ML20198B052

Safety Evaluation Supporting Amend 85 to License DPR-3

ML20198B052
Person / Time
Site:	Yankee Rowe
Issue date:	10/31/1985
From:	Office of Nuclear Reactor Regulation
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ML20198B014	List: ML20198B008 ML20198B037 ML20198B052
References
NUDOCS 8511060283
Download: ML20198B052 (3)

Text

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. # 'g UNITED STATES

'! o NUCLEAR REGULATORY COMMISSION

.I E waswswoTow. o. c.2oses SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 85 TO FACILITY OPERATING LICENSE NO. DPR-3 YANKEE ATOMIC ELECTRIC COMPANY YANKEE NUCLEAR POWER STATION DOCKET NO. 50-29

1.0 INTRODUCTION

By letter dated March 18, 1985, as supplemented May 9 and May 30, 1985, the

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Yankee Atomic Electric Company (YAEC) submitted a request for changes to the Yankee Nuclear Power Station technical specifications.

The amendment would modify the pressurizer safety valve setpoint tolerance to conform to section VIII of the ASPE Boiler and Pressure Vessel Code.

A Notice of Consideration.of Issuance of Amendment to License and Proposed No Significant Hazards Consideration Detemination and Opportunity for Hearing related to the requested action was published in the Federal Register en July 31, 1985 (50 FR 31078). No comments or requests for

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heartop were received.

.2.0 EVALUATION The licensee requested that the technical specifications for pressurizer

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safety valves be modified to increase the setting tolerance from,+0%, -3%_

to +3%, -3%. The licensee justified the change by perfoming a new analysis of the bounding overpressure transient which is a complete loss of load from

, full power.

The Yankee Nuclear Power Station has two sefety valves on the pressurizer.

They have staggered lift settings of 2485 psig and 2560 psig. Following a loss of load transient the turbine stop valves would close causing heatup of the secondary system and a reduction of heat flow from the reactor. system.

'The continued power production by the core and the reduct.on in heat removal would cause the reactor system pressure and temperature to increase. The reactor system pressure boundary is protected from overpressure by the action of the reactor protection' system to trip the reactor and by the opening of the pressurizer safety _ valves.

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Following a loss of load the reector protection system would receive signals to trip the reactor in the following order.

1. Direct reactor trip on turbine trip

2. Peduction in steam generator level

3. High reactor system pressure 4.- Fiah pressurizer level In the analysis the licensee assumed that the third incoming trip signal (high pressure) was effective. Other conservative assumptions were that the pressurizer relief valve and spray as v: ell es the steam dump to the conde_nser failed to functien. The pressurizer safety valves were assumed i toopenat33'abovetheirnormalliftsettirts.

_ _. The staff previously approved an overpressure analysis for the Yankee Station tolerance of 0%. This analysis also utilized the GEMINI II code. Since that time overpressure protection has been improved by the addition of the

_ . high pressure reactor trip and by the addition of higher cepecity pressurizer safety valves which ellow fcr approximately 40% greater steam flow.

Staff review of the GEMINI II code was completed in 1977 (Ref. 2). The code

-was approved for analysis of everpressure transients including loss of load.

The staff approval was. based on the ability of the code to correlate operating plant transient data as well as a staff e.udit. The code contains a non-equili-brium pressurizer model which. predicts more conservative results (higher pressures) than the plant data. The analyser predicted that the highest peak reactor systen pressure would be 2650 psig which is less than 110% of the design pressure of 2500 psig. This result is acceptable under the staff's standard review plen as meeting the overpressure protection requirement of General Design Criterion 15 of the Commission's Pegulations.

The staff concludes that the reanalysis of reactor system overpressure by

. the licensee is acceptable and the technical specificatiers for_ safety valve lift settirr telerance may be changed as the licensee requested.

3.0 ENVIRONMENTAL CONSIDERATION

4 This amendment involves a change to 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 staff has determined that the amendment involves no significant increase in the amounts, and no significant change in-the types, of eny effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiction

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exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there hes been no public comment on such findino. Accordingly, this 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 this amendment.

4.0 CONCLUSION

The staff has concluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the will-rot be endangered by operation in the proposed manner, andsuch (2) public activities will be conducted in corpliance with the Commission's regulations and the issuance of this amendrert will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ACKNOWLEDGEMENT Principal Contributors: W. Jensen and J. Clifford

6.0 REFERENCES

1. X. Goller, NRC, letter to G. Andognini, YAEC, July 30, 1974.

2. R. Reid, NPC, letter to R. Groce YAEC, May 27, 1977.

Dated: October 31, 1985 9