05000267/LER-1980-030, Revised LER 80-030/03X-3:on 800604,during Normal Operation, Loop 2 Steam Generator Penetration Leakage Greater than Tech Spec Limits.Caused by Inadequate Tech Specs.Tech Specs Modified

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Revised LER 80-030/03X-3:on 800604,during Normal Operation, Loop 2 Steam Generator Penetration Leakage Greater than Tech Spec Limits.Caused by Inadequate Tech Specs.Tech Specs Modified
ML20087P543
Person / Time
Site: Fort Saint Vrain Xcel Energy icon.png
Issue date: 03/28/1984
From: Don Frye, Mcbride L, Novachek F
PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20087P537 List:
References
LER-80-030-03X, LER-80-30-3X, NUDOCS 8404090146
Download: ML20087P543 (7)
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LER-2080-030, Revised LER 80-030/03X-3:on 800604,during Normal Operation, Loop 2 Steam Generator Penetration Leakage Greater than Tech Spec Limits.Caused by Inadequate Tech Specs.Tech Specs Modified
Event date:
Report date:
2672080030R00 - NRC Website
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10121 I On June 4.1980. durina nonnal oower operation while nerforming SR 5.2.16a-m. Loop 2 I steam generator penetration leakage appeared to be greater than 400 pounds per day. l o 3

. o A i Further testing revealed' the leakage was internal to thb steam generator and not I o s I through the seals. Leakage was greater than 700 pounds per day on several occasions I oe I during February and March,1981, and resulted in a plant shutdown on March 22, 1981, I

) o 7 l for further investigation. These events were reported as operation in a degraded I 1o#a 1 mode of LCO 4.2.9 per Fort St. Vrain Technical Specification AC 7.5.2(b)2. No affectc0I 7 seonpuDl1Cggthorspety. g similar reportable oCCurrenceg. sueCODE sueCODE CODE CODE sV8 CODE COMPONENT CODE o e I HI RI@ [2,j@ [,Z j @ l HI Tl El XI Cl H 18l@ L Fj@ [Z,,J 19 20 9 10 ft 12 13

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- CAUSE DESCRIPTION AND CORRECTIVE ACTIONS 27 l 'jE I The ourified helium leakaoe is internal to the nei. ation intersnace and -occurred I Public Service l l N I between the interspace and associated cold reheat steam pipina.

l l3 1 Company Charige Notice (CN) 1436 modified the steam generator interspace to operate at I Revisions to LCO 4.2.7 and 4.2.9 i

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!. , i 4 I primary coolant activity. No further corrective action is anticipated or required.

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/( - o REPORT DATE: March 28, 1984 REPORTABLE OCCURRENCE 80-30 ISSUE 3 OCCURRENCE DATE: June 4, 1980 Page 1 of 6 FORT ST. VRAIN NUCLEAR GENERATING STATION PUBLIC SERVICE COMPANY OF COLORADO 16805 WELD COUNTY ROAD 19 1/2 PLATTEVILLE, COLORADO 80651-9298 l REPORT NO. 50-267/80-30/03-X-3 l Revised Final i

- IDENTIFICATION OF OCCURRENCE:

During performance of plant test T-147, the steam generator penetration interspace leakage appeared to be in excess of the limit allowed by LCO 4.2.9 and the variance granted by the Nuclear Regulatory Commission on June 5,1980.

This was reported as operation in a degraded mode of LCO 4.2.9 per Fort St. Vrain Technical Specification AC 7.5.2(b)2.

EVENT DESCRIPTION:

On May 28, 1980, during performance of the scheduled surveillance on PCRV leakage, Loop 2 steam generator penetration was found to have a 1eak rate of 8.23 pounds per hour or 198 pounds per day. This was well within the 400 pounds per day limit of LCO 4.2.9.

On June 4, 1980, the surveillance was again performed and indicated steam generator penetration leakage to be 17.83 pounds per hour.

This was 428 pounds per day and in excess of the LCO 4.2.9 limit of 400 pounds per day. A special. test, T-145, using a pressure decay method, was written, approved, and performed later that day and l . showed the leakage rate to be 10.74 pounds per hour or 258 pounds per

day, within the limits of LCO 4.2.9. T-145 was a more accurate test and was used at this time to insure Technical Specification compliance. .

During performance ~of T-145, it was noted that the pressure decay j, rate did not decrease as reactor pressure was approached. This E indicated that a primary closure seal was not leaking. Further

, investigation revealed a noncondensible gas was in the hot reheat sample line. A sample was taken, analyzed, and determined to be

[. clean helium. The leakage flow path was believed to be from the

- penetration interspace to the cold reheat piping and out of the steam l generator via the hot reheat piping (see Figure 1).

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REPORTABLE OCCURRENCE 80-30 ISSUE 3 Page 2 of 6 On June 5, 1980, T-145 was again performed and revealed a leakage rate of 10.37 pounds per hour or 249 pounds per day. A letter was sent to the Nuclear Regulatory Commission explaining the situation and requesting temporary relief from LCO 4.2.9, permitting up to 700 pounds per day instead of 400 pounds per day for the suspected internal leakage path.

The Nuclear Regulatory Commission granted the requested temporary relief, agreed to the four administrative controls proposed by Public Service Company, and added one additional administrative control.

The five administrative controls agreed to are as follows:

1. SR 5.3.7, Secondary Coolant Activity, be conducted once each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in lieu of once per week.
2. A pressure decay test be conducted for the Loop 2 steam generator penetration closures on a weekly basis. This pressure decay test was utiTized to determine the leakage rate. Leakage rate increases of 25% over previous values required conducting pressure decay tests daily until it was established that the leak i rate had reached an equilibrium value within the 700 pounds per day.
3. SR 5.2.16, PCRV Closure Leakage, be conducted once every two weeks for Loop 2 steam generator penetration closure rather than monthly as a comparison to the pressure decay tests.

' 4. Radiation process monitors for the reheat steam system will be monitored once per shift for indication of primary coolant leakage into the secondary system.

! 5. Check and record the interspace differential pressure once per

, shift to comply with LCO 4.2.7.

To collect all the required data, T-145 was expanded and renumbered

. T-147. Both T-147 and SR 5.2.16 were performed during each power i increase, as required, and also as the leakage appeared to be somewhat dependent on power level.

The leakage was within the 700 pounds per day limit, except for three occasions during power level increases:

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1. On February 23, 1981, SR 5.2.16a-M was performed and indicated

!~ S02.4 pounds per day leakage. Test T-147 was performed as part of this surveillance and indicated the leakage to be 680.4 pounds per day. Later the same day, T-147 was again performed and indicated the leakage to be 777.6 pounds per day.

l l 2. ' On' February 25, 1981, T-147 was performed and indicated leakage of 709.7 pounds per day.

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3. On March 4, 1981, T-147 was performed and indicated leakage of 709.7 pounds per day.

Plant operation was changed in each of these three situations and the leakage rate was reduced to less than the 700 pound per day limit.

On March 21, 1981, at 0227 hours0.00263 days <br />0.0631 hours <br />3.753307e-4 weeks <br />8.63735e-5 months <br />, during plant testing, a turbine trip occurred. Prior to this trip, reactor power had been 69%, and the Loop 2 steam generator interspace leakage rate had been 568 pounds per day. The plant was restored to a stable condition, power increased, and the generator placed back on line at 0702 hours0.00813 days <br />0.195 hours <br />0.00116 weeks <br />2.67111e-4 months <br /> the same day. Reactor power was increased, and normal plant operation resumed during the remainder of the day shift. At 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />, T-147 was performed and indicated the steam generator interspace leakage to be 1312 pounds per day at 57% reactor power.

Various changes were made to plant operations, but the steam generator interspace leakage rate remained above the 700 pounds per day limit. An orderly plant shutdown was initiated at 1832 hours0.0212 days <br />0.509 hours <br />0.00303 weeks <br />6.97076e-4 months <br /> with the turbine generator off line at 0358 hours0.00414 days <br />0.0994 hours <br />5.919312e-4 weeks <br />1.36219e-4 months <br /> on March 22, 1981, and the reactor manually scrammed from approximately 2% power at 1145 hours0.0133 days <br />0.318 hours <br />0.00189 weeks <br />4.356725e-4 months <br />. The reactor was depressurized and investigation began on the cause of the excessive steam generator Loop 2 interspace leakage.

CAUSE DESCRIPTION:

Other. ,

The purified helium leakage was internal to the Loop 2 steam generator penetration and occurred between the penetration interspace and the cold reheat steam piping internal to the penetration (see Figure 1).

l CORRECTIVE l ACTION:

l Public Service Company Change Notice (CN) 1436'was initiated in l November, 1981, and has installed instrumentation, piping, valves, and control equipment to allow operation of the steam generator

! . penetration interspaces at a pressure slightly greater than cold reheat but less than reactor pressure. Additional capability to r- monitor the interspace helium was also provided. These modifications maintain steam generator interspace helium leakage to within the limits of LCO 4.2.9.

Revisions to the Fort St. Vrain Technical Specifications, LCO 4.2.7

_and LCO 4.2.9, were proposed by Public Service Company in

! January, 1982, and approved by the Nuclear Regulatory Commission in March, 1982, as Amendment No. 26 to the Fort St. Vrain Facility

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REPORTABLE OCCURRENCE 80-30'~

ISSUE 3 Page 4 of 6 Amendment No. 26 revised the Technical Specifications to:

(1) permit the interspace between primary and secondary closures of the steam generator modules to be maintained at a pressure slightly above cold reheat steam pressure; and (2) set a limit on the possible release of primary coolant activity through the primary closure seals of no greater than 1.4 curies

.per day.

No further corrective action is anticipated or required.

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REPORTABLE OCCURRENCE 80-30 ISSUE 3 Page 6 of 6 Prepared By: aarA Duane L. Frye

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