ML20084F678
| ML20084F678 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/22/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20084F667 | List: |
| References | |
| NUDOCS 9506020289 | |
| Download: ML20084F678 (2) | |
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UNITED STATES j
'j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20566 4 001
%,..... f" SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.181 TO FACILITY OPERATING LICENSE NO. DPR-51 ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT NO. 1 DOCKET NO. 50-313
1.0 BACKGROUND
By letter dated July 22, 1993, Entergy Operations, Inc. (the licensee) submitted a request for changes to the Arkansas Nuclear One, Unit 1 (ANO-1)
Technical Specifications (TSs).
The requested changes would change TS 5.2.1 which specifies a value for the volume of the ANO-1 reactor building. The current vglue,of the containment volume in Section 5.2.1 of the ANO-1 TSs is 1.91 X 10 ft. The licensee states that this is an approximate value and that a smaller value is used for the reactor building peak pressure analyses.
This results in a higher peak pressure which is conservative.
The original safety analysis report reactor building pgak pressure analysis used a reactor building internal volume of 1.8656 X 10 ft.
The licensee stated in the July 22, 1993, submittal that during a penetration design review, a non-conservative error was identified in the calculation of the reactor building net free volume.
Several other errors were subsequently discovered. When these errors were correc netfreevolumedecreasedfrom1.8656X10}edthecalculatejreactorbuilding to 1.81 X 10' f t.
2.0 DISCUSSION AND EVALUATION The reactor building net free volume is an input to the reactor building design basis accident, the post-loss-of-coolant accident (LOCA) hydrogen generation calculations, the maximum hypothetical accident dose calculation, design basis LOCA calculations to demonstrate that the criteria of 10 CFR 50.46 are satisfied, and reactor building leak rate testing.
The licensee discussed how each of these analyses were affected by the revised value for containment net free volume.
The analysis of the reactor building peak pressure is most affected by the decrease in calculated reactor building volume. The results of this analysis are a new peak pressure of 54.0 psig and a temperature of 284*F.
These values exceed the original safety analysis report (SAR) values but are less than the reactor building design pressure and temperature of 59 psig and 286*F, respectively. The licensee performed the reactor building pressure calculations using the COPATTA computer code. This code is used for the present SAR analyses (see Section 14.2.2.5.5.2).
9506020289 950522 PDR ADOCK 05000313 p
. Since design basis LOCA calculations are most conservative when the containment volume is maximized, the licensee did not perform any additional calculations for this case.
3.0 TECHNICAL CONCLUSION The staff has reviewed the licensee's analyses accompanying the proposed change to the reactor building net free volume. The licensee has addressed all analyses which would be affected by the change and has demonstrated that the facility still meets all applicable safety criteria. The methods used by the licensee are those used and approved by the staff previously and are therefore acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendment.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (58 FR 67843). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUUQH The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor:
R. Lobel Date: May 22, 1995