ML20004D456
| ML20004D456 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/05/1981 |
| From: | Green H TENNESSEE VALLEY AUTHORITY |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| RO-81-2, NUDOCS 8106090419 | |
| Download: ML20004D456 (3) | |
Text
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TENNESSEE VALLEY AUTHORITY e H A T T * *:OvuA. TENNESSEE 374ot 1750 Chestnut Street Tower II June 5, 1981
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fU Mr. James P. O'Reilly, Director R
ll-4 U.S. Nuclear Regulatory Commission g
1 Cffice of Inspection and Enforcement AN 0 81981m 7
Region II 3
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101 Marietta Street, Suite 3100 s
Atlanta, Georgia 30303
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Dear Mr. O'Reilly:
TENNESSEE VALLEY AUTHORITY - SEQUOYAH NUCLEAR PLANT UNIT 1 - DOC NO. 50-327 - FACILITY OPERATING LICENSE DPR SPECIAL REPORT 81-2 The enclosed special report provides information concerning two emergency core cooling system injections to the reactor coolant system.
This report is submitted in accordance with Sequoyah unit 1 Technical Specifications 6.9.2 and 3.5.3.
Very truly yours, TENNESSEE VALLEY AUTHORITY H. J. Green Director of Nuclear Power 4
Enclosure (3) cc (Enclosure):
Director (3)
Office of Management Information and Program Control U.S. Nuclear Regulatory Coraission
[
Wash.
, DC 20555 Director (40)
Office of Inspection and Enforcement U.S. Nuclear Regulatory Cor: mission
- Washincron. DC
'0555 Mr. Bill Lavallee l
Nuclear Safety Analysis Center I
Palo Alto, California 94303 l
NRC Inspector, Sequoyah 9
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8106000 M o
An Equal Opportunity smoloyer V
4 Special Report 81-2 Sequoyah Nuclear Plant Unit 1 ECCS Injections to the Reactor Coolant System This report provides information concerning two emergency core cooling system injections to the reactor coolant system. The first inadvertent injection occurred on April 23, 1981, and the second occurred on itay 1,1981.
A.
April 23. 1981 Safety Iniection Plant Status Mode 3 RCS Pressure 2234 psig RCS Temperature 539.80F BIT Temperature 170 F Event Description and Probable Consequences On April 23, 1981, during performance of a main turbine overspeed test, an inadvertent safety injection occurred. The signal which brought in the safety in.iaction was from a combination of high steamline flow coin-cident with low-low Tave. This safety injection caused the centrifugal charging pumps to pump water through the boron injection tank (1700F) into the Reactor, Coolant, System (539.8 F) for about 2-3 minutes.
Cause Description and Corrective Actions The automatic initiation occurred when the average RCS temperature fell below chu trip setpoint of 5400F coincident with a signal of high steam-line flow. The safety injection was terminated about 2-3 minutes after the SI signal was actuated, and the reactor coolant system was restored to preinitiation conditions.
B.
May 1, 1981 Safety Injection Plant Status Mode 2 RCS Pressure 2220 psig RCS Temperature 500 F BIT Temperature 170 F Event Description and Probable Consequences On May 1, 1981, during recovery from a generator trip test, an inadvertent safety injection occurred. The safety injection was brought in by the reception of two high steamline differential pressure signals. This safety injection caused the centrifugal charging pumps to pump water through the boron injection tank (170 F) into the reactor coolant system
-(500 F) and eventually brought in water from the refueling water storage tank.
A
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B.
May 1, 1981 Safety Injection, Continued Cause Description and Corrective Actions The automatic initiation occurred when two reactor coolant pumps were returned to service after the generator trip test was complete. Appar-ently, the two RCP's were started almost simultaneously causing hot water to be pumped into two of the four steam generators. This made the pressure inside the two steam generators rise rapidly and, as a result, two high steamline differential pressure signals were received which caused automatic initiation of the safety injection. The injection was terminated shortly after the SI signal was actuated, and the reactor coolant system was restored to preinitiation conditions.
l C.
Fatigue Usage Factor Sequoyah Technical Specification LCO 3.5.2 requires that the current vr.lue of the fatigue usage factor of safety injection nozz.'.es be included in the,5pecial, Report if the valve exceeds 0.7.
'Isage factors are not available for these nozzles because the primary piving was designed to the USAS B.31.1 Power Piping Code, which does not require fatigue analysis. However, IE Circular 78-05 states that Westinghouse Electric Corporation informed the Public Service Electric and Gas Company that 50 safety injections using 40 F water would be acceptable for the Salem I Plant which has a safety injection nozzle design similar to Sequoyah.
It i&'our opinion, based on IE Circular 78-05, conversations
~
with Westinghouse, and our.own evaluation that the fatigue usage factor 3
for the affected nozzles will not exceed 0.7 if the ECCS occurrences are less than 35 cycles (0.7 x 50). These two safety injections are only the third and fourth such events for Sequoyah,#' nit 1.
e 5
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