ML20077S542
| ML20077S542 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 09/01/1983 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| Shared Package | |
| ML20077S544 | List: |
| References | |
| NUDOCS 8309220123 | |
| Download: ML20077S542 (9) | |
Text
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-l 7590-01 ij UNITED STATES OF AMERICA i
NUCLEAR REGULATORY COMMISSION In the Matter of
)
PHILADELPHIA ELECTRIC COMPANY
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Docket No.
50-278 (Peach Bottom Atomic Power Station,
)
Unit 3)
)
ORDER CONFIRMING LICENSEE COMMITMENTS ON PIPE CRACK RELAiED 155UE5 1.
The Philadelphia Electric Company (the licensee) and three other co-owners are the 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 thermal.
The facilit'y is a boiling water reactor located at the licensee's site in York County, Pennsylvania.
II.
During the current 1983 refueling outage at Peach Bottom, Unit 3, augmented inservice inspection was performed on the recirculation and residual heat i
removal (RHR) system piping in accordance with Office of Inspection and Enforce-ment Bulletin 83-02.
The original sample size was expanded to 112 welds after ultrasonic (UT) indications were reported on welds in the original sampling.
This represents a total of 757. of all welds in these systems. A total of 37 welds in the recirculation and RHR systems were not inspected.
Examinations of these uninspected welds were either not necessary (e.g., conforming material as per NUREG-0313, Revision 1), or not practicable (e.g., high radiation, etc.).
8309220123 830901 PDR ADOCK 05000278 P
i 7590-01 1 Il The licensee also provided information on welas on the Reactor Water Clean-up a
j (RWCU) anc Core Spray Systems.
Both systems have had substantial portions of 1
their piping replaced with conforming material to reduce susceptability to intergranular stress corrosion c.acking (IGSCC).
A total of 25 non-conforming welds have not been inspected in these two systems.
The licensee provided technical justification for not inspecting these welds.
The staff has reviewed the licensee's submittal and has concluded that the uninspected welds are not likely to be cracked to the extent of compromising the safety of the plant.
Overall, out of a total of 112 welds inspected, a total of 15 were found to show linear indications: ten 12-inch riser to elbow welds, and five residual heat removal (RHR) system 20-inch suction line welds.
All indications were in the weld heat-affected-zone.
In the 1012-inch riser welds, all indications are oriented in an axial direction.
The deepest indication reported in the 12-inch riser welds is 92% of wall thickness with a length of about 0.75 incnes.
The deepest indication in the large-size pipe welds is 40%
l-of the wall thickness in a 20-inch RHR weld with a length of about 32 inches.
In the five 20-inch RHR welds, the indications are oriented predominatly in a l
circumferential direction.
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Evaluation by the licensee, submitted by letters dated August 9,1983, August 22, 1983 and August 30, 1983, indicates that the projected crack sizes, due to intergranular stress corrosion cracking (ILSCC) and fatigue I
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' crack growth, in two of the 20-incn RHR welds at the end of an 18-month l
fuel cycle would be within the ASME Code limits.
The licensee's evaluation also showed that the 1012-inch riser welds and the three 20-inch RHR suction line welds required repair for continued service because their calculated projected cracks would exceed the Code limits at the end of an 18-month fuel cycle.
However, all five RHR suction line welas and the ten 12-incn riser welds were repaired using a weld overlay process.
The minimum overlay thickness for the riser welds is 0.25-inch and for the 20-inch RHR weld it varies from 0.25-inch to 0.5-inch.
The overlay thickness is designed to meet the Code allowable limits in the new ASME Code Section XI IWB 3600 assuming the presence of through-wall cracks.
The length of the overlay varies approximately from 4 to 7 inches and is designed to reduce the stress at tne end of the overlay acting on the crack location.
RHR welds 10-0-7, 10-0-10, and 10-0-15 are bi-metallic welds (carbon steel to stainless steel) and the overlay is applied only to the stainless steel side of the welds.
Personnel of Region I confirmed that the weld overlay repairs were performed in accoraance*
with the qualified and approved procedures consistent with ASME Code requirements.
The staff has reviewed the licensee's subnittals including analysis of weld overlay design and the calculation of IGSCC crack growth, based e
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on current crack growth data, to support the ' continuing service for an
'i 18-month fuel cycle with the 15 overlay repaired welds.
Region I personnel confirmed that the licensee's UT procedures, calibration standards, equipment and IGSCC detection capabilities were acceptably demonstrated in accordance with IE Bulletin 83-02.
The licensee's overlay design analysis performed by General Electric is based on the conservative assumption that all cracks are
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through-wall cracks.
Consequently, its analysis did rot depend on and assumptions concerning UT sizing and the IGSCC crack gtowth rate.
The required minimum overlay thickness for each defective weld is calculated by using the methodology allowed in the new ASME Code Section XI IWB 3600 to meet the required Code safety margin.
For normal and upset condition, a safety margin of three is required and for the faulted and emergency condition, a safety margin of 1.5 is required.
Because the acceptable flaw in the normal condition based on new IWB 3600 is more liiniting, the acceptable flaw for the faulted condition need not be considered.
For RHR welds 10-0-5 and 10-0-6, the allowable flaw depth based on new IWB 3600 is conservatively calculated to be 75.57, and 78.1". of wall thickness respectively.
This calculation assumes the flaws to be fully circumferential in length and through the original pipe thickness in depth.
This assumption is very conservative because the worst reported I T indicatioT in these two welds is about 40*. through-wall in depth and U
5 less than half of the full circumference in length.
The Genera-1 Electric analysis has shown that a minimum overlay thickness of 0.5 inch is more i
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l-tnan enough to make the assumed tnrough-wall cracks (63% and 65% of the overlay repaired wall thickness) in tnese two welds meet the new
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calculated Code allowable limits (75% and 78%).
An allowable flaw i
depth of 82% was similarly calculated for RHR weld 10-0-7. This weld nas a worst reported flaw with about 35% of wall thickness in depth and about 7 inches in length.
With a minimum overlay thickness of 0.35 inch applied to this weld, the assumed fully circumferential through-wall crack (73% of the overlay repaired wall thickness) is well within the new calculated Code allowable limit (82%).
The in.dications reported by General Electric on RHR welas 10-0-10, and 10-0-15 may be overcalls because two independent UT examinations on the same two welds did not find any reportable indications.
Howeve r',
the licensee decided to apply an overlay with a minimum thickness of 0.25 inch to these two welds to increase the safety margin in structural strength and to prevent any potential leakage.
f.
We reviewed the weld overlay design calculation made by General Electric.
We concur in their conclusion that the overlay used will provice adequate reinforcement with Code required safety margin for a least the next fuel cycle of 3peration.
III.
_ Although the calculations discussed above indicate that the cracks j
in the 15 overlay repaired welds will not progress to the point of leakage 1
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.t during the next fuel cyc'le, and margins are expected to be maintained
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over crack growth which could compromise safety, uncertainties in crack sizing and growth rate still remain.
Further, not all welds were examined, and 9
significant cracks could be present in welds that were not examined.
Because of these uncertainties, we have 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 will provide adequate assurance that possible cracks in pipes will be detected before growing to a size that will compromise the safety of the plant.
The staff also has some concern regarding the long-term growth of IGSCC cracks and its effect on the long-term operation of the plant.
Therefore, we have determined that plans for corrective action or modification including replacement of the recirculation and other reactor coolant pressure boundary piping systems during the next refueling outage mQst be submitted for staff review at least three months before the start of the next refueling outage.
By letters dated August 9,1983 and August 24, 1983, the licensee committed to the above described conditions on leakage monitoring and early submittal of inspection and/or modification plans.
I have determined that the public health and safety requires that these commitments should be confirmed by an immediately effective Order.
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7590-01 ii
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IV.
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Accordingly, pursuant to Section 103,161i,161o and 182 of the Atomic Energy Act of 1954, as amended, and the Commission's regulations in 10 CFR Parts 2 and 50, IT IS HEREBY ORDERED EFFECTIVE IMMEDIATELY THAT:
1.
The licensee shall operate the reactor in accordance with the present requirements on coolant leakage in Sections 3.6.C and 4.6.C of the Technical Specifications, as modified by Attachment A to this Order.
2.
Plans for corrective actions and/or modification, including replacement of the recirculation and other reactor cooling pressure boundary piping systems during the next refueling outage shall be submitted for NRC review at least three months before the start of the next refueling outage.
3.
The Director, Division of Licensing', may in writing relax or terminate any of the above provisions upon written request
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from the licensee, if the request is timely and provides good cause for the requested action.
V.
The licensee may request a hearing within twenty (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 address.
A REQUEST FOR A HEARING SHALL NOT STAY THE IMMEDIATE EFFECTIVENESS OF THIS ORDER.
7590-01.
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If a hearing is to be held, the Commission will issue an Order Aj designating the time and place of any such hearing.
If a hearing is held y
concerning this Order, the issue to be considered at the hearing shall
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be whether the licensee should comply with the requirements set forth in Section IV of this Order.
This Order is effective upon issuance.
ORTHENUCQRREGULATORYCOMMISSION
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Darrell & Eisenhut, Director Division ( f Licensing Dated at Bethesda, Maryland this 1st day of September 1983.
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Attachnent A lij SURVEILLANCE AND LIMITTNG CONDITTON OF OPERATTON 1
FOR PEACH BOTTOM ATOMIC POWER STATION, UNIT 3
.1 4
- 7 7MITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3
- 3. 6.c. Coolant Leakage 4.6.c coolant Leakage
,l 1.
Any time irradiated fuel is 1.
Reactor coolant system leakage in the reactor vessel and reactor shall he determined by the coolant temperature is above primary containment (Drywell )
212 degrees F, the rate of sump collection and flow reactor coolant leakage to the monitoring syntem and recorded primary containment from unidenti-every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or less.
ficd sources shall not exceed 5 gallons per minute.
The rate of change of unidentified leakage shall not exceed 2 gallons per mi nute per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> surveillance period when the reactor is operated in the "Run" mode. In addition, the total reactor coolant system leakage into the primary contain-ment shall not exceed 25 gpm
.cveraged over any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> surveillance period.
2.
The primary containment (Drywell) sump collection and flow monitoring system shall be operable during reactor power operation.
From and after the time that this system is made or found to be inoperable for any reason, reactor
- power operation is permissible only during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless the system is made operable sooner.
For purposes of this para ara ph, the primary containment (Drywell) sump collection and flow monitoring system operability is defined as the ability to measure reactor coolant leskage.
3.
If the conditions in 1 or 2 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in at least Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in Cold Shutdown Condition within the
"'following 24" hours.
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