ML20154A778

Safety Evaluation Supporting Amends 112 & 110 to Licenses DPR-39 & DPR-48,respectively

ML20154A778
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	04/15/1988
From:	Office of Nuclear Reactor Regulation
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ML20154A754	List: ML20154A750 ML20154A767 ML20154A778
References
NUDOCS 8805160155
Download: ML20154A778 (3)

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UNITED STATES y

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WASHING TON, D. C. 20555

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HFETYEVALUATIONBYTHEOFFICEOFNUCLEARREACTORREGULATION RELATED TO AMENDMENT NO. 112 TO FACILITY OPERATING LICENSE NO. OPR-39 AND AMENDMENT NO.101 TO FACILITY OPERATING LICENSE NO. OPR-48 COMMONWEALTH EDIS0N COMPANY ZION NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET N05. 50-295 AND 50-304 1.0 _ BACKGROUND In October 1975, the Reactor Safety Study (WASH-1400) identified the potential intersystem loss of coolant accident (LOCA) in a pressurized water reactor as a significant contributor to risk resulting from core melt.

In this scenario, check valves in the safety injection systems J,.

could fail allowing the high pressure reactor coolant to connunicate with the low-pressure piping outside of containment. Rupture of the low-pressure system would result in a LOCA outside containment, at the b

sametimedisablingthatpartofemergencycorecoolingsystem(ECCS) needed to mitigate the effects of the LOCA, and possibly lead to core melt.

Due to the safety significance of these check valves, a confirmatory order b

was issued in April 1981, to require that operating reactor technical specifications (TS) be modified to impose limiting conditions for operations (LCO) and surveillance requirements (SR) on the pressure isolation valves (PIVs). On March 13, 1987, Generic Letter 87-06 was issued to all holders of operating licenses. The purpose of the letter was to obtain information from the licensees on the implementation status of their PIVs test programs.

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• cent NRC review of the PlV test procedures at the Zion Nuclear Station and i hn's responses to the Generic Letter, GL 87-06 indicated that the Zion TS end SR do not reflect NRC requireinnts on PIVs.

In response to this finding, Commonwealth Edison Company s1bmitted a letter dated November 13, 1987, in which changes were prc.cosed to TS-3.3.3.F and SR-4.3.3.F to include PIV test requirements.

The staff's evaluation of the proposed TS changes is provided below.

2.0 EVALUATION General Design Criterion (GDC) 14 of 10 CFR 50, Appendix A requires that i

"the reactor coolant pressure boundary shall be designed, fabricated, erected and tested so as to have an extremely low probability of abnonnal leakage, of rapidly propagating failure and of gross rupture." Further-more,10 CFR 50.55a and Subsection IWV of the ASME Code,Section XI, require that certain valves needed for shutting down a reactor or in mitigating an accident be leak rate tested, 8805160155 880415 i

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l O PIVs are defined for each interface as any two valves in series within i

the reactor coolant pressure boundary which separate the high pressure reactor coolant system from an attached low pressure system. There are essentially two types of PIVs. The first type is PlVs whose sizes are small or that are not connected to any accident mitigation systems. The second type is PIVs whose sizes are of some significance and are connected to an accident mitigation system.

Failure of the first type of PIVs would result in a LOCA that should be within the design basis accident and should have no impact on the nomal plant's ability to shut duwn the reactor or to mitigate the consequence of the accident. However, the failure of second type of PIVs would not only result in a LOCA that is beyond the plant's ability to achieve a normal shutdown of the reactor but also disable part of the mitigation system needed to mitigate the LOCA. Therefore, gross failure of the second type of PIVs would result in an accident that might be beyond the design basis accident and might ultimately lead to core nelt.

In order to comply with the re Section XI inservice testing (quirerents of GDC 14, the staff imposes ASME IST) requirements on the second type of PIVs in accordance with 10 CFR 50.55a because their failure would effect the

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plant's ability to shut down the reactor and to mitigate the consequence f

of an accident.

7 In the proposed TS changes for Zion Station, the licensee has included in TS-3.3.3.F all PIVs that separate the primary coolant system from the safety injection system, residual heat removal system and the accumulators. The licensee proposes to leak test all PIVs under SR-4.3.3.F by meeting the

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following conditions:

l a.

At least once each REFUELING CYCLE, b.

Prior to entering MODE 2 whenever the plant has been in MODE 5 for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if leakage testing has not been performed in the previous 9 months, i

Prior to returning the valve to service following maintenance, repair c.

or replacement work on the valve, and d.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or prior to entering MODE 2, following valve actuation where flow through any of the PIVs results in water being injected into the ReNtcr Conlant System, test those valves that have experienced flow.

(ExcludingM0Y-RH8701andMOV-RH8702).

The license has also established allowable limits of leakage rate and leakage rate changes.

The staff has reviewed the proposed list of PIVs against Zion designs and P& ids, and has found that it is complete.

The proposed testing conditions under SR-4.3.3.F meet NRC staff positions and exceed ASME Section XI require-Zion TS-1.40 defines that "when REFUELING CYCLE is used to designate ments.

a surveillance interval, the surveillance shall be perfomed at least l

once every 18 months". Testing PIVs once every 18 months meets the Section XI requirement of leak rate testing at least once every 2 years. The staff finds SR-4.3.3 F.a acceptable. The proposed allowable leakage rate

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. change of less than 50% of the margin between measured and maximum leakage rate meets the Section XI requirements and is also acceptable.

3.0 TECHNICAL FINDING Based on the above discussion, the staff finds that the proposea TS changes of TS-3.3.3.F are acceptable.

4.0 ENVIRONMENTAL CONSIDERATION

These amendments involve a change in the installation or use of the facilities components located within the restricted areas as defined in 10 CFR 20. The staff has detemined that these amendments involve no significant increase in the amounts, 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 Comission ic has previously issued a Notice of Consideration of Issuance of Amendment L,

to Facility Operating License and Opportunity for Hearing dated December 16, N

1987 (52 FR 47807). No public comment has been received. On j

the Commission published an Environmental Assessment and Finding of No Signi-ja.

ficant Impact relating to these amendments.

5.0 CONCLUSION

We have concluded, based on the considerations discussed above, that:

h (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 regu-lations and the issuance of these amendments will not be inimical to the comon defense and security or the health and safety of the public.

r PRINCIPAL CONTRIBUTOR:

J. Huang h

Dated: April 15, 1988 I

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