ML15238A618
| ML15238A618 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/13/1982 |
| From: | Novak T Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| RTR-NUREG-0630, RTR-NUREG-630 NUDOCS 8207200119 | |
| Download: ML15238A618 (4) | |
Text
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RIngram OELD Qgket ile ORB#4 PM AEOD RC PD Gray File IE L PDR DEisenhut ACRS-10 Dockets Nos. 50-269/270/287 ORB#4 Rdg HOrnstein EBlackwood Mr. William 0. Parker, Jr.
Vice President - Steam Production Duke Power Company P. 0. Box 33189 422 South Church Street Charlotte, North Carolina 28242
Dear Mr. Parker:
On November 1, 1979, the NRC staff met with fuel vendors and industry representatives to discuss 10 CFR 50 Appendix K cladding swelling and rupture models. The staff presented new models (Ref. 1) that we believed met the requirements of 10 CFR 50 Appendix K and 10 CFR 50.46.
Each fuel vendor was then asked to show how, in light of the new models, the plants analyzed with their ElIs continued to meet the performance limits of 10 CFR 50.46.
The Babcock & Wilcox (B&W) response (Ref. 2) concluded that the impact of the NRC curves was small and did not result in cladding temperatures in excess of 2200 F.
Based on this and the licensee's response (Ref. 3) and the fact that the use of the data did not produce a large effect when the overall conservatism in an ECCS evaluation model calculation was considered, we concluded that there was no significant safety concern for your facility.
Since the November 1st meeting, we have received revised EMs from all LWR fuel vendors except 88W (Refs. 4 and 5).
Recently, Consumer's Power Company submitted (Ref. 6) a B&! supplemental ECCS calculation that used the cladding models of NUREG-0630. Our evaluation of this submittal is given in Reference 7. The results of the Midland calculation showed a need to reduce linear heat generation rates by 1 kW/ft at the 2-foot, 4-foot and 6-foot core elevations even though partially offsetting margins were applied in other parts of their ECCS model.
The supplerental ECCS calculation for Nidland is applicable to all 8&N plants (OR and OL);
therefore, it may be that the swelling and rupture data described in NUREG-0630 may have more effect on the peak cladding temperatures cal culated by the B&W EM than previously considered in 1979 and 1980.
Accordingly, you are requested to provide within the next 30 days, a schedule for submitting supplemental ECCS calculations similar to those provided for Midland in Reference 6. These calculations may be generic to other B&W plants and they may employ offsetting margins derived from compensating models. Justification will be required for the use of unapproved compensating models.
6207200119 820713 PDR ADOCK 05000269 P
PDR OFFICE SURNAME DATEb FORM_31__(1-80) j 0240 R
NRC FORM 318 (10-80) NRCM 0240 OFF ICIAL R ECO RD COPY USGPO: 1981-335-960
r Mr. William 0. Parker, Jr.
-2 In addition, permanent modifications to your ECCS evaluation models should be developed and implemented, and an implementation schedule should be provided to us as soon as possible, but not later than 60 days after this 30-day period.
We will be pleased to meet with you or any of your representatives to discuss this matter further.
Please contact your NRC Project Manager if you wish such a meeting or if you have questions.
The reporting and/or recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, 0N clearance is not required under P. L.96-511.
Sincerely, Thomas M. Novak, Assistant Director for Operating Reactors Division of Licensing
Enclosure:
List of References cc w/enclosure:
Mr. J. H. Taylor, Manager Licensing Babcock & Wilcox P. 0. Box 1260 Lynchburg, Virginia 245054 Mr. H. G. Parris, Manager of Power Tennessee Valley Authority 500A Chestnut Street, Tower II Chattanooga, Tennessee 37401 See next page OFFICEO........
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DATE NRC FORM 318 (10-80) NRCM 0240 OFFICIAL CORD COPY USGPO: 1981-335-960
Duke Power Company cc w/enclosure(s):
Mr. William L. Porter Duke Power Company P. 0. Box 33189 422 South Church Street Office of Intergovernmental Relations Charlotte, North Carolina 28242 116 West Jones Street Raleigh, North Carolina 27603 Oconee County Library 501 West Southbroad Street Walhalla, South Carolina 29691 Honorable James M. Phinney County Supervisor of Oconee County Walhalla, South Carolina 29621 Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission, Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Regional Radiation Representative EPA Region IV 345 Courtland Street, N.E.
Atlanta, Georgia 30308 William T. Orders Senior Resident Inspector U.S. Nuclear Regulatory Commission Route 2, Box 610 Seneca, South Carolina 29678 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 220, 7910 Woodmont Avenue Bethesda, Maryland 20814 Manager, LIS NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 J. Michael McGarry, III, Esq.
DeBevoise & Liberman 1200 17th Street, N.W.
Washington, D. C. 20036 Mr. L. G. Lanese GPU Nuclear Corporation 100 Interpace Parkway Parsippany, New Jersey 07054
REFERENCES
- 1. D. A. Powers and R. 0. Meyer, "Cladding Swelling and Rupture Models for LOCA Analysis", NRC Report NUREG-0630, April 1980.
- 2. J. H. Taylor (B&W) letter to D. G. Eisenhut (NRC), November 2, 1979.
- 3. W. 0. Parker (DPCo) letter to H. R. Denton (NRC), Docket Nos. 50-269/
270/287, January 8, 1980.
- 4. R. J. Mattson (NRC) letter to L. Lanese (GPU) and R. G. Snipes (DPCo),
September 8, 1981.
- 5. L. C. Lanese (GPU) letter to R. J. Mattson (NRC), November 3, 1981.
- 6. J. W. Cook (CPCo) letter to H. R. Denton (NRC), Docket No. 50-329/330, April 2, 1982.
Enclosure:
B&W Report No. 12-1132424, Rev. 0, "Bounding Analysis Impact Study of NUREG-0630".
- 7.
"Safety Evaluation Report Related to the Operation of Midland Plant, Units 1 and 2",
NRC Report NUREG-0793, May 1982.