ML20076N368
| ML20076N368 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 07/08/1983 |
| From: | Purple R Office of Nuclear Reactor Regulation |
| To: | GEORGIA POWER CO. |
| Shared Package | |
| ML20076N371 | List: |
| References | |
| TAC-51407, TAC-57339, NUDOCS 8307210339 | |
| Download: ML20076N368 (8) | |
Text
.-
A 7590-01 UNITED STATES OF AMERICA f
NUCLEAR REGULATORY COMMISSION In the Matter of
)
)
)
Docket No. 50-366
)
(Edwin I. Hatch Nuclear Plant,
)
Unit 2)
)
ORDER CONFIRMING LICENSEE COMMITMENTS ON PIPE CRACK RELATES ISSUES I.
The Georgia Power Company (GPC or the licensee) and three other co-owners are the holders of Facility Operating License NPF-5 which authorizes operation of the Edwin I. Hatch Nuclear Plant, Unit 2 (Hatch or the facility) at steady state reactor power levels not in excess of 2436 megawatts thermal. The facility is a boiling water reactor located at the licensee's site in Appling County, Georgia.
II.
During the current 1983 refueling outage at Hatch Unit 2, augmented inservice inspection was performed on the recirculation reactor heat removal and reactor water cleanup system piping in accordance with Office of l
Inspection and Enforcement Bulletin 83-02. The original sample size was expanded to 108 welds after ultrasonic indications were reported on welds in l
the original sampling.
Welds most likely to crack were selected for the 8307210339 83070s PDR ADOCK 05000366 P
7590-01
. expanded inspection. Overall, out of a total of 108 welds inspected, a total of 39 were found to show linear indications which consist of 2312-inch riser welds, four 22-inch manifold end cap welds, nine 28-inch recirculation welds, two residual heat removal system 20-inch welds and one residual heat removal system 24-inch weld.
All indications were reported to be parallel to the weld in the heat-affected-zone. The deepest indication reported in the 12-inch riser welds is 327. of wall thickness.
The deepest indication in the large-size pipe welds is 42% of the wall thickness in a 22-inch manifold end cap weld.
Evaluation by the licensee, submitted by letters dated May 26 and June 8, 1983, indicates that the projected crack sizes, due to intergranular stress corrosion cracking (IGSCC) and fatigue crack growth, in the 12 large-diameter defective welds at the end of an 18-month fuel cycle would be within the ASME Code limits.
The licensee's evaluation also showed that the 2312-inch riser welds and i
the four 22-inch manifold end cap 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.
All 23 of the defective 12-inch riser welds and three of the four defective end cap welds were repaired using a weld overlay process.
The remaining end cap was replaced. The licensee's evaluation showed that each weld overlay was designed such that the weld joint meets the ASME Code Section III requirements, including fatigue. The predicted ultimate
~
o-7590-01 failure load based on tearing modulus approach was calculated for each overlay design. The ultimate failure load was shown to be at least three times the normal applied loads, which provides a safety margin larger than that inherent in the Code.
The staff has reviewed the licensee's submittals including analysis of weld overlay design and the calculation of IGSCC crack growth, based on current crack growth data, to support the continuing service for an 18-month fuel cycle with the 25 overlay repaired 12-inch riser and 22-inch manifold end cap welds, and the 12 unrepaired large-diameter defective welds.
The staff has performed independent calculations of crack growth, also based on current crack growth data, on the worst circumferential crack among the 12 large-diameter defective welds. Our calculated final crack depth at the end of an 18-month period meets the Code limit with adequate margin. Therefore, based on the staff's calculations and review of the licensee's analyses using current crack growth data, we conclude that the continuous service of the 121arge-diameter defective welds without repair for one 18-month fuel cycle would be acceptable because the Code design margin is maintained. However, recent field experience has indicated that the current crack growth data may not be conservative.
As a result, the staff will require that the piping be inspected well before the completion of the
7590-01 18-month cycle to evaluate changes in the crack depth for already cracked pipe (both repaired and unrepaired) and any indications of new cracks in previously unflawed pipe.
III.
Although the calculations discussed above indicate that the cracks in the 12 large-diameter unreinforced welds will not progress to the point of leakage during the next fuel cycle, and margins are expected to be maintained over crack growth which coulo compromise safety, uncertainties in crack sizing and growth rate still remain.
Further, not all welds were examined, and significant cracks could be present in welds that were not examined.
Because of these uncertainties, we have determined that interim inspection of pipes and monitoring in the containment for unidentified leakage is required; therefore, new limiting conditions for operation i
and surveillance requirements are hereby issued. These enhanced surveillance measures will provide adequate assurance that possible l
cracks in pipes will be detected before growing to a size that will compromise the safety of the plant.
The staf f 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 will require that plans for corrective action i
l
7590-01 or modification including replacement of the recirculation and other reactor coolant pressure boundary piping systems during the next refueling outage be submitted for staff review at least three months before the start of the next refueling outage.
By letters dated May 25, June 2 and June 8, 1983, the licensee committed to the above described conditions on leakage monitoring and early submittal of inspection and/or modification plans.
By letter dated July 8,1983, the licensee also committed to submit its plans for interim inspection of some of the welds for staff review.
I have determined that the public health and safety requires that these commitments should be confirmed by an immediately effective Order.
IV.
Accordingly, pursuant to Sections 103,1611,161o 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 i
THAT:
1.
The licensee shall operate the reactor in accordance with the present requirements on coolant leakage in Sections 3.4.3 and 4.4.3 of the Technical Specifications, as modified by Attachment A to this Order.
a 7590-01 2.
Plans for additional inspections during the present 18-month cycle to include examination of the repaired and unrepaired pipe shall be submitted for staff review within 30 days of issuance of this order.
These plans shall include provisions for identifying crack depth and crack propagation rates to support continued plant operation.
3.
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.
4.
The Director, Division of Licensing, may in w:iting relax or terminate any of the above provisions upon written request from the licensee, if the request is timely and provides good cause for the requested action.
l 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.
1 l
s,*
)
7590-01
_7 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.
If a hearing is to be held, the Commission will issue an Order desig-nating 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 whetner the licensee should comply with the requirements set forth in Section IV of this Orde r.
This Order is effective upon issuance.
FOR THE NUCLEAR REGULATORY COMMISSION a
Robert A. Purple, Deputy Director Division of Licensing Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland, this 8th day of July 1983.
.I 4
o Attachment A AUGMENTED REACTOR COOLANT LEAK DETECTION 1.
Grab samples of the containment atmosphere, as discussed in the ACTION statement of Technical Specification 3.4.3.1, (one leakage detection system inoperable) will be obtained and analyzed once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, instead of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as presently required.
2.
Primary system gaseous and particulate system channel checks (Technical l
Specification 4.4.3.1.a) will be performed once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, instead of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> as presently required.
l 3.
Actions required by Technical Specification 3.4.3.2.c will be implemented if reactor coolant leakage increases by 2 gpm unidentified leakage within any 24-hour period or less instead of within any 4-hour period as presently required.
4.
With greater than 5 gpm unidentified RCS leakage, or 25 gpm total RCS leakage averaged over any 24-hour period, (Technical Specification 3.4.3.2 Action b.) the leakage rate will be reduced to within these limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or the provisions of this ACTION statement will apply. Presently, 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is allowed to reduce leakage rates.
5.
Monitoring of primary containment floor drain sump and equipment sump levels will take place once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, instead of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, as presently required (Technical Specification 4.4.3.2.a).
6.
Monitoring of primary containment atmospheric particulate and gaseous l
radioactivity levels will take place once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, instead of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, as presently required (Technical Specification 4.4.3.2.b).
7.
In addition to the present requirements of Technical Specification 3.4.3.1, operability of at least one of the leakage measurement instruments associated with each sump shall be maintained per Surveillance Requi rement 4.4.3.1.b.
8.
Perform a sensor check at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for the primary containment sump level and flow monitoring system in addition to the surveillance presently required by Technical Specification 4.4.3.1.b.
l
,