ML19317F211
| ML19317F211 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/02/1976 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19317F208 | List: |
| References | |
| NUDOCS 8001080924 | |
| Download: ML19317F211 (3) | |
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SAFETY EVALUATION OF PROPOSED TECHNICAL SPECIFICATIONS CHANGE FOR OCONEE UNIT 3 CONCERNING CLADDING COLLAPSE The Duke Power Company has requested b ) approval of a change to the Technical Specifications for Oconee Unit 3 which would delete the limitation for cperation of Cycle 1 to 10,944 effective full power.
hours (EFPH).
The limitation was previcusly applied by the staff as an interim measure until supporting analyses were approved pertinent to cladding collapse resulting from fuel densification.
The fuel supplier, Babcock & Wilcox, has completed the analyses and provided them to the staff as Topichl Report BAW-10084(2)
The topical report was reviewed and approved by the staff for license applications with provisions specified on August 9,1974.(3) Subsequently, B&W resubmitted the topical report to the staff describing the analysis as refined by the staff provisions.
The Duke Power Company has applied the approved analysis to predict the most limiting time to cladding collapse for 0conee Unit 3.
The application has several identifiable conservations including those i
provisions previously identified by the staff.
The most limiting time tp
, cladding collapse exceeds the maximum projected three-cycle core life of 24,746.
Therefore, the staff concludes that the current limitation of 10,944 EFPH on Cycle 1 operations at Oconee Unit 3 can be deleted with no reasonable decrease in the health and safety of either the public or the plant personnel.
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m It should be noted that the prediction of cladding collapse, per se, does -not limit power operations of the reactor.(4)
If cladding collapse is predicted then:
a) limits should be established on the rate of power change, b) limits should be imposed on permissible steam generator leakage (PWR's),
c) ECCS eviluation model analyses should demonstrate that the calculated peak cladding temperature will not exceed 1800'F, and d) any operating restrictions necessary to assure the above limitations are incorporated into the Technical Specifications.
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3-REFERENCES 1.
Letter to B. C. Rusche, NRC, from W. O. Parker,. Duke Power Co.,
' dated April 16, 1976.
2.
BAW-10084, " Program to Determine In-Reactor Performance of B&W Fuels -
- Cladding Creep Collapse", May 1974.
- 3.
Letter to V. A. Moore, LWR-2, from V. Stello, TR, dated August 9,1974, subject:
TAR-970.
' 4.
WASH-1236," Technical Report on Densification of Light Water Reactor Fuels," November 14, 1972, Regulatory Staff, USAEC.
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