ML20100C491
| ML20100C491 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 03/25/1985 |
| From: | Andrews R OMAHA PUBLIC POWER DISTRICT |
| To: | John Miller Office of Nuclear Reactor Regulation |
| References | |
| LIC-85-122, NUDOCS 8503290110 | |
| Download: ML20100C491 (2) | |
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Omaha Public Power District 1623 Harney Omaha. Nebraska 68102 402/536 4000 March 25, 1985 LIC-85-122 Mr. James R. Miller, Chief Office of Nuclear Reactor Regulation Division of Licensing U. S. Nuclear Regulatory Commission Washington, DC 20555
References:
(1)
Docket No. 50-285 (2)
Letter fran OPPD (W. C. Jones) to NRC (Robert A. Clark),
dated December 29, 1982, LIC-82-410.
Dear Mr. Miller:
Error in LOCA-ECCS Analysis for Fort Calhoun Station On March 15, 1985, Omaha Public Power District received a preliminary notifica-t'on from our fuel vendor, Exxon Nuclear Co. (ENC), that a coding error had been discovered in the Fort Calhoun Station Unit No.1 LOCA-ECCS analysis.
This error could potentially result in a 55 F non-conservatism in the Peak Clad Temperature (PCT). ENC also notified the District that formal notification would be provided by March 19, 1985.
Based upon the preliminary report, the District immediately assessed the implications of the error and determined that the 10 CFR 50.46 PCT limit of 2200 F was not exceeded by a preliminary revised value of 2107*F (55 F + 2052 F from the Cycle 8 analysis), the Reactor Protec-tive System (RPS) setpoints remained valid, all Technical Specifications re-mained valid and no unreviewed safety question existed pursuant to 10 CFR 50.59.
The District also assessed the situation against the criteria of 10 CFR 50.72 and 50.73 and determined it was not reportable. The NRC Resident Inspec-tor, Mr. L. A. Yandell, was notified of the potential problem and our conclu-sions regarding the safety significance and reportability.
On March 19, 1985, the District received formal notification fran ENC of the coding error discovered in the LOCA-ECCS analysis for the Fort Calhoun Station Uni t No.1.
The error was contained in the EXEM/PWR model used to perfonn the Cycle 8 Large Break LOCA Analysis as described in Reference 2.
The EXEM/PWR I
model has been approved by the staff for use on Fort Calhoun Station.
Based upon our March 20, 1985, review of the Cycle 8 EXEM/PWR LOCA-ECCS analysis re-sults and our discussions with ENC, we have concluded that the LOCA models were l
correctly used in the Cycle 8 LOCA-ECCS analysis and that no credit was taken for heat transfer augmentation (HTAF = 1.0).
The coding error discovered by ENC was contained in the T00DEE2 model used for reflood/heatup, specifically in r
the heat transfer augmentation option section.
The ef fect of the error was to
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l Mr. James R. Miller LIC-85-122 Page Two enhance the heat transfer coefficient during reflood, thus underpredicting the Peak Clad Temperature (PCT). The correct coding is of the form:
HTAF = 1.0 + a (FLOCAL - 1.0) where:
HTAF = heat transfer augmentation factor a = a constant for augmenting the heat transfer FLOCAL = the local heat transfer factor. This variable would have been set equal to 1.0 for Fort Calhoun Station to turn off the augmentation option, i.e., to result in a multiplier of 1.0.
The incorrect coding used was of the form:
HTAF = b
- FLOCfd where:
b = a constant for augmenting heat transfer In the notification, ENC infonned the District that a coding correction to the fonn of the first equation was performed and the limiting case reanalyzed. We have reviewed the results of the re-analysis which showed an increase in the PCT of 52*F, from 2052' to 2104, and found them to be acceptable. The i n-crease in PCT was assessed against the conclusions of the Fort Calhoun Station Updated Safety Analysis Report. The USAR shows that the Cycle 8 analysis (PCT
= 2052*F) is bounded by the results of the Cycle 6 analysis (PCT = 2195*F).
Since the increase to 2104'F is still bounded by the Cycle 6 results, the conclusions of the USAR remain valid.
It is our understanding that ENC initially had reported the error pursuant to 10 CFR 50 Appendix K in that it resulted in calculated fuel cladding tempera-tures greater than 20*F in excess of those previously reported. Subsequently, ENC has reported the error to the NRC pursuant to 10 CFR Part 21. Since re-ceipt of the fonnal results from ENC, our conclusions remain the same; i.e., no unreviewed safety question (10 CFR 50.59) exists; the defect or non-compliance does not constitute a substantial safety hazard and is not reportable under 10 CFR Part 21; nor does this discrepancy meet criteria of 10 CFR 50.72, or 10 CFR 50.73 requiring reporting to the Commission.
Based upon the reanalysis, the District finds that the PCT remains below the 2200*F limit and Fort Calhoun Station is still in compliance with 10 CFR 50.46.
Sin erely, f46 R. L. Andrews Division Manager Nuclear Production RLA/rh cc:
LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.
Washington, DC 20036 Mr. E. G. Tourigny, NRC Project Manager Mr. L. A. Yandell, NRC Senior Resident Inspector