ML19351A360

Application for Amend to License NPF-38,consisting of Tech Spec Change Request NPF-38-103,revising Table 4.3-1 of Tech Spec 3/4.3.1, Reactor Protective Instrumentation & Table 4.3-2 of Tech Spec 3/4.3.2, ESF Actuation Sys....

ML19351A360
Person / Time
Site:	Waterford
Issue date:	10/16/1989
From:	Dewease J LOUISIANA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19351A361	List: ML19351A360 ML19351A363
References
W3P88-1849, NUDOCS 8910200116
Download: ML19351A360 (7)

Text

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m I Louh Power & Ught Coy.pany 317 Barorve Street P.O. Boa 60M0 Now Orleans, LA 70160 0MO Tel 604 695 2781 J. G. Dewease Genum Vee Pepsedent-uaw ovaesm l W3P88-1849 A4.05 '

October 16, 1989 QA U.S. Nuclear Regulatory Commission ATTN! Document Control Desk Washington, D.C. 20555  ;

Subject:

Waterford 3 SES ,

Docket No. 50-382 License No. NPF-38 Technical Specification Change Request NPF-38-l_03_

Gentlement In accordance with 10 CFR Part 50.90 Louisiana Power & Light company i submits a request to amend Facility. Operating License NPF-38.

This minendment revises the Reactor Protective System (RPS) Inotrumentation Surveillance Requirements (Table 4.3-1 of Technical Specification 3/4.3.1

" Reactor Protective Instrumentation") and the Engineered Safety Feature i Actuation System.(ESTAS) Inntrumentation Surveillance Requirements (Table 4.3-2 of Technical Specification 3/4.3.2 " Engineered Safety Feature Actut4 tion System Instrumentation"). The proposed amendment changes the surveillance frequency for several of the channel functional tests from ,.

monthly to quarterly. The basis for this chenge comes from CEN-327: ,

RPS/ESFAS Extended Test Interval Evaluation, submitted by the Combustion  !

Engineering Owners Group.

Please feel free to direct any questions or comments on this matter to [

Steven E. Farkas at (504) 464-3383. l truly yours,

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> IO J. . Dewcase wu a-  :

S for Vice President clear Operations

,M JGD/SEF/pi [

! qsn Attachmentst NPF-38-103 '

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'0% cc Messrs R.D. Martin i M F.J. Hebdon I i

gN D.L. Wigginton E.L. Blake 1r W.M. Stevenson [

'g,h NRC Resident Inspectors Office I Administrator Nuclear Energy Division (State of Louisiana)

,American Nuclear Insurers

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UNITED STATES OF AMERICA NUCLF.AR REGULATORY COMMISSION l

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In the matter of )

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( Louisiana Power & Light Company ) Docket No. 50-382 ,

! Waterford 3 Stet.m Electric Station )

I AFFIDAVIT

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J.G. Dewease, being duly sworn, hereby deposes and says that he is Senior Vice President - Nuclear Operations of Louisiana Power 6 Light Companyl i that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached Technical Specification Change Request NPF-38-103;  !

that he is familiar with the content thereof and that the matters set  :

forth therein are true and cerrect to the best of his knowledge, information and belief. ,

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_ ! WC IJC.Dewtase senior Vice President

/ Nuclear Operations  !

STATE OF LOUISIANA) '

).ss PARIS!! 0F ORLEANS )

a Notary Publi 1. d or the Parish Subscribed and and named State above sworn to before this /dme,I,

/ day of ,

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1989. i b' ' 4 ~

otary Public" C

My Commission expires ,. .

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L DESCRIPTION AND SAFETY ANALYSIS r

OF PROPOSED CHANGE NPF-38-103 h

This document justifies a revision ~to Table 4.3-1 of Technical Specification 3/4.3.1: Reactor Protective Instrumentation and Table 4.3-2 g of Technical Specification 3/4.3.2 Engineered Safety Features Actuation i

System Instrumentation, t ..

Existing Specificatiens See Attachment A U Proposed Speeffications f

See Attachment B

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i ' Description, The operability of the Reactor Protective System (RPS) and Engineered Safety Features Actuation System (ESTAS) instrumentation and bypasses ensures thatt l . "- 1) the~ associated Reactor Trip and Engineered Safety Features Actuation

!> actions perform when the parameter monitored by each channel, or p- combination thereof, reaches its setpoint s2) the specified coincidence logic is maintained

3) redundancy permits a channel to be out-of-service for either testing or maintenance 1 4) sufficient system functional capability is available from diverse parameters.

ESTAS and RPS operability criteria assure the reliability, redundancy and diverst:y assumed in the facility design for the protection against, and 21cigation of, accident and transient conditions.

The surveillance requirements ensure the system capability stays comparable to the original design standards. The RPS and ESFAS instrumentation channel functional tests are performed monthly. These surveillances are time consuming. More importantly, they collectively throughout the industry cause many inadvertent reactor trips and engineered safeguards actuations challenging safety systems. Therefore, this high frequency results in an adverse impact on equipment life and unit availability. The proposed license amendment increases the interval for these surveillances from monthly to quarterly and maximites instrumentation reliability.

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The NRC recognized the burden imposed by some technical specifications I and is evaluating how to modify or restructure technical specifications to l reduce that burden without affecting the health and safety of the public.

In 1982. Westinghouse and the Westinghouse owners Group submitted a generic  !

j- topical report requesting test interval extensions for selected components l in the RPS and EST/.S. This request was supported through the use of ,

probabilistic risk assessment methodologies. As a result, the NRC l

foitiated a renarch project to determine how probabilistic risk assessment  ;

methodologies can lead to optimal technical specification surveillance  ;

intervals and allowed component outage times. Plus, the project evaluates  !

how to grant requests for technical specification relief based on such

, probabiliotic risk analyses. Pending completion of this research, the NRC ,

i indicated that they will favorably review technical specification relief requests based on probabilletic risk analyses, as long as these analyses demonstrate no significant increase in the risk to the public. .

[' Following the low power ATWS event in 1983, the NRC issued Generic Letter 83-28 specifying actions for licensees and applicants. One of these ,

actions reviews the Reactor Trip System (RTS) test intervals to deterutne i, if they help achieve high RTS availability. CE, under contract to the CE Owners Group.(CE00), performed an analysis on RTS availability given the thirty day test interval required by the Technical Specifications.

Sensitivity analyses indicated that the RTS unavailability is insensitive to changes in individual component failure rates. That conclusion affects the RTS test program rationale. Less testing reduces challenges to safety systems and improves plant availability. In January 1985, CE, under contract to the CEOG, began an analysis on the impact of extending the surveillance test intervals for selected components in the RPS and ESFAS.

This analysis became LEN-327: RPS/ESFAS Extended Test Interval 1. valuation.

In CEN-327, each of the four RPS fault tree models developed previously for the CEOG expanded to cover all RPS electronic trip parameters. The new models were then used to determine the RPS reliability for the current and the proposed test interval. The model takes into account common mode failures, operator errors, reduced redundancy, and random component failures. The analysis intends to justify extending the surveillance test intervals from thirty days to ninety days. Based on the NRC position at that time, the CE00 declared acceptable analysis results if extending the test intervals did not increase the risk to the public (as measured by the core damage frequency). Trip parameter circuits (between the sensors and actuated device) were analyzed assuming a ninety day test interval with staggered testing and a sixty day test interval with sequential testing.

Both of these test schemes passed the "no increase in core damage frequency" criteria.

Fault tree models were also constructed for each of the ESFAS signals.

l Similar to the RPS fault tree models, each ESFAS fault tree model specifically addressed common mode failures, operator errors, reduced l redundancy, and random component failures. Once the models were l constructed CE measured the ESTAS reliability assuming the current and proposed test intervals. For ESFAS, CEN-327, Section 5.e, specifically I recommended increasing the surveillance frequency for the bistables.

I bistable relays, logic matrix relays, actuation logic circuits and manual l- actuation devices. See the enclosed figure. However, CEN-327 holds that

!, the test interval for process measurement sensors and subgroup actuation I relays remain the way technical specifications currently require.

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CEOG submitted CEN-327 to the NRC for review in June 1986. The NRC l

,s contracted with EC&G Idaho to perform an er. tensive technical evaluation. l q EC&G Idahe then submitted their report to the NRC. The NRC then performed additional reviews and prepared a draft Safety Evaluation Report (SER). '

The SER concluded that extending the RPS and ESTAS test intervals was ,

acceptable, j The NRC reviewed CEN-327 and felt that changes in the RPS reliability that l result from extending the test interval from thirty days to ninety days I were acceptable. And, they were inclined to approve a ninety day RPS test j

, interval with sequential testing. Later, CE received additional feedback i from the NRC. While the NRC was inclined to find acceptable the RPS test '

i interval extention from thirty days to ninety days, with sequential testing, they could not include this in the SER for CEN-327 because CEN-327 ,

recommended more restrictive test intervals for certain trip parameters. 4 However, the NRC proposed to move forward on the 'M if CE issued a l i:upplement to CEN-327. The supplement to CEN-327 (sent to the NRC on March d 4 3, 1989) presents updated system reliability information for the RPS. And ,

it recommends a ninety day test interval with sequential testing for KPS. l L

The analysis results presented in CEN-327 and CEN-327 Supplement 1 ,

demonstrate that the surveillance test intervals for RpS and ESFAS

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components can be increased without increasing public risk. In fact, for  !

the test intervals propooed, the overall impact is a slight decrease in f public risk as measured by a net decrease in core melt frequency.  !

Extending the test interval does not change the trip per test frequency,  ;

but it does reduce the trip per year frequency, j The analyses and conclusions presented in both CEN-327 documents support this license amendment request to extend logic test frequencies as shown in i Attachment B. ESTAS actuation logic stays at the sequential monthly test j interval because the CEN-327 analysis did not include those relays. '

Supplement I showed the analysis applied to RPS from sensor (not including 1 the sensor) to CEDM breaker (not including the breakers). RPS logic can be tested through the final contact in the breaker opening logic. A similar .

test in ESFAS would start equipment in the plant.

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t This submittal closely follows the rationale provided in CEN-327. That ,

topical report has extensive applications to all CE plants. If the NRC  !

approves CEN-327, many plants will realize a significant safety benefit by ,

reducing the net core melt risk. Specifically, Waterford seeks to reduce its core melt risk by proposing this technical specification amendment. l l  :

Safety Analysis ,

1. Will operation of the facility in accordance with the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated? i Responset No.

Increasing the surveillance interval for RpS and ESTAS has two i principle effects with opposing impacts on core melt risk. The first l

< impact is a slight increase in core melt frequency that results from the increased unavailability of the instrumentation in i

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question. The unavailability of the tested instrumentation translates to a failure of the reactor to trip (AWS) or a failure of the l appropriate engineered safety feature to actuate when required. The l opposing impact is the corresponding reduction in core melt frequency  !

..that would result because of the reduced exposure to test induced  ;

transients.  ;

Representative fault tree models for Waterford and the corresponding j core melt frequency increases and decreases were quantified in CEN-327. The unavailability assumption described above includes the j

' increased relay service time (relays are normally energized). For the j specific relaxations proposed in Attachment B, an extended l surveillance interval was found to result in a net reduction in core  :

nelt risk. A lower potential for test induced trips over-shadows negative effects from increasing relay operating time.

Therefore, the proposed chsnge will not involve a significant increase i in the probability or consequences of any accident previously  !

evaluated. l i

L 2. Will operation of the Facility in accordance with this proposed change  !

create the possibility of a new or different kind of accident from any j accident previously evaluated?

Response: No.

This amendment request does not involve any changes in equipment and will not alter the manner in which the plant will be operated. For l this reasou, this, amendment will not create the possibility of a new i or different kind of accident from any previously evaluated.  ;

3. Will operation of the facility in accordance with this proposed change l involve a significant redaction in a margin of safety? ,

Response No.

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No margin of safety, based on the analyses of record and the existing  ;

r limits of the technical specifications, will be reduced. There are no changes to the equipment or plant operations that will result. The i only impact of this change, an described in detail above, is a ,

reasoned balance between two factors. One, increased core melt risk .

because of slightly increased equipment unavailability. And two, an [

even larger reduced core melt risk because of a reduction in plant

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exposure to test-induced trancients. The conclusions in the Waterford 3 FSAR remain valid and the safety limits continue to be met. Margins of safety are, therefore, not reduced.

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Safety and Stanificant Hagards Determination

%. l y,.', - Based.pn the above Safety Analysis, it is' concluded thatt (1) the proposed 9e '

change:does not constitute a significant hazards consideration as defined by 10 CFR 50.923 and (2) there is a reasonable assurance that the health 4:.' and safety'of,the public wil) not,be endangered by the proposed change and .i (3) this action will not result in a condition.which significantly alters  ;

'the impact of the station on the environment as described in the NRC Final Environmental Statement.' .

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