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Tt:NNESSEE VALLEY AUTF.ORITY CHATTANOOGA, TENNESSEE 37401 4 August 29, 1973 gygygggg PAATNERSHip Mr. F. E. Kruesi, Director Directorate of Re atory Operations U.S. Atomic Ener Comission Washingtdn, DC 205145
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.
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Very truly yours, t
J. E. Gilleland Assistant to the Manager of Power Enclosure CC (Enclosure):
Mr. Norman C. Moseley, Director Directorate of Regulatory Operations U.S. Atomic Energy Comission
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Region II - Suite 818
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230 Peachtree Street, MI.
Atlanta, Georgia 3030'4
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..e E::CLOSUi1E Browns Pcrry I:uelcar_P] ant DESIG:: DENCIE::CY REPO?.T - TARGET RCCX SAFETY R_E.T.TEF VATYE__S__
GE conducted tests at Moss Ianding to establish the correlttien fer safety valve tire of cperation between using gasccus nitrogtn and stean.
Qirir.g the stean tests, GE found that the safety-re2ief valves ( ranufactured by Target Rock Ccepany) failed to teet the tin.ing recuirc ents stated in the Browns Ferry FSAR.
The tiring requirc cnts are given in sectica L.4.5 of the TSAR, which states "The delay tire (es:i: n elspacd tire between overpressitre signal and actual valve notic:1) and the response tine (raximun
~ valve stroke tire) tre cach equal to or 1 css than 0.2 second." Centrary
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to what is stated in the FSAR, the respcnse time should be 0.3 seccnd.
The correc; tires of 0.2 second for delay and 0.3 sccend for valve strche
'wcre used in the overprotection rcycrt subnitted in response to Q estion 4.1 dated Decenber 6, 1971.
(' Die repcrt also incorrectly states the valvc stroke time is 0.2 second.) However, the data fren the M ss Landing stcan tests had reasured delay times up to 0.8 second and rcspense tire vp to 0.1 seccad and a total clapsed tinc of up to 0 9 second.
In accordance with the rules of 10CIR50 55(e), this is a design deficiency.
An investl.gation was initiated to deterdne why the delay tires and respcuse times were both excessive and erratic.
Tue,investigr. tion indicated that condensation was cc11ceting behind the nain valve pisten.
Tnis resulted
- in addition.1 friction frcn both viscous drag and inertial effects.
Since the additional fr'ictica was dependent on the arount of condencate collected, both the excessive tices and erratic behavior were attributable to the condensation.
A modificatica to the design of the valve was made as shown on the attached sketch.
Inis codification per=its the cendensate to return to the inlet of the valve. All eleven valves for Browns Ferry Unit 1 have been so codified and successfully tested to rect a delay time of 0.4 second and a response time of 0.1 second.
(Justification for the ncv tices is discussed below.)
'lhey will be installed before Unit 1 exceeds 1 percent cf rated power.
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i If the design deficiency had not been detected, the effect of the increased total elapsed timed (frca 0.5 second to 0 9 secend) for self-actuation (overpressure safety mode) of the power relief /cafety valves would have affected the transients in the FSAR as fo110ws:
1.
Turbine Trip at resirn Fever a.
Without turbine steam breass--the peak pressure at the valves vould have increased frca 1168 psig to 1198 psig--reducing the margin between the peak pressure and the setpoint (1230 psig) of the two safety valves.
However, no steam would have been released to the drywell.
The peak nucicar system pressure wculd 1
have been 12!0 psig at the bottom of the vessel, which is well 4
below the peak vessel limit of 1375 psig.
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With tu'rbine steam bvrass--the peak pressure at the valves would have increased fren 1113 psis to 11hl pais--a change of 28 psis.
This would net have rneased any ster.: to the drpell through the two safety velves.
The peak nuclear systc pressure would have been 1189 psig, which is well below the vessel limit of 1375 psig.
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2.
IS.in Sten-Line Isclatien 'lalve Closure The safety valves are sized using this transient to assure that the vessel pressure 14-it of 1375 psig is not exceeded.
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Pressure scra.--the peak nuclear systen pressure at the bottes of the vessel would 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 bottom of the vessel that approached but would not have exceeded the 1375-psig limit, but the targin would have been stall.
The original analysis had a cargin of at least '72 psi for the equilibrium fuel
- cycle, i
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Flux scrad--the peak nuclear system pressures d the bottom of the i
vessel are always less for this transient than for pressure scrata.
The same general trend would have existed as the pressure scram case discussed above, but = ore cargin would have existed.
Although the targins for the vessel pressures would have been decreased, the vessel would not have exceeded the 1375-psig li=it for the transients.
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Also, the reactor dore transients associated with these abnormal cperational occurrences would have increased in severity by le.:s than 1 percent for MCliFR, fuel centerline temperature, and surface heat flux.
Since there would still be nargin for these core transients without fuel da.nge or fission '
product releases, the health and safety of public would not have been endangered if the design deficiency had not been detected.
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'Ihe design rodification was evaluated vit sensitivity studies for a 0.5-secend total clapsed tire for valve actuation in the overpressure mode (0.11 seccad for delay time plus 0.1 seccad for res,bonse ti e).
In addition, the overpressure protection provided by the eleven. power relief / safety valves was changed to:
Capacity at 103 Fercent Entber of Valves Set Pressu-o (PSIG) of Set Pressure (PFH) 4 1080 800,000 4
logo 808,000 3
1100 815,000 and thece chen;.cs.:ere incimied as part of the stu.?ics.
Tne results of these sensitivity studies gave only a h-psi not increase in the turbine trip without steam bypass case and caly a 1-psi net increase in ':e case of a rain steem isolaticn valve closure accupied with a reactor scram from.
high pressure.
The overall result of the valve performance changes on vessel pressure is, therefore, concluded to be small when compared to the vessel overpressure rargins contained in the FSAR.
The MCEF2, fuel center-line te=perature, snd surface heat flux vill also re= sin substantially unchanged; the change is less than 1 percent toward a core severe condition.
Since there was considerable targin for these core transients before, the results of these transients. after the modifications indicate essentially the same conclusions that are presently stated in the FSAR.
'lhe change in valve, capacity rating was made in order to be consistent with the ASME Boiler and Prccsure Vessel Code.
These cr.pacities reflect the Code method of rating these valves based on the actual valve orifice ama.
These new capacities were used to update the calculational procedures to correspond to as-purchased valve ratings rather than minir.t specified capacities.
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Since analyses have shown that these r.edifications result in essentially the same margins in the safety considerations as the original design, the l
codifications vill provide the sar.e degree of assurance as before that j
the health and safety of the public will be protected.
'Iherefore, the power relief / safety valves for Units 2 and 3 vill be modified accordingly 8
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and the FSAR will be amended to reflect these nadifications to the valv and valve data.
Also, the results of these studies vill be incorporated in th'c appropriate figures and discussion raterial of the FSAR and submitted during October 1973.
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