ML20207F816

Safety Evaluation Supporting Amends 65 & 54 to Licenses NPF-10 & NPF-15,respectively

ML20207F816
Person / Time
Site:	San Onofre
Issue date:	08/10/1988
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20207F807	List: ML20207F802 ML20207F816 ML20207F824
References
TAC-68425, TAC-68426, NUDOCS 8808230229
Download: ML20207F816 (3)

Text

.

/pnmatou

~

fg UNITED STATES 8

NUCLEAR REGULATORY COMMISSION g

D aj WASHINGTON. D. C. 20555

"%...../

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPOR11NG AMENDMENT NO. 65 r; FACILITY OPERATING LICENSE NO. NPF-10 AND AMENDMENT N0. 54 TO FACILITY OPERATING LICENSE NO. NPF 15 SOUTHERN CALIFORNIA EDISON COMPANY, ET AL.

SM ONOFRE NUCLEAR GENERATING STATION UNITS 2 AND 3 DOCKET NOS. 50-361 AND 50-362

1.0 INTRODUCTION

By letter dated June 14, 1988, 1988, Southern California Edison Company (SCE), et al.as.uppleme 13, and July 25:

licensees) submitted a request to revise Technical Specification 3/4.1.3.4 "CEA Drop Time," for the San Onofre Nuclear Generating Station (SONGS)

Units 2 and 3.

The change would increase the time requirement for insertion of control element assemblies (CEA's) upon receipt of a reactor trip signal from 3.0 seconds to 3.2 seconds. Specifically, this represents the time from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position from the fully withdrawn position.

Previously each CEA was withdrawn from the core to its full out position and dropped by opening its individual circuit braaker.

Beginning with Unit 2 Cycle 4 startup, a new method of measuring CEA drop times was used in which the reactor trip breakers are the point at which power is inter-rupted to the CEA gripper coils rather than the individual breakers. This new method uses the reactor trip breakers and, therefore, more accurately reflects the operation of the reactor protection system during a scram.

During the Unit 2 startup tests, the CEA drop times obtained using the new method were longer than those obtained using the previous method. This has been found to be due to the fact that the circuit dissipating gripper coil stored energy has a longer time constant when tripped by the reactor trip breakers than when tripped by the individual circuit breakers.

Unit 3 is currently in a refueling outage and will use the new test method during Cycle 4 startup. Since a review of previous Unit 3 CEA drop time measure-ments indicated that there is a potential for at least one CEA to fail to meet the 3.0 seconds requirement, the proposed change would increase the specified drop time to 3.2 seconds in order to avoid possible delay of Unit 3 restart.

8808230229 880810 PDR ADOCK 05000361 P

PNU;

o 2-2.0 EVALUATION The staff ha. reviewed SCE's reevaluation of chose Chapter 15 Design Basis Accidents which could be adversely impacted by the increased CEA scram time. The reevaluation of these events incorporated the increased CEA drop times in a conservative manner. Also incorporated was a revised CEA reactivity versus aosition curve bases on space-time neutron kinetics calculations for tiose events which previously used static calculations.

The staff has previously approved this methodology to determine CEA scram characteristics for SONGS as well as for other Combustion Engineering plants. The reevaluation has shown that, for most events, this revised scram reactivity prediction is conservative relative to the reference analysis scram reactivity dats at the crucial time in the transient (during the closest approach to a safety limit), or sufficient conserva-tisms exist in the analyses to offset the effects of the increased CEA drop time on the consequences of previously &nalyzed accidents.

For the most severe increased main steam flow event, which is the inadver-tent opening of all of the turbine bypass valves at full power, a low DNBR trip is generated by the Core Protection Calculators (CPC's) to terminate thermal margin degradation. The additional 0.3 second delay before CEA motion begins would result in additional margin degradation. To compensate for this, the licensee has proposed to increase the CPC DNBR power uncertainty penalty addressable constant labeled BERR1 in the CPC algorithms to 1.005. Adjustments to the value of addressable constants by SCE, without prior NRC approval, is permitted by Technical Specification 6.8.1.j, provided the new value is within the software limit values. This is the case with thi: adjustment. This correction factor effectively provides a CPC low DNBR trip early enough to compensate for an increased holding 0011 decay time of up to 0.3 seconds.

For the uncontrolled CEA withdrawal event from a subcritical condition, an additional 0.3 second holding : oil decay time results in increases in peak core power from 65% to 106% of rated thermal power and peak heat flux from 16% to 33%. However, the staff's acceptance criteria remain satisfied since milimum DNBR remains greater than 1.31 and fuel centerline melting does not occur.

The increase in CEA drop time results in an increase of less than 2% in rtquired margin to accommodate a total loss of forced reactor coolant flow and less than 1% to accomodate the pre-trip power excursion during a steam line break.

The licensee will, therefore, increase the Cycle 4 Core Operating Limit Supervisory System (COLSS) required overpower margins accordingly.

The licensee has verified that suf't :ent margin exists in the current TS limits when COLSS is out of serv 6 o accomodate the increased delay time so that no change to this 14 4 is required.

h -

3.0 CONCLUSION

S The staff finds the proposed increase in CEA drop time acceptable based on the above evaluation of its effect on the reference safety analyses.

These analyses either remain bour, ding or continue to satisfy the staff's acceptance criteria provided that the adjustment to the BERR1 addressable constant and the increase in COLSS inargins discussed above are made. The staff recomends that the Bases for Specification 3.1.3.4 h modified to reflect these required changes.

4.0 CONTACT WITH STATE OFFICIAL The NRC staff also advised the Chief of the Radiological Health Branch, State Department of Health Services State of California, of the proposed determination of no significant hazardu consideration. No coments were received from the State or from the public.

~

5.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact has been prepared and published in the Federal Register (53 FR 29971, August 9,1988). Accordingly, based upon the environmental assessment, the Comission has determined that the issuance of this amedment will not have a significant effect on the quality of the human environment.

6.0 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 Comission's regulations, and the issuance of these c.nendments will not be inimical to the comon defense and security or to the health and safety of the public.

Principal Contributor:

L. Kopp l

l Dated:

August 10, 1988 l

l l

l l

l