ML19318A886
| ML19318A886 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 06/17/1980 |
| From: | William Jones OMAHA PUBLIC POWER DISTRICT |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8006240324 | |
| Download: ML19318A886 (3) | |
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Omaha Public Power District 16 2 3 H A R N E Y OMANA, NCORASMA 68102
?tLEPMONE S36 4000 AREA CODE 402 June 17, 1980 Mr. Darrell G. Eisenhut, Director Division of Licensing V. S. Nuclear Regulatory Commission Washington, D. C.
20555
Reference:
Docket No. 50-285 Gentlemen:
The Omaha Public Power District received a letter from the Commission, dated May 7,1980, putting forth five additional TMI-2 related requirements for operating reactors.
For your information, the attached summary provides the District's commitment and associated implem.entation schedules for meeting these requirements.
Sincerely, i
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W, C. Jones Division Manager Production Operations WCJ/KJM/BJH:jm
- Attach, cc: LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N. W.
Washington, D. C.
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800624 032 (
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ATTACHMENT I.A.I.3, Shift Manning The criteria for shift manning has not been received by the Omaha Public Power District. After the staff recommendations are received, the District will forward comments to the staff.
I.A.3.1, Revise License Examinations Content The District is currently revising the training program at Fort Calhoun Station to incorporate the recommendations of the NRC's March 28, 1980, letter on operator training programs.
,I_.,C,5, Feedback of Operating Experience to Plant Staff Procedures will be developed as necessary to assure that important operating experience originating both within and outside the organi-zation are provided to operators and other personnel and are incorpo-rated into training and retraining programs.
The documentation of methods used to accomplish this will be submitted to the NRC by January 1, 1981.
II.K.3.1, Automatic PORV Isolation The District will design a system for automatic PORV isolation if the need is shown by the PORV failure reduction study discussed under i tem II.K.3.2.
This design will be completed and submitted to the NRC by July 1, 1981.
II.K.3.2, PWR Vendor Report Regarding PORV Failure Reduction The District will participate in a study on PORV failure reduction.
This study will probably be performed under the auspices of the CE Owners Group and will include an evaluation of safety valve failure rate.
It will be completed and submitted to the NRC by January 1,1981.
II.K.3.3, Reporting of Safety and Relief Valve Challenges Safety and relief valve challenges will be promptly reported to the NRC per the immediate reporting requirements of the Technical Specifi-cations, Section 5.
A history of safety and relief valve challenges will be included in the monthly operating report. This reporting program will commence on January 1,1981, and will cover a period of time beginning April 1, 1980.
II.K.3.5, Automatic Reactor Coolant Pumo Trio During LOCA
~ The District is participating with the Ci Owners Group on the resolution of the automatic trip of reactor coolant pumps during a LOCA issue. The Owners Group is in continuing discussion en this issue with the. Reactor. Systems Branch of the Division of System Integration of NRR.
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- An integral part of the resolution of this. issue is the modeling of the LOFT L-3-6 test for verification of small break LOCA models.
Since this test is not scheduled until December 17, 1980, it is anticipated that the CE Owners Group's study will not be ready by January 1,1981. This study.will be submitted on a schedule consistent with the LOFT test and the requirements of the Reactor Systems Branch.
II.K.3.17, Report on Use of ECCS By January 1, 1981, the District will submit a report to the NRC detailing outage dates and lengths of outages for all ECC systems for the years 1976, 1977, 1978, 1979, and 1950.
II.K.3.30, Revised Small Break LOCA Methods The District will address this item with the CE Owners Group for the Fort Calhoun Station. We de not feel that it would be useful for our fuel vendor (Exxon Nuclear Company) to submit revised small break LOCA methods for the Fort Calnoun Station.
Previous analyses have shown that the small break is not limiting with respect to fuel and, there-l fore, that core thermal power (the sole LOCA analysis area in which our fuel vendor has been involced) is completely defined by the large break LOCA. Further, it is well known that fuel characteristics are of secondary importance in determining the plant response to a small break LOCA; the most important parameters being the licensed core power level, the performance characteristics of the ECCS, the normal primary coolant loop operating temperature, and the elevation of the core with respect I
to the hot and cold legs.
This work will be completed'and submitted to the NRC by July 1, 1983.
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II.K.3.31. Plant Specific Calculations to Show Comoliance With 10CFR50.46 The District will address this issue with the reactor vendor (CE) for Fort Calhoun Station. A small break LOCA is not a limiting event from the point of view of the fuel; therefore, we expect that this requirement will be satisfied through bounding analyses of specific plant. groupings.
The results of the small break LOCA analysis will be submitted to the NRC by January 1,1983, or one year following staff approval of the LOCA analysis model.
III.D.3.4, Control Room Habitgoility The District will asNre that control room operators are adequately protected against the ef ects of accidental release of toxic and radio-active gases and that tne plant can be safely operated or shutdown under design basis accident conditions.
The required evaluation will be submitted to the NRC by January 1, 1981. Modifications, if needed, will be identified and scheduled for completion by January 1,1983.
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