ML19257B496
| ML19257B496 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 12/31/1979 |
| From: | Harold Denton NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | Ahearne J, Gilinsky V, Kennedy R NRC COMMISSION (OCM) |
| Shared Package | |
| ML19257B493 | List: |
| References | |
| REF-10CFR9.7 NUDOCS 8001160428 | |
| Download: ML19257B496 (2) | |
Text
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UNITED STATES g,,
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NUCLEAR REGULATORY COMMISSION r,,
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%'% g j,f December 31, 1979 MEMORANDUM FOR:
Chairman Ahearne Commissioner Gilinsky Commissioner Kennedy Commissioner Hendrie Commissioner Bradford THRU:
Executive Director for Operations MN $v-b N.-
FROM:
Harold R. Denton, Director Office of Nuclear Reactor' Regulation
SUBJECT:
POINT BEACH UNIT 1 - STEAM GENERATOR TUBE DEGRADATION In the Confirmatory Order for Modification of License dated November 30, 1979 (Order), certain requirements were made pertaining to the operation of Point Beach Unit 1.
In the Safety Evaluation appended to that Order, certain remedial actions were discussed. Among these remedial actions we noted that the licensee planned to operate the facility at a reactor coolant pressure of 2000 psia rather than 2250 psia to reduce the internal pressure stresses by about 157, during operation (Action No. 3 - pp.15).
This action was to be initiated upon NRC approval of an amendment request dated November 2,1979, which requested permission to operate at that pressure.
In the same Safety Evaluation we discussed " Measures for Reducing the Rate of Degradation" on pp. 22 and 23. We indicated that the acceptability of this proposed operation would be addressed separately, The Order of November 30, 1979, was based on information resulting from the steam generator tube inspection following the October 1979 leak.
On December 11, 1979, another steam generator leak occurred.
An eddy current test was performed on both steam generators which resulted in eddy current indications below the tube sheet (in the tube crevice) in both steam generators.
Twenty tubes were plug'ged in steam generator A and fifteen tubes were plugged in steam generator B.
Since there appears to be evidence of continuing intergranular corrosion attack, the NRC staff has now found that it is not only desirable, but prudent and necessary, to take immediate action to require the reactor coolant pressure to be reduced from 2250 psia to 2000 psia since this will have the effect of substantially reducing the differential pressure across all tubes in both steam generators.
As explained below, operation of Unit 1 at a reactor coolant pressure of 2000 psia is acceptable from an accident analysis point of view.
The applicable criteria for transient and accident are still satisfied.
Contact:
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Chm, Cmrs, OPE, OGC, Magee, Chilk, Records k
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tie Comission December 31, 1979 Since the licensee's amendment request was for operation at either 2250 psia or 2000 psia, he has recently withdrawn his amendment request and has made a cor.raitment to operate Unit 1 at a reactor coolant system pressure of 2000 psia.
_Sumary Evaluation Operating the facility at a reactor coolant pressure of 2000 psia rather than 2250 psia will reduce the internal pressure stresses during operation by about 15%.
The systems performance aspects of this change was evaluated by the licensee.
Two trip' setting changes are required. These relate to DNBR protection and low "
reactor pressure protection.
Several accidents were reanalyzed using standard Westinghouse methods.
These included the LOCA and the Rod Ejection Accident.
We have concluded that the appropriate reactor protection criteria continue to be met.
.D_egraded Tubes Above the Tube sheet Since our November 28, 1979 Commission meeting, the staff has learned of five degraded tubes, located at or above the surface of the tube sheet, that were detected and plugged during the October-November plant outage. The licensee have examined this infomation and detemined that it is not associated with intergranular corrosion attack, the phemonenon occurring in the tube sheet crevice. The staff discussed this experience with the licensee in late December, and concurred that this degradation was associated with the thinning, or wastage phenomenon occurring during the 1972-1975 time frame.
Conclusion The staff proposes to issue a Confirmatory Order, following the Comission briefing new scheduled for January 2,1980. g
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Y(w Harold R. Denton, (bre([torZd.
d-c Office of Nuclear Reactor Regulation ec:
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1753 135
OUTLINE i
POINT BEACH UNIT 1 - STEAM GENERATOR STAFF LiSCOVERY OF DEFECTS AT AND AB0VE TOP OF TUBE SHEET -
IMPACT ON SAFETY EVALUATION DATED NOV. 30, 1979 A.
DISCUSSI0!! 0F 5 DEFECTS AB0VE TUBE SHEET WHEN DEFECTS WERE DISCOVERED AtlD UNDER WHAT CIRCUMSTANCES SAFETY SIGNIFICANCE STAFF'S EVALUATION / ACTIONS B.
SUPPLEMENTAL CONFIRMATORY ORDER REDUCE PRIMARY SYSTEM PRESSURE SAFETY EVALUATION ADMINISTRATIVE VEHICLE 1753 136
CHRONOLOGY OF EVENTS POINT BEACH UNIT 1 - STEAM GENERATOR 10/05/79 POINT BEACH UNIT 1 SHUTDOWN FOR REFUELING AND STEAM GENERATOR INSPECTION CONFERENCE CALL WITH LICENSEE 10/29/79 11/05/79 MEETING WITH LICENSEE 11/16/79 LER 79-017/0lT-0 ISSUED BY LICENSEE 11/19/79 LER RECEIVED BY NRC (ISE) 11/20/79 MEETING WITH LICENSEE 11/23/79 LICENSEE SUBMITTAL 11/28/79 COMMISSION BRIEFING 11/30/79 CONFIRMATORY ORDER AND SUPPORTING SER 11/30/79 LER LOGGED INTO TERA 12/01/79 RESUMPTION OF POWER AT POINT BEACH UNIT 1 gel 1ERIC MEETING WITH 18E (POINT BEACH - ONE TOPIC 12/04/79 0F DISCUSSION) 12/07/79 NRR BECAME AWARE OF LER 12/07/79 TELEPHONE CONVERSATION WITH ENVIRONMENTAL DECADE REGARDING AXIAL LOCATION OF DEFECTS 12/11/79 POINT BEACH UNIT 1 SHUTDOWN DUE TO STEAM GENERATOR LEAK 1753 137
(2)
CHRONOLOGY OF EVENTS POINT BEACH UNIT 1 - STEAM GENERATOR 12/13/79 CONFERENCE CALL TO LICENSEE - REQUESTED EXPLANATION OF LER DATA DISCUSSION OF 5 TUBE DEFECTS AB0VE THE TUBE SHEET 12/13/79 WITH THE ENVIRONMENTAL DECADE 12/16/79 TRIP TO POINT BEACH UNIT 1 12/18/79 ADDITIONAL REQUEST FOR INFORMATION REGARDING 5 TUBE DEFECTS - REQUIRED BEFORE RESTART 12/21/79 RECEIPT OF LICENSEE'S SUBMITTAL OF DECEMBER 21, 1979 REVIEW 0F LATEST SUBMITTAL 12/21-22/79 1753 138
SUMMARY
OF PRIMARY AtlD SECONDARY OPERATING CONDITIONS NORMAL OPERATING CONDITIONS 100% POWER PRIMARY SECONDARY PRESSURE = 2250 PSIA PRESSURE = 800 PSIA
- HOT LEG TEMPERATURE = 597 F TEMPERATURE = 5000F COLD LEG TEMPERATURE = 542 F AP = 1450 PSI OPERATING CONDITIONS 11/30/79 TO PRESENT 80% POWER PRIMARY SECONDARY PRESSURE = 2250 PSIA PRESSURE = 600 PSIA
- HOT LEG TEMPERATURE = 557 F TEMPERATURE = 450 F COLD LEG TEMPERATURE = 510 F AP = 1650 PSI PROPOSED OPERATING CONDITIONS 80% POWER PRIMARY SEC0flDARY PRESSURE = 2000 PSIA PRESSURE =600PSfA*
HOT LEG TEMPERATURE - 557 F TEMPERATURE = 450 F COLD LEG TEMPERATURE - 510 F aP = 1400 PSI 1753 139 REDUCED PRIMARY AND SECONDARY OPERATIllG PRESSURES 1)
REDUCE THE POTENTIAL FOR CORROSION DUP.ING 40RMAL OPERATION 2)
MITIGATE THE EFFECTS OF POSTULATED ACCIDENTS (MSLB AND LOCA) ON TUBE INTEGRITY SEC0flDARY SYSTEM PRESSURES ARE DICTATED BY SATURATED CONDITIONS