%

PA ATN E A S HIP

• St Mr. F. E. Kruesi, Director Directorate of Regulatory Operations U.S. Atomic Energy Comission Washington, DC 20545

Dear Mr. Kruesi:

TVA made an initial report to the AEC-DRO Region II office by telephone regarding deficiencies in the Target Rock Safety Relief Valves at the Browns Ferry Nuclear Plant. In accordance with paragraph 50.55(e) of 10 CFR 50, we submit the enclosed final report of the deficiencies.

Very truly yours, I f ,/

D As.sistant E. Gilleland to the Manager of Power Enclosure CC (Enclosure):

Mr. Norman C. Moseley, Director Directorate of Regulatory Operations U.S. Atomic Energy Comission Region II - Suite 818 230 Peachtree Street, NW.

Atlanta, Georgia 30303 8308300121 750401 PDR ADOCK 05000 S

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1}EPICI!:::CY E!1TT - TARGI:T RnCK %'.FP.~T -O:M EP \'AIXES GS conducted tents at Posn Landing to establich the corre]ation for f safety valve time of cperation betwcen using cacco trcada tnd : team. ,P,drir.:;

the steam tests, GE found that the cafety-relief valves ( tanufactured by i

f Tarr,et Rock Ocrpany) failed to meet the tiring recuireacnta stated in the

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Browns Ferry FP*R. Tne tining requirements are given in section h'.h.5 l

of the FSAR, which states "Tne delay tire (naxi: ara elapsed tire beiween J overpresc'ure signal and actual valve notien) and the recponce tire (n2ximun j  ?

t valve ntroke tire) are ench equal to or less than 0.2 cecond." fcntrary 1

to what is stated in the FSAR, the retponse tire should be 0 3 cec nd. Tne

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correct times of 0.2 cecond for delay and 0 3 cecend for valve siroke -

were used in the overprotection reprt subnitted in rehponse to Q.wstien h.1 dated Dccember 6, 1971. (51c report also incorrectly states the valve ctroke tire is 0.2 second.) However, the data from the Moss Landing stean tects had teasured deley tinen up to 0.8 second and response time up to 0.1 coccnd and a total elapsed tine of up to 0 9 cecond. In accordance with

.- the rulcc cf 30CFR50 55(c), this is a design deficiency.

An investigation was initiated to determine why the delay tines and response times were both execccive and erratic. Tne investigation indicated that condensation was collecting behind the rain valve picton. Tnis reculted in additional friction from both viccouc drag cad inertial effects. Since the additicnc.1 friction was dependent on the a cunt of condencate collected, both the exceccive tirec and erratic behavior were attributable to the condencation. .

A modificatica to the design of the valve van rade as shown on the attached sketch.

This rodificatica permits the condencs.to to return to the inlet of

, the valve. All eleven valves for Browns Ferry Unit 1 have been so codified and cucceccrully. tented to neet a delay time of 0.4 second and a response time of 0.1 second. '

(Juctification for the new times is discucced below.)

'lhey will be installed before Unit 1 execeds 1 percent of rated power. -

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If the design deficiency had not been detected, the effect of the increased total elapsed timed (from 0.5 second to 0 9 sc a nd) for scif-actuation (overpressure safety mode) of the power relief / safety valves veuld have I

affected' the transients in the FSAR as follows: .

1. Turbine Trip at Desien Power

+ a. Without turbine stcan bvoass--the peak pressure at the valves vould have increased from 1168 psig to 1198 psig--reducing the

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margin between the peak pressure and the sctpoint (1230 psis)

, of the two safety valves. HovcVer, no steam would have been released to the dryvell. The peak nuclear system pressure would have been 2240 psig at th'e bottom of the vessel, which is well below the peak vessel limit of 1375 psic.

.b. With tu'rbine steam bycass--the peak pressure at the valves would have increased frcm 1113 psis to 1141 psic--a change of 28 psis.

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Bis vould not have released any steam to the dryvell through the two safety valves. The peak nuclcar syste= pressure vould have

> been 1189 psig, which is well belou the vessel limit of 3375 psig.

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2. Main Sten 1 Line Isolation Valve Closure' s

, The safety valves are sized using this transient to assure that-tbc E vessel pressure W it of 1375 psis is not exceeded. -

it. . pressure scram--the peak nuclear system pressure at the bottour of the vessel vould have been about 1300 psig in the first fuel cycle.

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Subsequent fuel cycles would have had higher peak pressures at'the '

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bottom of the vessel tlynt approached but would not have exceeded the 1373 i psig limit, but the cargin votild.have been small. The original

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j, analysis had a cargin of at least 72 psi for the equilibrium fuel

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,/ b. Flux scred--the peak nuclear system pressures rt the bottom of. the -

- vessel are always less for this transient than for pressure scrams.

h e same general trend would have existed as the pressure scram case discussed above, but = ore margin vould have existed. -

3 . Altliough the margins for the vessel pressures would have been decreased,

-the vessel vould not have execeded the 1375-psis li=it for the transients.

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a Also, the reactor core transients ancociated with these abnoir.al operational occurrences would have increased in severity by lens than 1 percent for ICfFR, fuel centerline te::perature, and surface hr .t flux. Since there would

, still be margin for these core transients without fuel damage or fission i

product relcaces, the health and safety of public vould not have been .

, endangered if the design deficiency had not been detected. "

Mic design modification was evaluated via acnsitivity studien for a

, 0.5-necond total clepsed time for valve actuation in the ovelpresnurc tnde (0.4 second for delay time plus 0.1 second for reshonse time). In addition,

, the overpressure protection provided by the cleven. power relier/cafety valves was changed to: "

. Capacity at 103 percent Number of Valvec Set Prc sure (pSIG) of Set Prescure (P;d ._

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1o80 800,000 4 1090 . 808,000 p 3 '

1100 815,000

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cnd these changen were included as part of the studies. Toe results of -

, these sensitivity studies gave onAy a h-pci not increase in the turbine

,h trip without ctcan bypass case and only a 1-pai net increase in the case ,

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>s of a main steam isolation valve closure scoupled with a reactor scram from h high pressure. The overall result of the valvo perfor:rance changes on - -

h vessel pressure is, therefore, concluded to be small when- coupared to the

g vessel overpressure margins contained in the PJAR. The MCh?R, fuel center-y liric temperature, and surface heat flux will also remain substantially ji unchanged; the change is less than 1 percent toward a more severe condition.

Since there was considerable margin for these. core transients before, the 4 results of these transients. after the modifications indicate eccentially the samt conclusions that azu presently stated in the FSAR. *

'Ihc change in valve, capacity rating was made in order to be consistent with the ASIE Boiler and Pressure Vessel Code. Diese capacitics reficct the Code method of rating these valves based on the actual valve orifice a:va.

Shose new capac'itics were_used to update the, calculational procedures to correspond to as--purchased valve ratings rather than miaimum specified capacitics.

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Since analyses have shown that these rodificaticas result in essentially the stune raarcins in the safety considerations as the original desita, the Inodifications will provide the saae degree of assurcnce as before that the henith and safety of the public will be protected. 'Dierefore, the pouerrelier/cafetyvalvesforUnits2cnd3villbenodifiedaccordiucly and ihe FSAR will be amended to reflect thenc I:odifications to the valve and valve data. Also, the results of these studies will be incorporated in tlic appropriate figures and discuccion material of the FSAR and submitted iloiing October ic//3 ,

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