ML20087F361
| ML20087F361 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 03/15/1984 |
| From: | Purple R Office of Nuclear Reactor Regulation |
| To: | COMMONWEALTH EDISON CO. |
| Shared Package | |
| ML20087F363 | List: |
| References | |
| NUDOCS 8403190077 | |
| Download: ML20087F361 (6) | |
Text
7590-01 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of
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COMMONUEALTH EDISON COMPANY
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Docket No. 50-249 (Dresden Nuclear Power Station, Unit No. 3)
ORDER CONFIRMING COMMONWEALTH EDISON COMPANY'S COMMITMENTS RE IGSCC INSPECTION I.
The Connonwealth Edison Company, (CECO, the licensee) is the holder of Facility Operating License No. DPR-25, which authorizes the licensee to operate the Dresden Nuclear Power Station, Unit No. 3 (the facility) at power levels not in excess of 2527 megawatts thermal (rated power).
The facility is a boiling water reactor located at the licensee's site in Grundy County, Illinois.
II.
As a result of inspections conducted at 18 operating boiling water reactors (BWRs) in conformance with recent.0ffice of Inspection and Enforcenent Bulletins (IE Bulletin No. 82-03, Revision 1, " Stress-Corrosion Cracking in Thick-Wall, large-Dianeter Stainless Steel Recirculation Systen Piping at BUR Plants," and IE Bulletin No. 83-02, " Stress Corrosion Cracking in large-Dianeter Stainless Steel Recirculation System Piping at BWR Plants"),
a potential safety concern regarding intergranular stress corrosion cracking (IGSCC) in prinary system piping was identified.
These bulletins requested selected' licensees to perforn a number of actions regarding inspection and testing of pipe welds.
8403190077 840315 PDR ADOCK 05000249 G
7590-01
-2 Results of these and other inspections pursuant to IE Bulletins 82-03 and 83-02 have revealed extensive cracking in large-diameter recirculation and residual heat renoval systen piping.
In almost every case, where inspec-tions were performed, IGSCC was discovered and, in nany cases, repairs, analysis, and additional surveillance conditions were required.
In view of the foregoing and the fact that the facility is.similar in design to plants where IGSCC has occurred, there was a significant potential for IGSCC to exist in this facility.
Therefore, inspection was required to detenaine the extent of IGSCC and to ascertain, if necessary, the degree of remedial action.
On August 26, 1983 an Order was issued to the licensee which required that the facility be shut down by September 30, 1983 and an IGSCC inspection be perforned. The facility was shut down on Scotember 30, 1983 pursuant to Section III.B of the Order and an IGSCC inspection was performed pursuant to Section III.C of the August 26, 1983 Ordar.
By letter dated December 9,1983, the licensee provided its plan for inspection and repair of welds covared by the Order of August 26, 1983. The plan provided that, to the extent practicable, the ultrasonic testing (UT) program of.exanination would encompass 100% of the Type 304 Stainless Steel piping welds of 4-inch and greater size in the recirculation system and the -
ASME Code Class 1 portions of the residual heat renoval systems, core spray external to the reactor pressure vessel and the reactor water cleanup system.
Specific welds which were not to be inspected were identified and explan-ations for their exclusion had previously been provided in the licensee's letter of October 26, 1983.
On February 15, 1984, the staff met with the
7590-01 l 1 licensee to discuss its progran and its findines, and to receive clarifying
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inforne tion.
By letter of !! arch 5, 1984, the licensee subnitted a report on the inspection and repair of welds covered in the Order of August 26, 1983.
Sy letter dated !' arch 12, 1984, the licensee provided additional information on the results of the inspection.
The NRC staff has reviewed and evaluated all the above reports and information provided by the licensee. That review is docunented in our Safety Evaluation dated Parch 15, 1984.
By letter dated March 15, 1984, the NRC notified the licensee that the facility could be returned to power.
Althouch the calculations performed by the licensee and evaluated by the staff indicate that the cracks in the repaired and unrepaired welds will l
not progress to the point of leakage during the operating cycle, and wide nargins are expected to be maintained over crack prowth which could compromise safety, uncertainties in crack sizing and growth rate remain.
Because of these uncertainties, we have determined that the following a;tions should be taken:
(1) The ASME Code-required system pressure tests and nondestructive exaninations on overlay repaired welds should be, satisfactorily completed prior to startup.
(2) The limiting conditions _for operation and surveillance Tr4uirements irposed by the August 26, 1983 Order should be continued. These 4
enhanccd surveillance measures will provide adequate assurance that j
possible cracks in pipes will be detected before growing to a size that ~ will. compromise the safety of the p' ant.
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7590-01 The staff also has sone concern regarding the long-term growth of IGSCC-cracks and its offect on the long-tern operation of the plant.
Therefore, we have determined that plans for inspections, corrective action and/or nodification including replacement of the recirculation and other reactor coolant pressure boundary piping systems during the next refueling outage must be submitted at least 90 days before the start of the next refueling outage.
In addition, the staff has determined that a justification for continued operation must be subnitted to NRC for review and approval prior to startup after the next refueling outage.
By letters dated March 5 and March 9, 1984, 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.
III.
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 THAT:
A.
Notwithstanding the current Technical Specifications for the facility the following compensatory measures shall be implemented:
- 1. -The reactor coolant system leakage shall be limited to a 2 gpm j
increase in unidentified leakage within any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period (leakage shall be moritored and recorded once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />).
i Should this leakage limit be exceeded, the unit shall
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7590-01
. 4 inmediately start an orderly shutdown. The unit shall be placed 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 within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
At least one Primary Containment sump collection and flow nonitor.ing system shall be operable. With the prinary containment sump collection and flow monitoring system inoperable, restore the inoperable system to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or immediately initiate en orderly shutdown I
and be in at least het 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
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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.
Plans for inspection, corrective actions, and/or nodification, including replacement of the recirculation and/or coolant pressure boundary piping systems, during the next refueling outage which is currently expected to begin approximately April 1985 shall be submitted at least three months before the s; art of that outage.
C.
At least one nonth prior to startup of the facility after its next refueling outage, a justification for c'ontinued operation shall be submitted fcr NRC review and approval.
D.
The.nirector, Division of Licensing, may, in writing, relax or terminate eny of_the above provisions upon written request from the licensee, if the request is tinely and provides good cause for the requested action.
IV.
The' licensee may request a hearing on this Order within 20' days of the date of publication of this Order in the Federal Recister. Any request for a hearing shall be addressed to the Director, Office of Nuclear Reactor
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7590-01 i
Regulation, U.S. Nuclear Regulatory Conmission, Washington, D.C.
20555. A copy shall also be sent to the Executive Legal Director at ti;e same address.
1 A REQUEST FOR HEARING SHALL NOT STAY THE IMMEDIATE EFFECTIVENESS OF THIS nRDtR.
If a hearing is to be held, the Commission will issue an Order designating the tine ard place of any such hearing.
If a hearing is held concerning this Order, the issue to be considered i
at the hearing shall be whether, on the basis of the matters set forth in Section II of the Order, the licensee should conply with the requirements set forth in Section III of this Order. This Order is effective upon 4
issuance.
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FOR THE NUCLEAR REGULATORY COMitISSION RobertA. Purple,DeputfDrector Division of Licensing Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland this 15 day of March 1984.
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