ML20113D273
| ML20113D273 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/27/1996 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20113D260 | List: |
| References | |
| NUDOCS 9607030079 | |
| Download: ML20113D273 (3) | |
Text
p2 Etog UNITED STATES g
j NUCLEAR HEGULATORY COMMISSION 2
WASHINGTON. D.C. 206664 201 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0. 201 TO FACILITY OPERATING LICENSE DPR-57 AND AMENDMENT NO.142 TO FACILITY OPERATING LICENSE NPF-5 GEORGIA POWER COMPANY. ET AL.
EDWIN I. HATCH NUCLEAR PLANT. UNITS 1 AND 2 e
DOCKET NOS. 50-321 AND 50-366
1.0 INTRODUCTION
By letter dated February 21, 1996, as supplemented by letters dated May 1, 4
1996 and June 4,1996, Georgia Power Company, et al. (the licensee), proposed license amendments to change the Technical Specifications (TS) for the Edwin I. Hatch Nuclear Plant, Units 1 and 2.
The proposed revision would change the Drywell Air Temperature Limiting Condition for Operation (LCO) from less than or equal to 135*F to less than or equal to 150*F. The proposed change would provide a margin for the primary containment Drywell Air Temperature LCO when prolonged summer and high river temperatures are experienced. Also, a strictly editorial correction to a Final Safety Analysis Report (FSAR) reference would be made.
The reference to FSAR Section 6.2 would be changed to FSAR Section 5.2.3.2 for Hatch Unit 1.
The May 1 and June 4, 1996, letters provided clarifying information that did not change the scope of the February 21, 1996, application and the initial proposed no significant hazards consideration determination.
i 2.0 EVALUATION The licensee stated that, as part of the Plant Hatch Power Uprate Program, the pertinent design basis analyses for the containment were performed assuming an initial drywell average air temperature of 150*F, an initial drywell pressure of 1.75 psig, and a suppression pool temperature of 100*F.
l Of the analyses that were performed, those that are affected by the proposed increase in drywell temperature include:
short-term containment pressure / temperature response to the design basis loss-of-coolant accident (DBA-LOCA)
DBA-LOCA containment dynamic loads long-term containment pressure / temperature response to DBA-LOCA 4
drywell pressure / temperature response to small steamline breaks 4
9607030079 960627 PDR ADOCK 05000321 P
1 The short-tern LOCA response evaluation utilized the M3CPT computer code. The long-tern LOCA response and the small steamline break evaluations utilized the SHEX computer code. ' The results of the licensee's analyses are:
(1) Operation with the drywell air temperature less than or equal to 150*F will not result in any safety concerns associated with primary containment system performance.
(2) Peak drywell prc -.ures will remain below design drywell pressures, and drywell structure temperatures will. remain below design temperatures.
(3) For Unit 2, the peak ambient drywell air temperature is below the drywell structure design temperature of 340'F.
1 4
(4)
For Unit 1, the peak ambient drywell air. temperature is slightly above the drywell structure design temperature of 281'F during the initial 15 seconds of the limiting accident. An evaluation concluded that the actual drywell structure design temperature is not exceeded. This condition was previously approved by the NRC staff during its review of the power uprate submittal per the staff's safety evaluation Section 3.7.1.2 dated August 31, 1996.
The licensee's evaluation of the effects of increased initial drywell temperature on equipment qualification found that drywell temperature will increase by only l'F during the postulated worst-case scenario (i.e., small steamline break), and that the resultant drywell temperature remains below the existing equipment qualification temperature envelopes for Hatch Units 1 and 2.
The licensee will continue to monitor ambient drywell temperatures and the qualified lifetimes of components will be adjusted as necessary based on..
the actual temperatures that exist.
The licensee's evaluation also concluded that the proposed increase in drywell temperature will not have an adverse effect on the primary containment structure and pipe supports.
The licensee evaluated reactor water level instrument calibration assuming drywell temperatures up to 170*F. The results showed that a change in calibration endpoints from 135'F to 170*F had a negligible effect upon setpoint available margins. This temperature bounds the expected actual temperature in the vicinity of the instrument sensing lines, assuming the drywell average allowable temperature is less than 150*F. The licensee concluded that since the instrument calibration impact is negligible, no changes in instrument setpoints are required.
The NRC staff has reviewed the information submitted by the licensee in i
support of the proposed increase in drywell temperature for Hatch Units 1 and 2, as discussed above. The licensee has evaluated the effects of increased drywell temperature on the containment and equipment located in the drywell, and has found.that the effects are insignificant and of no consequence to public health and safety.
Based on the information that has been provided, the NRC staff agrees with the licensee's assessment and the use of computer codes SHEX and M3CPT for the above purpose.
Therefore, the proposed increase in drywell temperature is acceptable.
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3.0 STATE CONSULTATION
j In accordance with the Commission's regulations, the Georgia State official was notified of.the proposed issuance of the amendments. The State official had no comments.
4.0 ENVIRONMENTAL CONSIDERATION
4 The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has detennined that the amendments involve no significant increase in the amour.ts, 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 amendments involve no significant hazards consideration, and there has been no public comment on such finding (61 FR 18167 dated April 24,1996).
Accordingly, the amendments meet 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 amendments.
5.0 CONCLUSION
The Comission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the j
public will not be endangered by operation in the proposed manner, (2) such I
activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
J. Tatum R. Goel K. Jabbour Date: June 27, 1996
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