ML19246A596
| ML19246A596 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 05/24/1979 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | YANKEE ATOMIC ELECTRIC CO. |
| Shared Package | |
| ML19246A593 | List: |
| References | |
| NUDOCS 7907060159 | |
| Download: ML19246A596 (6) | |
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7590-01 UNITED STATES OF AMERICA NUCLEAR REGULATORY CCMMISSION In the Matter of
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MAINE YANKEE ATCMIC PCWER COMPANY
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Docket No. 50-309 (Maine Yankee Atomic Pcwer Station,)
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TERMINATION OF CRCER TO SHCW CAUSE I.
The Maine Yankee Atcmic Power Company (the licensee) is the holder of Facility Operating License No. CPR-36 which authorizes operation of the Maine Yankee Atomic Power Station (the facility) at power levels uo to 2630 megawatts thar :al (rated power). The facility, which is located at the licensea's site in Lincoln County, Maine, is a pressurized, vater reactor used for the commercial generation of electricity.
II.
Because certain safety related piping systems at the facility had been designed and analyzed with a computer code which incorrectly sum ed earth-cuake loads algebraically, the potencial. existed for cocoramising the basic defense in depth ar0vided by redundant safety systems in the evenc of an earthcuake. This compromising resulted from the possibility that an earthquake of the type that the plant :ust be deeigned for, could.
cause a pice rupture as well as degrade emergency cccling systems designed to :icigate sucn an accident. Therefore, by Order af the Director of Nuclear Reacccr Regulation (the Director) for the Nuclear Pegulatory Comission (NRC), dated Yarch 13,1979 (la FR 16506, "aren 19,1979),
One licensee was creered to shcw cause:
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7590-01 (1) Why the licensee should not reanalyze the facility piping systems for seismic loads on all potentially affected safety systems using an appropriate piping analysis computer code which does not cctbf : loads algebraically; (2) Why the licensee shculd not make any modifications to the facility piping systems indicated by such reanalysis to be necessary; and (3) Why facility operation should not be suspended pending such reanalysis and completien of any required mcdifications.
In view of the importance to safety of this matter, the Order was made immediately effective and the facility was required to be placed in the cold shutdown condition and remain in that mode until further Crder of the Ccamission.
III.
The facility is currently in the cold shutdcwn condition. Pursuant to the March 13, 1979 Order, the licensee filed a written answer to the Order by letter dated Acril 2,1979.
In this res;cnse the licensee stated that it has reanaly:ed all potentially affected safety systems for seismic leads using an appropriate method which tces not sum loads algebraically and these reanalyses indicate that two piping restraihts needed to be mcdified to account for ;ase plate fl exi b il i ty.
These acdifications have been comaletec. Technical J
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7590-01 succort for these conclusions was provided in the " Interim Repcrt by Stone & Webster, April 1,1979", " Containment Spray Piping Analysis of Pipe Supports H-51 and H-53, April 2,1979", and the licensee's submittals dated April 3, 12, 13, 19, 27 and May 2, 4, 5,15 and 18,1979.
Based on the acove, the licensee concludes there is no basis for continued suspension of facility operation as contemplated by the Order, and proposes:
(1) That the Director modify or rescind so much of his Order of March 13, 1979, as requires the continued shutdown of the facility.
(2) That tt u Director grant to the licensee such other and further relief as is proper in the circumstances.
The NRC staff has reviewed the licensee's submittals. This review included an evaluation of the codes which c:mpute pipe stresses resulting from the facility's response to an earthquake. The means by which piping responses are ccmbined in the codes that are currently a basis for the facility design are summarized belcw:
NUPIPE-SW This code ccmbines intramodal* res:cnses by tile square root of the sum Of tne squares (SRSS) and ccmbines intermcdal' responses by SRSS cr absolute sum for closely spaced mcdes.
"Moces are cefined as dynamic Dioing deflections at a given frecuency.
Intr 3 modal res:cnses are the ccaconents o' 0,Me, mcment and deflection within a mode.
Intermedal res:cnses are the conocnents of f:rce, mcment and deflection for all modes.
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PSTRESS/SHCCK 3 In this code the intramcdal responses are calculated by adding the absolute value of the responses due to tne vertical earthquake ccmconent to the root-mean-square of the responses due to the two hori:cntal earthquake ccm:cnents. The intermodal ccmconents are calculated bj the root-mean-square method.
PSTRESS/SHCCK 1 One of four versions of this code was reviewed.
In this version the largest modal response 's added (absolute sum) to the root-mean-square value of all other acdal resp nses.
Intramcdal re-sponses due to multi-directional earthquake excitation were not calculated since the code cnly produced responses parallel to a given earthquake cceponent excitation.
Because this code is not equivalent to current practice, the.1RC staff requested that the licensee demonstrate the conservatism of pipe stress as determined by this code. This was done by reanalysis of certain piping systems using currently acceptable metho ds.
STRUCL-SHAXE This code ccmbines intramodal respcnses by absolute sua and the inter cdai ras;onses by SRSS.
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. The NRC staf.' has determined that an algebraic summation of responses was not incorporated into any of the above listed codes. The NRC staff has further concluded that these codes provide an acceptable basis for the facility piping design.
Modifications of two piping supports (H-51, H-53) for the containment spray system were determined to be necessary as a result of the re-analysis. The modifications consisted of welding two stiffners to each support base plate to reduce the base plate flexibility. The modifications were compi.ted in accordance with the Yankee Operational Quality Assurance Program (YOQAD-1A) and are acceptable.
Bastd on the NRC staff's Safety Evaluation dated May 24,1979, the staff finds that, in accordance with the Order of March 13, 1979, the licensee hcs reanalyzed all potentially affected safety systems using an appropriate piping analysis which does not combine loads algebraically and has made those modifications to the facility piping systeins indicated by such reanalysis to be necessary.
The licensee's answer to the Order did not request a heariiig. The New Hampshire Legislative Utility Consumers' Council petitioned on April 2,1979, to be permitted to intervene in any proceeding which might arise frcm the Show Cause Order, but did r,ct request a hearing.
No other person recuested a hearing.
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IV.
Accordingly, pur.;2nt to the Atomic Energy Act of 1954, as amended, and the Comission's Rules and Regulations in 10 CFR Parts 2 and 50, IT IS DETERMINED THAT: The public health, interest or safety does not require the continued shutdown of the facility, AND IT IS HEREBY ORDERED THAT:
Effective this date the Order to Show Cause of March 13, 1979, and the proceeding thereon are terminated.
FOR THE NUCLEAR REGULATORY C0FNISSIO'i Harold R. Denton, Director Office of Nuclear Reactor Regulation Dated at Bethesda, Maryland this 24th day of May 1979.
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