ML19317G796
| ML19317G796 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 11/30/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19317G763 | List: |
| References | |
| NUDOCS 8004010624 | |
| Download: ML19317G796 (3) | |
Text
UNITED STATES
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'4 NUCLEAR REGULATORY COMMISSION y'
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h W ASHINGTON, D. C. 20665 1.
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SAFETY EVALUATION OF THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMEt'T NO.16 TO FACILITY OPERATING LICENSE NO. DPR-54 AND REANALYSIS Of Et1EPGENCY CORE C00LIflG SYSTEli PERFORMANCE SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SEC0 NUCLEAR GENERATING STATION DOCKET NO. 50-312 Introduction By letter dated September 27, 1977, the Sacranento Municipal Utility District (the licensee) requested changes to the Technical Specifications appended to Facility Operating License No. DPR-54 for the Rancho Seco Nuclear Generating Station (the facility).
The proposed amendment would change the surveillance testing of the reactor internal vent valves.
By letter dated August 30, 1977, the licensee referenced a reanalysis of the Emergency Core Cooling System (ECCS) performance with revised reactor coolant system pressure drop characteristics using the same ECCS model previously approved for the facility.
Evaluation By letter dated September 27, 1977, the licensee proposed a change to the surveillance testing of reactor internal vent valves that makes the requir.ed opening differential pressure for the reactor internals vent valves equivalent to 1.0 psid.
The present Technical Specification requires manual actuation of the. vent valves to verify that the valve begins to open from the fully closed position with a force equivalent to <.15 psid and is fully open with a force equivalent. to <.30 psid.
The licensee has shown that this change has no significant effect on the peak cladding temperature (PCT) during the limiting Loss of Coolant Accident (LOCA),
i.e., <3 F.
This is not a significant increase and does not cause the limiting LOCA PCT to exceed any of the 10 CFR 50.46 criteria, nor does this change affect which LOCA break is limiting.
Based on this information and the continued surveillance requir2ments on the reactor internals vent valves, we find this change to be acceptable.
8004010 b h
s The licensee's analyses, which were presented as justification for operation, were conducted in compliance with flRC's regulations and approved methods and, furthermore, are conservative relative to flRC regulations.
The proposed changes to the Technical Specifications are acceptable on the bases that the health and safety of the public will not be endangered by operation in the proposed manner.
By letter dated August 30, 1977, the licensee referenced a reanalysis of the ECCS performance with revised reactor coolant system (RCS) pressure drop characteristics using the same ECCS model previously approved for Rancho Seco.
This reanalysis was performed because of an identified error in the input value to the reactor vessel inlet nozzle U-baffle pressure loss characteristics.
The reanalysis shows that lower peak cladding temperatures (PCT) would be obtained for the worst break analysis during a postulated LOCA. The trends of the break spectrum, sensitivity, and LOCA limits studies for the previously approved analysis for Rancho Seco remain valid.
Therefore, only the limiting size break needed reanalysis.
The reduction in PCT as compared to that for the generically approved ECCS analysis (B&W-10103)* was due to the enhanced core flow during blowdown (more cooling), lower metal-water reaction rates (because of lower temperatures, less heat generation due to exothermic reaction),
and improved reflooding of the core (cooling attained sooner).
These benefits are based on an improved system pressure distribution; i.e.,
the reanalyzed RCS pressure drops are less than that assumed from B&H-10103.
The revision to the RCS pressure drops is based on both experimental and analytical verification techniques.
Pressure drop measurements made during the Oconee 1 hot functional testing are valid for Rancho Seco, since it,is identical to Oconee 1.
The pressure drop characteristics within the reactor vessel and the Once Through Steam Generator were analytically established to match this data. Additionally, there were vessel model flow tests which further substantiate the decrease in pressure drop observed in the hot functional test data and established by analysis.
Originally, the reactor vessel inlet nozzle was erroneously assumed to be a long leg U-baffle. As shown by tests, the change in pressure drop for this component between originally assumed and-as-built conditions is substantial.
All the changes to RCS pressure drops have been verified experimentally and analytically.
- Babcock & Wilcox Topical Report, "FCCS Analysis of B&W's 177-FA Lowered-Loop NSS" (BAW-10103), June 1975.
m
. We have reviewed the RCS pressure drops and their impact on the ECCS performance analysis.
We agree with the licensee that the ECCS calculations for the current Rancho Seco fuel loading are in con-pliance with the criteria of 10 CFR 50 Section 50.46 and Appendix K.
Although the reanalysis has lower PCT than those of B&W-10103, the -
allowable Linear Heat Generation Rate limits for Rancho Seco will be maintained at the same values as previously approved.
We find this analysis acceptable.
Environmental Consideration the amendment does not authorize a change We have determined that in effluent types or total amounts nor an increase in power level Having and will not result in any significant environmental impact.
made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environme'ntal impact statement, or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consecuences of accidents previously considered
- gnificant decrease in a safety margin, the and does not involve ?
amendment does not in,olve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: flovember 30, 1977