05000344/LER-1986-003, Forwards Clarification of Util Position Re 10CFR50.49 Compliance for 11 ECCS Valves Discussed in Rev 1 to LER 86-003,per 860905 Enforcement Conference & Concerns Raised in Insp Rept 50-344/86-32

Forwards Clarification of Util Position Re 10CFR50.49 Compliance for 11 ECCS Valves Discussed in Rev 1 to LER 86-003,per 860905 Enforcement Conference & Concerns Raised in Insp Rept 50-344/86-32

ML20214U677
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	09/11/1986
From:	Withers B PORTLAND GENERAL ELECTRIC CO.
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
TAC-42502, NUDOCS 8610010373
Download: ML20214U677 (6)

LER-2086-003, Forwards Clarification of Util Position Re 10CFR50.49 Compliance for 11 ECCS Valves Discussed in Rev 1 to LER 86-003,per 860905 Enforcement Conference & Concerns Raised in Insp Rept 50-344/86-32

Event date:
Report date:
3442086003R00 - NRC Website

text

-

MCEP.'E0 l: '.1

- ummmun Ihk E mE E3 Fi? E Rt 2: 19 f3 art D. Wahers Vce Presdert oEG;0,'; 1i , 7 September 11, 1986 Trojan Nuclear Plant Docket 50-344 License NPF-1 Mr. John B. Martin Regional Administrator, Region V U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek CA 94596-5368

Dear Sir:

TROJAN NUCLEAR PLANT PGE Position Regarding Compliance With 10 CFR 50.49 of Eleven ECCS Valves As requested during the Enforcement Conference on September 5, 1986, attached is PGE's position regarding compliance with 10 CFR 50.49 of the eleven ECCS valves discussed in LER 86-03, Revision 1. The question of compliance with 10 CFR 50.49 of these components was raised in Inspection Report No. 50-344/86-32.

We appreciate the opportunity to clarify our position on this matter and are ready to answer any questions you may have.

Sincerely, N. ._. ,

Bart D. Withers Vice President Nuclear Attachment I

l c: Mr. Lynn Frank, Director State of Oregon Department of Energy 8610010373 860911 4

PDR ADOCK 0500 121 S W Spren Shot. Pecad. Owgan 97204

Trojan Nuclear Plant Mr. John B. Martin Docket 50-344 September 11, 1986 License NPF-1 Attachment Page 1 of 5 PGE POSITION REGARDING COMPLIANCE WITH 10 CFR 50.49 0F ECCS VALVES A. Introduction The following is provided to clarify PGE's position with regard to the environmental qualification (EQ) problems discussed in LER 86-03 .

Rev. 1 (Reference 1) for 11 motor-operated valves in the Emergency Core Cooling System (ECCS). The following explanation also provides the basis for why these 11 valves meet 10 CFR 50.49 requirements.

The discussion below centers around two questions:

1. Were the 11 ECCS valves in compliance with 10 CFR 50.49 prior to the recently identlfled pump runout concerns?

l 2. Are the 11 ECCS valves now in compliance with 10 CFR 50.49 with the procedure changes that.have been made to resolve pump runout concerns?

B. Were the 11 ECCS Valves Previously in Compliance With 10 CFR 50.49?

It is PGE's position that the 11 ECCS valves were previously in com-pliance with 10 CFR 50.49 up until the time procedure changes were implemented to resolve newly identified pump runout concerns (as described in Reference 1). The basis for this position is discussed.

in the following and does not include pump runout considerations because this was not known at the time of the original EQ evaluation (Reference 2).

The 11 valves in question are relied upon for transition from cold leg injection to cold les recirculation, and for transition from cold leg recirculation to hot leg recircuiation. The valves are also used to isolate passive failures. In evaluating accomplishment of the recir-culation functions, a degraded core condition was assumed in the original EQ evaluation per NRC EQ guidelines. However, a degraded core condition was not assumed for the passive failure function since j this conflicted with the FSAR accident analysis and was not explicitly required by NRC requirements or EQ guidelines. Each of these cases

+

are addressed separately below.

For the transition from cold leg injection to cold leg recirculation, the original EQ evaluation assumed a brake failure does not occur because the equipment would receive no more than 105 rads during this period (available vendor data was reviewed to ensure no radiation i sensitive materials such as Teflon were present in the brake construc-tion). Hence, the valves needed to operate are qualified to perform

Trojan Nuclear Plant Mr. John B. Martin

~ Docket 50-344 September 11, 1986 License NPF-1 Attachment Page 2 of 5 their function, since they perform this function prior to exposure to a significant radiation environment. There is also no single active failure during the transition from injection to cold les recirculation or during cold leg recirculation that would impair core cooling capability.

For the transition from cold les recirculation to hot leg recircula-tion, the original EQ evaluation recognized that operability of the RHR cold leg injection valves (MO-8809 A&B) and the RHR cross-tie valves (MO-8716 A&B) were necessary. It is also recognized that hot les injection could be accomplished via the safety injection (SI) pumps, and all valves necessary for this alignment were qualified.

Therefore, should the MO-8809s and MO-8716s fail as is (the evaluation assumed that brake failure causes the valve to remain in its pre-failure position) during hot les recirculation because of the radia-tion environment, hot leg injection could still be accomplished via i the qualified SI path. It was confirmed at the time that one SI pump i injecting to the hot legs from an RHR pump would meet mininum hot leg flow requirements. This conclusion is also valid with application of a single active failure during transition between the recirculation modes or during the hot les recirculation mode.

1 With respect to the passive failure isolation function these valves perform, the 11 ECCS valves are not required to isolate a leak under degraded core conditions for several reasons. First, 10 CFR 100 requires assumption of a fission product release from the core (substantial meltdown of the core) with an " expected demonstrable leak rate from the containment". No other assumptions are required with

[ respect to passive failures in the ECCS outside Containment. Second,

! 10 CFR 50 Appendix A requires assumption of a passive failure in the ECCS; however, no leak rate or core release assumptions are speci-fled. Thirdly, 10 CFR 50.49 contains no requirement for assuming a passive failure in conjunction with a design basis event.

The accident analysis in Section 15.6 of the UFSAR (same as the original) performs the offsite dose calculation, assuming one train of ECCS leakage at the maximum operational leakage rate. The resulting offsite doses are well below 10 CFR 100 values. Trojan's original

(. licensing basis for offsite dose determination due to a leakage from the ECCS components outside Containment followed the guidance of Regulatory Guide 1.70, Rev. 1. Regulatory Guide 1.70, however, did not specify leakage rates to be assumed. Trojan's original design basis, therefore, specified one train of ECCS leaking at the maximum operational leakage. This leakage was assumed to be constant for the course of the accident as specified in the original FSAR Sec-tion 15.4.1.2.6 and the UFSAR, Section 15.6.5.6.1. Later NRC guidance in NUREC-0800 (July 1981) specified that for a plant that does not provide an ESF atmospheric filtration system, the dose assessment

Trojan Nuclear Plant Mr. John B. Martin Docket 50-344 September 11, 1986 License NPF-1 Atta:hment Page 3 of 5 should also include the leakage from a gross failure of a passive component. This leakage should conservatively be assumed to be 50 spm starting 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident, lasting for 30 minutes. This subsequent requirement exceeds Trojan's original design bases speci-fled in UFSAR Chapter 15.6 and no offsite doses were calculated based upon these assumptions.

UFSAR Section 6.3.3 indicates that offsite doses were calculated for a passive leak, as defined in Section 3.1, and the offsite doses do not

, exceed 10 CFR 100 guidelines. Section.3.1 defines a passive leak as i 50 gpm for 30 minutes. However, the statements in Section 6.3.3 have been in the FSAR since the original submittal and do not reflect the

! NRC requirements at the time Trojan was licensed, nor the accident

' analysis contained in Chapter 15.6. The. statements in Section 6.3.3 are not discussed in the NRC Safety Evaluation Report for Trojan.

This section of the FSAR was originally proposed by Westinghouse.

It is unknown why.Section 6.3.3 was not edited correctly to reflect

, Trojan's design.

! Application of the 50 spm criterion (for 30 minutes) in Plant design is limited to consideration of ECCS function to meet minimum core cooling requirements. The passive. failure analysis in the FSAR

(Table 6.3-15) demonstrates that the ECCS can sustain a single passive failure during the long-term phase and still retain an intact flow path to the core to supply sufficient flow to keep the core covered

! and assure the removal of decay heat. For the'50 spm passive failure criterion, there is no FSAR assumption concerning core source term, j nor is one required by 10 CFR 50 Appendix A.

The original EQ evaluation for the valves (Reference 2) recognized the correct FSAR design bases by not requiring valve function "for leakage isolation in the long-term recirculation mode for degraded core LOCA events since this function is currently undefined in the plant design bases and emergency procedures". Since the FSAR accident analysis is predicated on a degraded core source term and a maximum operational (unisolable) leak, no valve operation is required under the assumption of degraded core conditions. For isolation of larger leak rates up to 50 gpm, no degraded source term is assumed in the accident analysis and the valves are not subject to a harsh radiation environment.

10 CFR 50.49 defines the scope of equipment to be qualified and requires that equipment qualification be based on a degraded core source term. The equipment required to be qualified is that relied upon to remain functional during and following design basis events to ensure (a) the integrity of the reactor coolant pressure boundary, (b) the capability to shut down the reactor and maintain it in a safe shutdown condition, and (c) the capability to prevent or mitigate the

,s Trojan Nuclear Plant Mr. John B. Martin Docket 50-344 September 11, 1986 License NPF-1 Attachment Page 4 of 5 consequences of accidents that could result in potential offsite exposures comparable to 10 CFR 100 guidelines. Equipment installed at Trojan which is necessary to acccmplish these functions is qualified on the bases of a degraded core source condition. HowcVer, the 11 ECCS valves are not required to be qualified under a degraded core condition for isolating leaks up to 50 gym because 10 CFR 50.49 does not explicitly require a passive failure be assumed in conjunction with the design basis event (ie, a LOCA). Indeed, EQ of the valves becomes a moot point since application of such a criterion in Plant design (ie, a 50 spm leak for 30 minutes with a degraded. core source tecm) would dictate the need for new equipment (ESF filtration sys-tems) to meet 10 CFR 100 guidelines. This goes beyond the scope of the EQ rule, which does not define safety functions for systems and components beyond those defined in the FSAR and require new systems to nieet these functions.

C. Are ths 11 ECCS Valves Currently in Compliance With 10 CFR 50.497 It is PCE's position that the 11 ECCS valves are currently in com-pliance with 10 CFR 50.49 in light of the procedure changes that have been implemented to resolve pump runout concerns.

This is becausa the Event Specific Emergency Procedures (ES-1,3 and ES-1.4) for cold and hot les recirculation have been revised to reposition the necessary ECCS valves prior to, or immediately after the onset of sump recirculation, before a significant radiation environment occurs to assure that none of the 11 ECCS valves have to be repositioned to eliminate pump runout potential during recirculation modes.

As noted in Reference 1, a procedure change is being evaluated to either ensure simultaneous hot and cold leg recirculation are main-tained or to address the recently identified requirement to alternate between cold and hot leg recirculation every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In conjunction with this change, the current procedure requirement for splitting the ECCS trains with the MO-8923 valvos is also being re-evaluated. This change would leave the MO-8923 valves open during all recirculation modes, and would rely on the MO-8809 valves for preventing pump runout l

(either MO-8809 A or B would be closed during the switchover from cold j leg injection to sump recirculation). This change is preferred over the current procedures in that it would provide maximum ECCS operating flexibility. Both procedure changes described are consistent with the discussion above in Paragraph B. Moreover, a material analysis per-i l formed on the 11 ECCS valves indicates they would most likely perform

( an isolation function in a high radiation environment.

J.

Irrespective of the foregoing, in order to ensure maximum ECCS

[

capability during recirculation modes, and also to conservatively

Trojan Nuclear Plant Mr. John B. Martin Docket 50-344 September 11, 1986 License NPF-1 Attachment Page 5 of 5 ensure that passive failure isolation capability is assured for all conceivable accident scenarios, the 11 ECCS valves will be qualified for degraded core radiation conditions during the 1987 refueling outage.

D. References

1. Licensee Event Report 86-03, Rev. 1, dated August 29, 1986.

2. PGE internal memorandum, " Environmental Qualification Evaluation of Limitorque Motor Operator Brakes", JWL-659-81M, dated September 21, 1981.

i i

DRS/sf 0760P