ML20070K974
| ML20070K974 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 07/22/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070K973 | List: |
| References | |
| NUDOCS 9407280119 | |
| Download: ML20070K974 (2) | |
Text
i.
O " "'00g A
fx A
0 S
UNITED STATES 5'
NUCLEAR REGULATORY COMMISSION
(,
,/
WASHINGTON, D.C. 2055 50001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 109 TO FACILITY OPERATING LICENSE N0. NPF-2 SOUTHERN NUCLEAR OPERATING COMPANY. INC.
JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 1 E0CKET NO. 50-348
1.0 INTRODUCTION
By letter dated June 17, 1994, Southern Nuclear Operating Company (SNC) submitted a request for changes to the Joseph M. Farley Nuclear Plant, Unit 1 (Farley 1), Technical Specification (TS) to nuclear enthalpy rise hot channel factor (F"(1) increase the limit of the
) for VANTAGE 5 fuel as given in TS 3.2.3, from 1.65 to 1.70 (for full power,o,peration) and (2) change the associated action statement to more closely follow the guidance of the Westinghouse improved standard TS (NUREG-1431) for this specification. Also proposed were changes to the Bases for TS 2.11 (Safety Limits) to reflect the F" increase, and the Bases for TS 3.2.3 (Nuclear Enthalpy Hot Channel Factory to reflect the changes to the action statement. The current Farley I core contains both VANTAGE 5 and LOPAR fuel assemblies.
The change is proposed because increased margin is available for VANTAGE 5 fuel and it provides additional flexibility in core design and operation, including the current cycle in which measured F"3, values have closely approached the 1.65 limit.
2.0 EVALUATION To justify the proposed change, SNC has evaluated all relevant transient / accident analyses, and has reanalyzed the large break loss-of-coolant accident (LBLOCA) and fuel handling accident after determining that all current analyses of events involving F,, except for these two events, 3
bound the effects of a 1.70 F"3,.
The most limiting LBLOCA was reanalyzed using the Westinghouse 1981 evaluation model with BART and BASH, which is the current model of record for Farley 1 VANTAGE 5 fuel.
The revised peak clad temperature is 1957 F.
The metal-water reaction amounts are also within 10 CFR Part 50 limits.
The fuel handling accident analyses was determined to be the only radiological event not already bounding the use of a 1.70 F"a.
The accident was reanalyzed with the new value and all acceptance criteria were met.
It was determined that the increase in F"3 will cause no significant increase in dose above the refueling canal and spen,t fuel pool.
Thermal calculations were done for the fuel pool and no significant increase in clad temperature was found.
9407200119 940722 ADOCK0500g8 T2DR
e, 9 The staff review has concluded that SNC has appropriately examined the current status of Farley I safety analyses with a full power F"3 value of 1.70 and has provided appropriate analyses for events and conditions not already covered by existing analyses.
The SNC has also proposed to change the action section of TS 3.2.3.
The changed statement closely parallels the action statement of the new Westinghouse Standard TS (NUREG-1431). The primary effect of the change is to increase the time to reduce power to less than 50 percent, when F"a is above the (power dependent) limit, from 2 to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
The time change and the language of the statement are in accordance with the staff approved version in NUREG-1431 and are acceptable.
3.0
SUMMARY
The staff has reviewed the information submitted by SNC for Farley 1 proposing TS changes relating to F"a.
Based on this review, it has been concluded that appropriate information was submitted and the proposed changes to F",, ad associated TS 3.2.3 action statements are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the State of Alabama 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. 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 i
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 J
public comment on such finding (59 FR 32249).
Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR i
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 l
the amendment.
6.0 CONCLUSION
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:
H. Richings Date: July 22, 1994
,