ML20210D101

Safety Evaluation Supporting Issuance of Amend 83 to License DPR-54

ML20210D101
Person / Time
Site:	Rancho Seco
Issue date:	02/03/1987
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20210D062	List: ML20210D057 ML20210D085 ML20210D101
References
TAC-61473, NUDOCS 8702100018
Download: ML20210D101 (3)

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gv j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR RFACTOR REGULATIDF SUPP0PTING AMENDMENT NO. 83 TO FACILITY OPERATING LICENSF NO. OPP-54 SACRAMENTO MUNICIPAL UTILITY DISTPICT RANCHO SECO NUCLEAR GENERATING STATION CGCKET NO. 50-312

1.0 INTRODUCTION

By letter dated January 24, 1986, Sacramento Vunicipal Utility District (SVVD or the licensee) proposed a license amendment to Facility Operating License No. DPR-54 for the Rancho Seco Nuclear Generatina Station (RS).

In response to an NRC staff request, SFUD provided suoplemental information in a letter dated September 19, 1086.

The proposed amendment would:

(1) increase the setpoint for trip (i.e., shutdown) of the reactor on high pressure in the reactor coolant system from 2300 osin to 7355 psia; and (2) increase the armino threshold fnr anticipatory reactor trip (ART) on turbine trip from 20% of full power to 45% of full power.

2.0 EVALUATION Rancho Seco's Technical Specifications (TSs 2.2 and P.3) currently have a value of 2300 psig for the hiah pressure trip setpoint, and a value of 20% of full pcwer for the arming threshold for the ART on turbine trip.

These values were based on chanaes required by the NRC staff (Refererce 1), subsequent to the accident at Three Mile Island Unit 2, to reduce challenges to and openina of the power operated relief valve (PORV).

In April 1986, we completed our review of two Babcock & Wilcox (B&W) topical reports related to SMUD's request:

(1) " Justification for Raisina Setpoint for Peactor Trip on High Pressure," BAW-1890, September 1985 (Reference ?); and (2) " Basis for Raisina Arming Threshold for Anticipatory Reactor Trip or Turbing Trip," BAW-1893, October 1985 (Reference 3). The results of our review of the topical reports are contained in two Safety Evaluation Poports (References 4 and 5).

In those Safety Evaluation Reports, we:

(1) reviewed the basis for the proposed chanaes; (?) reviewed B&W's method of analysis of the effect of the proposed high pressure trip setpoint on PORV openinos; (3) compared the results of Ponte Carlo simulation for PORV openings with the NRC requirements contained in NUREG-0737 (Reference 6); and (4) reviewed the results of B&W's analysis of tha armina threshold for the ART on turbine trip. The NRC requirements include:

(1) the POPV will open in less than 5% of all anticipated overpressure transients IReference 6, Item II.K.3.7); and (2) the probability of a small-break loss of coolant accident caused by a stuck-open POPV will be less than 0.001 per reactor-year (Reference 6, item II.K.3.2).

In the Safety Evaluation Reports (References 4 and 5), we concluded on a generic basis that the proposed changes met the NRC reauirements, and should berefit plants by potentially reducing the reactor trip frequency.

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In regard to Ranchc Seco, the licensee states in their supplemental response dated September 19, 1986, that the accidents analyzed in Chapter 14 of the Final Safety Analysis Report (FSAR) were based on the high pressure trip setpoint of 2355 csig (not the current setpoint of 2300 psig) and no credit for ART. The increase in the high pressure trip setpoint and the increase in arming threshold for ART will not increase the frequency of challences to the main steam safety valves. Consequently, the accidents analyzed in the FSAR for Rancho Seco bound the proposed TS changes.

We have reviewed the proposed chances to TSs 2.2 and ?.3 for Pancho Seco.

The proposed changes would:

(1) increase the setpoint for trip of the reactor on high pressure in the reactor coolant system from 2300 psig to 2355 psig; and (2) increase the armino threshold for ART on turbine trip from 20% of full power to 45% of full power. As discussed above, we find that the proposed changes meet the apnlicable regulatory guidance and requirements and are, therefore, acceptable.

3.0 ENVIRONMENTAL CONSIDEPATION This amendment involves changes in the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20.

We have determined that the amendment involves no significant increase in the amounts, and no sionificant chance 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 findina that this amendment involves no significant hazards consideration and there has been no public comment on such finding.'

Accordingly, this amendment meets the eligibility criteria for cateoorical 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

We have 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, and (2) 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: February 3, 1987 Principal Contributors:

E. F. Branagan, Jr.

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. kEFERENCES 1.

"Fuclear Incident at Three Mile Island - Supplement," IF Bulletin 79-05B, April 21,-1979.

2.

" Justification for Raisin 9 Setpoint for Reactor Trip on High Pressure,"

BAV-1890, September.1985.

3.

" Basis for Raising Arming Threshold for /nticipatory Reactor Trip on Turbine Trip," BAW-1093, October 1985.

4.

. Letter from D. M. Crutchfield, NPC to J. H. Tavlor, Babcock and Wilcox Company, April 22, 1986.

5.

Letter from D. M. Crutchfield, NPC to J. H. Taylor, Rabcock and Wilcox Company, April 25, 1986.

6.

" Clarification of TMI Action Plan Pequirenents," NUREG-0737, November 1980.

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