W3P90-1528, Forwards Addl Info Re Tech Spec Change Request (Tscr) NPF-38-108 Concerning Removal of Automatic Closure Interlock Functional Surveillance from ECCS Ts,Per 901008 Telcon

Forwards Addl Info Re Tech Spec Change Request (Tscr) NPF-38-108 Concerning Removal of Automatic Closure Interlock Functional Surveillance from ECCS Ts,Per 901008 Telcon

ML20058H301
Person / Time
Site:	Waterford
Issue date:	11/07/1990
From:	Burski R ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
W3P90-1528, NUDOCS 9011150163
Download: ML20058H301 (5)

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Novcinber 7,1990 U.S. Nuclear Regulatory Commission j ATTN: Document Control Desk Washington, D.C. 20555 ,

Subject:

Waterford 3 SES  :

Docket No. 50-382 l' License No. NPF-38 Technical Specification Change Request (TSCR) NPF-38-108 Gentlemen: .

During the week of October 8,1990, several conversations occurred between .l representatives of Entergy Operations, Inc. and an NRR staff member i concerning an amendment request to remove the automatic closure interlock-functional surveillance from the Emergency Core Cooling Systems Technical t Specifications. Additional information was provided on several issues. This letter serves as documentation of this information. t P

Attachment A documents the issues and information discussed during 'the ';

conversations. If there are any questions concerning the responses, please ,

feel free to contact D. A. Rothrock on (504) 739-6693.  :

Very truly yours,

k. "

.f RFB/DAR/ssf .

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Attachment:

Additional Information Concerning NPF-38-108 '

l cc: Messrs. R.D. Martin, NRC Region IV' D.L. Wigginton, NRC-NRR E.L. Blake  :

R.B. McGehee Ms. 'M. Chatterton, NRC-NRR  ;

NRC Resident Inspectors Office ..

Administrator Nuclear Energy Division (State of Louisiana) .

American Nuclear Insurers e

901115o163 901107 ,

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' Attachment to l W3P90-1528 J Page 1 of 4 j 1

Attachment A . _ . .

Additional Information Concerning NPF-38-108

o Q Are there any design basis accidents dependent upon the functioning of the automatic closure interlock (ACI)?

A No.

o Q Are the alarms independent from the indicator on the panel in the control room?

A Yes.

o Q Are any changes.being made to the alarm system for the removal of the ACI function?

A No, the alarm will not be altered.

I o Q the Waterford 3 request, W3P90-0234, statesithat th4re is an-l alarm to notify the operator when the Shutdown Cooling System -

l (SDCS) suction valves are mispositioned.. In: the- event that an ,!

operator fails to completely close both suction valves and isolate l the SDCS when the primary system is becoming pressurized,-

l operating procedures will outline operator actions following ,

receipt of an alarm. 1 1

What are these actions?

A The Annunciator Response Procedure is under revision to reflect the removal of the ACI function. Included as a part of this revision will be guidance to the operators for. evaluation of an i alarm receipt. The operators will be instructed to discontinue.

pressurization, and close the isolation valves when they are not fully closed and pressure is increasing past the alarm setpoint. -

o Q Provide more information concerning-the probabilistic risk' analysis for Interfacing. System LOCA mentioned in- W3P90-0234.

A W3P90-0234 requests NRC . approval to- delete the surveillance requirements for the ACI on the SDCS suction valves.

Waterford 3 plans to remove the ACI to improve overall~ plant safety. In the letter, Waterford 3 presented the results.of an ,

'1 assessment performed to determine the effects of 'ACI removal-upon' Interfacing System Loss of Coolant. Accident (ISLOCA)'

frequencies. This result.is based upon comparison of two-configurations:

Att:chment to W3P90-1528 Page 2 of 4

1. SDCS suction valves with alarm and ACI:

ISLOCA frequency = 1.12

• 10-7 / year

2. SDCS suction valves with alarm only:

ISLOCA frequency = 1.12

• 10-7 / year The results show a negligible 0.09% increase in ISLOCA frequency due to removal of the ACI. They also show that there is a 39% decrease in both SDCS unavailability and low temperature overpressure protection (LTOP) unavailability with ACI removed. This implies a net increase in reactor safety.

With an alarm present, as in the current Waterford 3 design, ACI is a negligible contributor to reactor safety. The dominant contributor to ISLOCA frequency is a catastrophic failure of both SDCS suction isolation valves with the reactor at power. Neither ACI nor alarms can provide defense against such a failure, nor  ;

is that the intended function. Furthermore, the alarm offers -

protection against equipment failures (such as the SDCS isolation -

valves falling to closo); the ACI does not. -

At power, an ISLOCA via the SDCS suctiot lines can occur by  !

the following mechanisms:

i Both isolation valves in series are left open, il The motor operated valve is left open, and the hydraulic operated valve in series ruptures, iii The hydraulic operated valve is left open, and the motor operated valve in serleo "uptures, or iv Both valves rupture.

The first mecharism is not a credible initiator for an ISLOCA.

If both valves are left open during reactor startup, the SDCS relief valve, located downstream of the two valves, will open to relieve the increasing pressure and discharge reactor coohnt to the containment sump. The setpoint of the relief valve is 430 psia. Upon relief valve aatuation, indications of increasing containment sump level and decreasing reactor coolant system (RCS) volume control tank level will alert the operator that the RCS pressure boundary has not been secured during startup.

Dte to these indications and the affecta of the relief valve dischstge upon RCS pressure, startup will be suspended until  ;

the RCS pressure boundary is established by closing the SDCS 1 suction valves. Leaving both valves in series open is not credible. l I

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, Attachme t:toI W3P90-1528 '

Page 3 of 4i The frequency of an' ISLOCA related to the 'two SDCS suction

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lines at Waterford 3 'can then. be+ estimated based upon' the-remaining three mechanisms: .

l F(ISL) = 2' * (aQ 2 + aQ3 l + aQ3 )

where:-- ,

-l n

' o F(ISL) E . Frequency of. ISLOCA$via - SDCS :

suction lines a E: Catastrophic failure ra't e for motori 1 operated or' hydraulio operated - , ,

valves <

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Q1 E Probability'; hat motor operated 3' valve is ne' closed' Q 3' E- Probability that L hydraulic : operated . ,

valve is notEclosed; Q3 E P robaMilty/ that ; hydraulic : operated : j

':vs, 4 fails - given that .motorf '

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. opert ted . valve Lhas failed : 1 4i Two cases, with anu without ACI, were analyzed' for Waterford  !

3.  ;

3 Case 1: Alarm and ACIi and Case 2: Alarm only.

-l Waterford 3 letter W3P90-0234 discussed - the: results ' ofL Case 11 i which represents the current Waterfordj3. configuration! The l same alarm characteristics have b'een assumed for Casei 2?as  !

Case 1. -

I Variable Q 3 has the same value for. both Cases.1 TFault tree: ,

analyses are used to determine Q t and-:Q2 . The results are:' i 1

4 Case 1 Case Q2 1.00

• 10-* 21.10

• 10T*,

j Q2 2.40:* 10-7 3.= 38:: *- 107 7

Q3 2.04 *-10-' 2.04

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, Attachment toi $j W3P90-1528 : R

- Page 4 of!4-In all cases, Q 3 is the: dominant: term. This' term, representing catastrophic failure of both-initially closed valves;- contributes :

over 99% of- the total ISLOCA risk for! Cases 1- and' 2. The .

. o increases in variable Q 2 and Qi orf Case 2l compared to- Case'1 are rr.inor and quantify- the ffects of ACI removal, sin'ce; the -

same alarm characteristics are assumed? for. both cases. I m

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The results indicate a negligible ~ 0.09% difference in ISLOCA' ' +

probability for the two cases:.

cl

-l Case 1 F(ISL) = '1.1155

• 10-7 / year j Case 2 F(ISL) = 1.1165, *f10~7 /ye rf '

.. 1 As proposed- in W3P90-0234, ACI removal will result in:a net

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increaseiins reactor' safety, based on"the substantial decrease in Vj SDCS and. LTOP. unavailability .and .the : negligible impact- on - >J ISLOCA frequencies. ACI removal'is' being' pursued consistent.-

with the -guidelines recomtrended- by:.the .NRC'in a January 28,.

1985 memorandum from G.W.~ Sheroni C) af tof Reactor Systems ,

Branch.

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