BECO-91-136, Application for Amend to License DPR-35,changing Tech Spec 3.5.D.2, RCIC Sys,Under Exigent Circumstances to Provide Limited Extension of 7-day LCO

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Application for Amend to License DPR-35,changing Tech Spec 3.5.D.2, RCIC Sys,Under Exigent Circumstances to Provide Limited Extension of 7-day LCO
ML20085L511
Person / Time
Site: Pilgrim
Issue date: 10/24/1991
From: Gina Davis
BOSTON EDISON CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20085L516 List:
References
BECO-91-136, NUDOCS 9111040196
Download: ML20085L511 (5)


Text

10CFR50.90 l$

BOSTON EDISON j

Pdgnm Nuclear Power Station j

Rocky Hill Road i

ri mouth, Massachusetts 02360 y

i october 24, 1991 George W. Davis sene vice ermoent ~ Nuaea' BECo 91-136 U.S. Nuclear Regulatory Commission Document Control Desk Hashington, DC 20555 License DPR-35 Docket 50-293 Proposed Change to Technical Specification 3.5.D.2, Reactor Core Iso _13_ tion Coolina System. Under Exiaent Circumit s gi In accordance with 10CFR50.90, Boston Edison proposes the following changes to Appendix A of Operating License DPR-35. The proposed changes revise the Pilgrim Nuclear Power Station (PNPS) Technical Specifications (TS) to provide a temporary extension to the 7-day limiting condition for operation (LCO) of TS 3.5.D.2 as discussed in our letter dated October 15, 1991.

The NRC granted Boston Edison a temporary waiver of compliance pending review of these proposed changes to preclude an unnecessary shutdown of PNPS.

Boston Edison requests approval of this proposed change under exigent circumstances.

Descrintion of Chmti:

The proposed changes to PNPS TS consists of a footnote providing a limited extension of the 7-day LCO of TS 3.5.0.2 for Reactor Core Isolation Cooling (RCIC) inoperability due to the potential for an undesirable DC bus voltage transient-induced trip of the RCIC inverter.

The extended LCO is limited to 97 l

days or until modifications can be completed or testing conducted to verify DC l

bus voltage transients will not exceed the RCIC inverter trip setting during a loss of coolant accident (LOCA) coincident with a loss of offsite power (LOOP),

whichever occurs first.

These proposed changes are shown on revised TS Page l

108 in Attachment B.

l l

Reason for Chance:

Boston Edison has determined for a LOCA coincident with a LOOP, the RCIC inverter may be subject to a momentary voltage transient on the supply DC bus i

when the DC bus is powered by its normal battery charger.

The transient may be sufficiently high to cause an inverter trip.

Presently, the backup battery charger powers the RCIC inverter.

Extrapolated test data from two similar battery chargers indicates the RCIC inverter trip setpoint would not be reached with +.he backty charger in servie for the LOCA with a LOOP scenario.

However, since suf ficient test data for the battery charger in question was not available to demonstrate the extrapolation was conservative and sufficient margin to the trip setpoint existed, RCIC was declared inoperable on October 9, 1991.

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BOSTON EDISON COMPANY U. S. Nuclear Regulatory Commission Page 2 of 3 During a LOCA with a LOOP, the "A" Core Spray Pump will start approximately one-third of a second after the emergency diesel generator circuit breaker closes onto 4160V Emergency Bus AS.

The battery charger powered from Emergency Bus A5 will energize at approximately the same time Emergency Bus A5 experiences a voltage transient due to the Core Spray Pump motor start.

This proposed TS change extends the 7-day ICO to allow the design and installation of a modification to prevent inverter trips.

Exclanation of Exiaent Circumstances:

The Code of Federal Regulations [10CFRSC.91(a)(6)] requires licensees requesting the issuance of a license amendment without the normal 30-day public notice to provide an explanation of the exigent circumstances.

As requested in our letter dated October 15, 1991, Boston Edison was granted a temporary waiver of compliance for TS 3.5.D.2 pending the NRC's review of this proposed TS change.

The circumstances surrounding the need for this change are fully described in the October 15, 1991 letter.

Boston Edison entered the present LCO on October 9, 1991.

Because the need for the change was not known to us prior to this time, the exigent circumstances could not have been avoided.

Boston Edison has made a good faith effort to prepare this request and submit it for NRC approval as expeditiously as practicable while adhering to the appropriate requirements for completeness and accuracy. Attachment A to this letter provides our determination of no significant hazards. Attachment B provides the proposed TS change.

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Attachments:

A) Determination of No Significant Hazards Consideration B) Proposed Technical Specification Page 108 Commonwealth of Massachusetts) l County of Plymouth

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l Then personally appeared before me, George H. Davis, who being duly sworn, did state that he is Senior Vice President - Nuclear of Boston Edison Company and that he is duly authcrized to execute and file the submittal contained herein in the name and on behalf of Boston Edison Company and that the statements in said submittal are true to the best of his knowledge and belief.

My commission expire :

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DATE NOTARi/UBLIC [

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BOSTON EDISON COMPANY O. S. Nuclear Regulatory Cornission Pagt 3 of 3 cc:

Mr. R. Eato1, Project Manager Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Mail Stop:

1401 U. S. Nuclear Regulatory Congnission 1 White Flint Ncrth 11555 Rockville Pike Rockville,-HD 20852 U. S._ Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406 Senior NRC Resident inspector Pilgrim Nuclear Power Station Mr. Robert H. Hallisey, Director Radiation Control Program Massachusetts Department of Pub.ic Health 30S South Street Jamaica Plain, MA 02130

ATTACHMENT A TO BECo 91-136 ihLtEDdnat10!Lof No SignificAntjfAzarALCon11hrdlan The Code of federal Regulations (10CFR50.91) requires a licensee requesting a license amendment to provide an analysis, using the standards in 10CfR50.92, that determines whether a significant hazards consideration exists.

The following analysis is provided in accordance with 10CFR50.91 and 10CFR50.92 for this proposed Technical Specification change.

1.

The operation of the Pilgrim Nuclear Power Station (PNPS) in accordance with the proposed amendment will not involve a significant increase in the prcbability or consequences of an accident previoc;1y evaluated.

This condition does no* i rease the probability of occurrence of an t

accident previously en

'd in the safety analysis report because accident initiators are a.pacted.

This condition does not increase the consequences of an accident previousiy i

evaluated in the safety analysis report because the Reactor Core Isolation Cooling (RCIC) System will perform its mission for all design basis events for which it is credited with performing a safety function.

The RCIC system safety objective is to provide makeup water to the reactor vessel following reactor vessel isolation to prevent the release of radioactive materials to the environs as a result of inadequate core cooling (FSAR 4.7.1).

RCIC shall operate automatically to maintain I

sufficient coolant in the reactor vessel so the integrity of the radioactive material barrier is not compromisec.

RCIC is designed to cope with a control rod drop accident, a loss of feedwater flow transient, and a loss of offsite power (LOOP) transient (fSAR Appendix G).

Each of these t

events results in an isolated reactor vessel with no assumed breach of the pressure boundary.

Reacter water level will drop as a result of the initiating events followed by a " boil down" as the safety relief valves relieve on high pressure. RCIC is designed to automatically restore water level by providing flow in excess of the boiling rate.

Technical Specification 3.5.0 requires RCIC to be operable whenever there is irradiated fuel in the reactor vessel, reactor pressure is greater than ISO psig, and reactor coolant temperature is greater than 365'f.

RCIC is not a core standby cooling system (CSCS) and is not credited in accident analyses for coping with any loss of coolant accident (LOCA) events.

The condition may increase the probability of malfunction of the RCIC during a LOCA coincident with a LOOP; however, RCIC is not credited in the safety analysis for LOCA events.

2.

The operation f PNPS in accordance with the proposed amendment will not create the possibility of a new or dif ferent kind of accident from any accident previously evaluated.

Page 1 of 2

High DC voltage induced trips of the RCIC inverter result in the RCIC turbine going to its minimum speed condition.

This creates no increased potential for intersystem LOCA or any other event because pump discharge check valve 1301-50 and turbine discharge pressure prevent back-leakage from the RPV by design. On a RCIC inverter trip, RCIC flow will decrease, valve 1305-50 will close to isolate reactor pressure and the minimum flow valve will open with the turbine at minimum speed.

3.

The operation of PHPS in accordance with the proposed amendment will not involve a significant reduction in the margin of safety.

Technical Specification Bases 3.5.C states RCIC is iequired as an alternative source of makeup water to the high pressure coolant injection (HPCI) system in the case of loss of all offsite AC power.

Technical Sp '.ification Bases 3.5.D confirms this function and further states for all other postulated accidents and transients, the automatic depressurization system (ADS) provides redundancy for the HPCI.

Since RCIC remains operable for a LOOP, the margin of safety is not significantly reduced.

Therefore, this proposed license amendment does not involve a significant hazard consideration.

Page 2 of 2

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