ML20209C254
| ML20209C254 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 03/20/1986 |
| From: | Bernero R Office of Nuclear Reactor Regulation |
| To: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| Shared Package | |
| ML20209C152 | List:
|
| References | |
| FOIA-87-185 GL-84-11, NUDOCS 8704280631 | |
| Download: ML20209C254 (8) | |
Text
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5 7590-01 UNITED STATES OF AMERICA I
NUCLEAR REGULATORY COMMISSION In the Matter of PHILADELPHIA ELECTRIC COMPANY Docket No. 50-278 (PeachBottomAtomicPowerStation, Unit 3)
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4 ORDER CONFIRMING LICENSEE ComITMENTS ON PIPE CRACK RELATED ISSUES I.
The Philadelphia Electric Company (the licensee) and three other co-owners are holders of Facility Operating License No. DPR-56 which authorizes the operation of the Peach Bottom Atomic Power Station, Unit 3 (the facility), at steady-state power levels not in excess of 3293 megawatts themal. The facility is a boiling water reactor located at the licensee's site in York County, Pennsylvania.
II.
During the current 1985 refueling outage at Peach Bottom, Unit 3, augmented inservice inspection was performed on the recirculation and residual heat 4
B704280631 870424 PDR FOIA MORIART87-185 PDR 4
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removal (RHR) piping in accordance with the guidance given in Generic
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Letter 84-11. Overall, out of a total of 132 welds ultrasonically inspected, a total of 40 were found to show ultrasonic indications:
ten welds were thermal sleeve attachment welds in the recirculation inlet safe ends,18 welds were in the recirculation piping system (4-12" riser welds and 14-28" welds) and 12 welds were in the RHR piping system (4-20" welds and 8-24" welds).
The licensee's consultant, General Electric Company (GE), performed crack growth analysis for each of the 40 cracked welds. The results of the analyses (October 11,1985) indicated that 22 cracked welds (10 inlet safe ends, 5 recirculation welds and 7 RHR welds) did not require repair because the final crack sizes in those welds at the end of the 18-month operating period were shown to meet the requirements in Generic Letter 84-11. GE also performed the weld overlay designs for the licensee. Eighteen (18) cracked welds required weld overlay repairs. The overlay weld metal was 3091 stainless s
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steel material, resistant to intergranular stress corrosion cracking (IGSCC).
The overlay repair was designed to provide a full structural pipe reinforcement and meets the allowable limits of ASME Code Section XI IWB-3640.
Cracks were identified on both sides of the piping to junction box of the "A" core sparger header. GE performed an evaluation which concluded that no modifications or repairs were required for safe operation. However, the licensee has installed two brackets at the location of the cracks to provide further assurance of core spray operability and safe reactor operation.
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. Similar brackets were installed on the "B" core spray header although no cracks were identified in that piping.
The staff has reviewed the licensee's submittals including the inspection results, GE's fracture mechanics evaluation of IGSCC, weld overlay designs, and the licensee's safety evaluation-of cracking in the recirculation inlet safe ends, recirculation welds and RHR welds, to support the continued operation of Peach Bottom Unit 3 for one fuel cycle (18 months) in the present configuration. The NRC staff confirmed that the licensee's ultrasonic procedures, calibration standards, equipment and IGSCC detection capabilities were acceptably demonstrated. The staff's detailed analysis is contained in the concurrentl; issued Safety Evaluation. Our analysis, including independent crack growth calculations for three welds of most concern, indicates that safe operation with these three welds is attainable for a period of 8 to 10 months, but not necessarily for a full 18 month fuel cycle.
III.
Because the uncertainties in crack sizing and growth rate still remain as discussed above, the staff has determined that improvements in the monitoring in the containment for unidentified leakage are required; therefore, new limiting conditions for operation and surveillance requirements have been developed. These enhanced surveillance measures should provide assurance that possible cracks in pipes will be detected before growing to a size that will' compromise the safety of the plant.
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4-The staff also has concerns about continued operation of Unit 3 for a full 18 month cycle with unrepaired welds. Therefore, we have determined that a mid-cycle shutdown (9 months from startup plus or minus 1 month) should be required in order to inspect certain welds by UT. The plans for mid-cycle examinations will be required to be submitted for NRC review and approval within 90 days of returning. Unit 3 to startup. Be.cause of our concern with the continued operation of three cracked inlet safe ends and welds (#10-0-03, #2-AS-08 and #2-BD-12), for a full 18 month fuel cycle, the licensee has committed by letters dated November 27, 1985, and January 7, 1986, to implement the following action for increasing the margin of safety in operating Unit 3 during the next fuel cycle:
1.
A mid-cycle inspection of the three unrepaired welds (RHR weld
- 10-0-3, recirculation piping welds #2-AS-08 and #2-BD-12) as well as the two inlet safe ends (safe ends F and J) during a period between 8 and 10 months after startup.
2.
The licensee has committed to strengthen the allowable leakage limits and take immediate corrective, actions as soon as the primary coolant leakage limits are exceeded. These augmented leakage limits will reduce the maximum allowable unidentified leakage from 5 gallons per mimute (gpm) to 2 gpm and the rate of increase of unidentified l'eakage in a 24-hour period from 2 gpm to 1 gpm. The details of the more restrictive leakage limits and the corresponding action statements are provided in Attachment A.
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The details for the mid-cycle inspection are to be provided to the NRC staff for review and approval within 90 days of Unit 3 being returned to service.
Based upon the above evaluation and our consideration of the licensee's comitments discussed above, the staff has detennined that the continued operation of the 12 unrepaired welds.and the 10 flawed safe ends for one fuel cycle (18 months) is acceptable provided the mid-cycle inspection results do not report excessive crack growth.
In view of the foregoing I have determined that the public health and safety requires that the licensee's comitments should be cor fir <.ed by an imediately effective Order.
IV.
Accordingly, pursuant to Sections 103, 1611, 161o and 182 of the Atomic Energy Act of 1954, as amended, and the Comission's regulations in 10 CFR 2.204 and Part 50 IT IS ORDERED EFFECTIVE IPEEDIATELY THAT:
1.
The licensee shall operate the reactor in accordance with the present requirements on coolant leakage in Se.ctions 3.6.C.1 and 3.6.C.3 of the Technical Specifications, as modified by Attachment A to this Order.
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' 2.
A mid-cycle shutdown and inspection shall be initiated by the licensee within 9 months (plus or minus 1 month) of returning Unit 3 to service from the current outage.
3.
Plans for the mid-cycle shutdown and inspection shall be submitted for NRC review and approval within 90 days of returning Unit 3 to service.
i The Director, Division of Boiling Water Reactor Licensing, may in writing relax or terminate any of the above provisions upon request from the licensee, if the request is timely and provides good cause for the requested action.
V.
The licensee or any other person who is adversely affected by this Order may request a hearing within 20 days of the date of publication of this Order in the Federal Register. Any request for a hearing shall be addressed to the Director, Office of Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission, Washington, D. C.
20555. A copy shall.also be sent to the Executive Legal Director at the same addreas. A REQUEST FOR A HEARING SHALL NOT STAY THE IMMEDIATE EFFECTIVENESS OF THIS ORDER.
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f If a hearing is to held, the Commission will issue an Order de~signating the time and place of any such hearing.
If a hearing is held concerning this Order, the issue to be considered at the hearing shall be whether the licensee should comply with the requirements set forth in Section IV of this Order.
This Order is effective upon issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
- r Ro ert M. Bernero, Director Division of BWR Licensing Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland, this 20th day of March 1986.
Attachment:
Temporary Reactor Coolant System Leakage Specifications
Docket No. 50-278 ATTACHMENT A PEACH BOTTOM ATOMIC POWER STATION - UNIT NO. 3 TEMPORARY REACTOR COOLANT SYSTEM LEAKAGE SPECIFICATIONS In lieu of Peach Bottom Technical Specifications 3.6.C.1 and 3.6.C.3 on reactor. coolant leakage, Unit 3 shall conform to the following Limiting conditions for Operation during fuel cycle 7.
1.
Reactor coolant system leakage to the primary containment i
shall be limited to:
a.
With reactor coolant temperature above 212 degrees F, the leak rate from unidentified
- sources shall not exceed 2 gallons per minute.
b.
When the reactor is operated in the "Run" mode, the rate of change of unidentified
- leakage shall not exceed 1 gallon per minute per 24-hour surveillance period.
c.
With reactor coolant temperature above 212 degrees F, the total leak rate shall not exceed 25 gpm averaged over any 24-hour surveillance period.
2.
Action Statement If the reactor coolant system leak rate is greater than a.
the limits in 1.a and 1.b above, reduce the leak rate to within the limits within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least Hot Shutdown within the next 12. hours and in Cold. shutdown within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
If the reactor coolant system leak rate is greater than the limit in 1.c above, be in at least Hot Shutdown I
within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and Cold Shutdown within the l
following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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- The unidentified leakage as determined by the drywell sump collection and flow monitoring system, may be adjusted by an identified leakage measurement.
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