ML20133G662

Forwards Endorsement of Conclusions of NEDC-30844,BWR Owners Group Response to Item 4.5.3 of Generic Ltr 83-28 Re Surveillance Requirements for Reactor Protection Sys

ML20133G662
Person / Time
Site:	LaSalle
Issue date:	08/06/1985
From:	George Alexander COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
0476K, 476K, GL-83-28, NUDOCS 8508090024
Download: ML20133G662 (3)

Text

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\ Commonwealth Edison

~/ One First N;tional Plaza. Chictgo, Ithnois j[,~ O '] Address Reply to: Post Office Box 767

/ . Chicago, Illinois 60690 August 6, 1985

• Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory:. Commission Washington, DC 20555

Subject:

LaSalle County Station Units 1 and 2 Response to Request for Additional Information Generic Letter No. 83-28 NRC Docket Nos. 50-373/374 Reference-(a): March 1985 A.-Schwencer letter to D. L. Farrar (b): May 16, 1985 G. L. Alexander letter to H. R. Denton

Dear Mr. Denton:

Reference (a) requested additional information pertaining to Generic Letter 83-28 Items 2.2.2 and 4.5.3. Reference (b) responded to that request, but stated that the company had not completed our review of NEDC-30844 which was the BWR Owner's Group response to Item 4.5.3.

Attached is our response endorsing-the conclusions of the study and presenting our arguments on why_ Tech Spec surveillance requirements are not needed.

Please address any questions that you or your staff may have.concerning our response to this office.

One signed original with Attachment and fifteen, copies are being provided for your use.

• Respectfully, Greg lexand r Nuclear Licensing Administrator bs Attachment' O

cc: RIII Inspector- LSC i 0476K 8500090024 850006 PDR ADOCK 05000373 p PDR

ATTACHMENT During January 1985 the BWR Owners Group submitted NEDC-30844 as a response on the reliability basis for existing test intervals and allowable out-of-service times for the reactor protection system (RPS) of the standard BWR product lines. That generic report addressed the availability of RPS, the uncertainties , reduced redundancy during test or repair, operator errors and component wear out. The RPS reliability model was included in that report; it utilized PRA techniques to represent the function of the equipment, its surveillance testing for Tech Spec compliance, a'nd its operability features while undergoing maintenance or repair.

As prior guidance in this general topical area, the NRC Staff released NUREG-1024, " Technical Specifications - Enhancing the Safety Impact", which recommended that surveillance test requirements and Technical Specification Action statements be reviewed to assure that they have an adequate technical basis and that they do indeed minimize risk. The use of reliability analyses was recognized by the staff in a March 1984 meeting with the BWROG representatives who outlined the additional implementing generic studies beyond the concerns of generic letter 83-28. The common methodology tied these efforts together; detailed evaluations were planned for RPS and ECCS equipments.

One of these implementing reports, NEDC-30851P (May 1985) was submitted to NRC's C.O. Thomas from J.M. Fulton (BWROG chairman) on 31 May 1985. This implementing report provides the results of reliability anhlyses to justify extended test intervals and extended out-of-service times for the RPS technical specifications. Risk is thereby reduced because unnecessary scrams are decreased, excess test cycles are eliminated to conserve cyclic wear on equipment, and plant personnel and resources are conserved by less diversion and radiation exposure.

Specifically, current weekly and monthly sensor channel functional tests can be extended to quarterly tests; allowable repair and test times of 1 and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for RPS sensor channels can be extended to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> respectively. The net change in significant hazards as indicated by core damage frequency is a decrease by one percent.

Plant capacity factor increase is roughly one-tenth of that. These Tech Spec changes are supported by the analytical basis contained in this report. No online tripping of the reactor is anticipated now that start up testing is completed.

To make these generic reports applicable to any particular plant, each utility was to provide a verification report with plant specific interpretations of the generic analysis. For LaSalle, the plant unique evaluation is reported in MDE-83-0485 (April 1985) which is also a GE proprietary document. Correct informational references are identified there in and formal checklists used to assure comprehensive coverage of plant differences from the generic model used in NEDC-30844 and NEDC-30851P. Appendix A of the report gives an assessment of the effect on reliability (probability of damaging event) resulting from these differences. Edison agrees that this safety evaluation of

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the LaSalle RPS represents the plant and that the conclusions are pertinent for LaSalle. The results indicate that the LaSalle RPS conclusions do not significantly deviate from the RPS safety conclusions represented in the generic reports. Therefore, these generic analyses are applicable to LaSalle.

Specific comments were requested on Tech Spec changes which relate to the back-up scram solenoids. Edison's position is that these l devices are strictly a prudent back-up for the primary scram solenoids of the RPS. In fact, other secondary scram air header venting solenoids are a part of the Alternate Rod Insertion system (ARI) which is part of the LaSalle ATWS-2A modification. The ARI is totally ,

independent of RPS. It is a Class IE, seismically qualified system l which incorporates the recirc pump trip (RPT) and the scram discharge '

volume up-grade.. The LaSalle ARI is a power-on scram system that l operates similar to the backup scram solenoids of the RPS except with I entirely independent power sources (batteries).

An additional back-up reactivity control system, designed for maintenance purposes but useful for operational transients with extreme ,

coincident failures, is the Standby Liquid Control system (SBLC).

This manually operated equipment is strictly a back-up to RPS and ARI to cover the failure of multiple control rods to mechanically insert into the core. SBLC is covered by the Tech Spec for that reason.

NEDC-30844 Appendix H concludes that the contribution of the backup scram solenoids to overall failure to scram is trivial. Therefore,

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it appears inconsistent to require Tech Spec monitoring and reporting i of these backup devices. Edison's position is that the combined I reliability of the RPS and the ARI makes the ATWS event extremely remote for electrically based (or logic) faults and, further, that j the already vanishingly small probability for multiple rod mechanical failures negates any measureable benefit from applying Tech Spec coverage to these backup scram solenoids. Absent any benefit, such Tech Spec coverage is a needless burden.

To summarize: Edison endorses the two NEDC generic evaluations and the LaSalle specific evaluation of RPS reliability as reported above. Moreover, Edison considers these reports as adequate justification for the relaxing a Tech Specs surveillance intervals and out-of-service times in the LCO's for RPS. Edison also agrees that these evaluations justify the exclusion of Tech Spec coverage on the back-up scram solenoids, especially with the inclusion of the ARI modifications which respond to the new ATWS rule. On-line reactor trips are not contemplated nor needed to authenticate the RPS and scram capability which is authenticated via the Tech Specs as exercised on the scheGule extentions established by these studies.

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