05000456/LER-1989-013-04, :on 891005,discrepancy W/Design of Steam Generator Blowdown Sys Identified,Minimizing Auxiliary Feedwater Flow Requirements.Caused by Preservice Design Deficiency.Temporary Design Changes Made

:on 891005,discrepancy W/Design of Steam Generator Blowdown Sys Identified,Minimizing Auxiliary Feedwater Flow Requirements.Caused by Preservice Design Deficiency.Temporary Design Changes Made

ML19354D471
Person / Time
Site:	Braidwood
Issue date:	11/03/1989
From:	Querio R, Thout J COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BW-89-2087, LER-89-013-04, LER-89-13-4, NUDOCS 8911090274
Download: ML19354D471 (5)

LER-1989-013, on 891005,discrepancy W/Design of Steam Generator Blowdown Sys Identified,Minimizing Auxiliary Feedwater Flow Requirements.Caused by Preservice Design Deficiency.Temporary Design Changes Made

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(1) 10 CFR 50.73(a)(2)(x)
4561989013R04 - NRC Website

text

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Commonwealth Edison.

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Braidwood Nucb r P wir St: tion N

RoutX1. Box 84 1

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Brac:vi;b. tilinois 60407.

Telephone 815/458-2801 November 3,1989 BW/89-2087 F

U. S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555

Dear Sir:

The enclosed Licensee Event Report from Braidwood Generating 1

Station is being transmitted to you in accordance with the requirements of 10CFR50.73(a)(2)(v) which requires a 30-day written report.

This report is number 89-013-00; Docket No. 50-456.

Very truly yours, t

. E. Querto Station Manager Braidwood Nuclear Station g

REQ / ADS /sjs (7126z).

Enclosure: Licensee Event Report No. 89-013-00 L

cc:

NRC Region 111 Administrator i

NRC Resident inspector INPO Record Center CECO Distribution List 3

4 9911090274 891103 TM PDR ADOCK 05000456 i \\

S PDC

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LICEWSEE EVfMT REPORT (LER)

Facility Name (1)

Docket Number (2)

_ Pane (3)

Jr114v.eell 1 0151 OLaLDL.allu J_l_.et}J_la_

Title (4) Inadequate incorporation of Isolation Requirements For the Steam Generator 81owdown System Ove to a Preservice Design Deficiency

.. Event.Date (5)

LER Nimber (61 Repgr.t J ate (7)

Other FACil11101 In10lYtdJEL

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Revision Month Day ~

Year Facility. East &_JosiaL!haptttris)

Sequential /j//j/j/

/j/j//j Mo!.th Day Year Year

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._.Jraidioed_L._. JULoLolalJL5L2

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_ai113 oIo ii1 of 3 819 0151oLotof 1 I-THl$ REPORT IS SUBMITTED PUR$UANT TO THE RE()UIREMENTS OF 10CFR OPNW LChei one or.nore of.the.followingl.(11) k I"

6 20.402(b) 20.405(c) 50.73(a)(2)(iv) 73.71(b)

POWER 20.405(a)(1)(1) 50.36(c)(1)

3. 50.73(a)(2)(v) 73.7)(c)

LEVEL 20.40$(a)(1)(ii) 50.36(c)(2) 50.73(a)(2)(vli)

Other ($pecify l0 20.405(a)(1)(li t )

50.73(a)(2)(1)

_., 50.73(a)(2)(vill)(A) in Abstract ital

=///////////////,/,////////,//_20.405(a)(1)(lv) 50.73(a)(2)(li)

_ 50.73(a)(2)(vili)(B) below and in

//////////////}/}////////}///

20.405(a)(1)(v)

__.50.73(a)(2)(lit) 50.73(a)(2)(x)

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(IC E$(E_ CONTACT F08 THIS LER (121 Name TE(EEljQHE NUPSER AREA CODE Jialbgyt. TechnIca1 Silfi lDSlattr ExL. 2489 8l115 4l5lBl_Lil3LoL1 E0t!ELLILONLutiLIDRleflLC0tlP9tfMIJAILURLDESLRIBID._1fLIlill_ REPORT (13)

CAUSE

SY$ TEM COMPONENT MANUFAC-REPORTABLE

CAUSE

SYSTEM COMPONENT MANUFAC.

REPORTABLE TVRER J0. HERD 1.

TURER

_10_.NERQ1.

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$UPPLEMEN?AL_RLEQRT EXPEETED (14]

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AB3 TRACT (Limit to 1400 spaces, i.e approximately fifteen single-space typewritten lines) (16)

On October 5,1989 a discrepancy with the design of the Steam Generator Blowdown (SD) system was identified.

The Safety Analysis Report (UFSAR) specifies $D isolation on the ir.itiation of the Auxl11ery Feedwater (AF) l System. This minimizes the AF flow requirements. The current design does not provide for a. stomatic l

1 solation on all AF Initiations. The $0 system mas isolated. Changes were made to Escrgency Procedures l

r; quiring SD isolation on AF initiation. After these changes SD Operation was permitted. Engineering cvaluated the affects on the Accident Analysis of the UF$AR. The impacted scenarios were Loss of Normal m

Feedater (LONF), Loss of Non-emergency AC Power (LOAC), and Feedwater Pipe Break (FSPB). This was due to the lack of a $G Lo Lo Water Level $D isolation on AF initiation. Calculations were perfomed using actual plant data and taking credit for manual SD isolation within 10 minutes for LONF and LOAC events or credit for r: aching a Phase A initiating setpoint within 6 minutes for FSPB events. It was determined that the UFSAR acceptance criteria was met. Braldwood $tation was justified to continue operation provided the BwCP's were rsvised and the permanent modification is completed on a prudent schedule. The modifications are scheduled.

The cause was a preservice design deficiency for unknown reasons. The UFSAR will be revised to reflect the changes. There are no previous occurrences.

2915m(103189)/2

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's LICEMME EVENT REP 0tf (LEtt TEXT Ciufflitl& TION (arm Rev 2.8 h~

' FACILITY NAME (1)

DOCKET NUMBER (2)

LER MUMBER (6)

~

Paan (31

//

Revislan Year

///

$sqteential /j//

ff fff Numbat.

///

Number

/

.Sie.ldw00d,1 -

M_L}_DJ_0J_0 1 41 51 6 a I 9' oI1I3 oIo el 2 _or el 4 TEXT-Energy Industry Identification System (Ell 5) codes are identified in the text as (XX)

~

A.

PLANT CONDITIONS PRIOR 70 EVENT:

e Unit: Braldwood 1:

Event Date: October 5, 1989; Event Time: 1200;

' Mode: 6 - Refuel; Rx Power: 0%;

RC$ (AB) Temperature / Pressure: Ambient Unit: Braidwood 2;,

Mode: 1 - Power Operation:

Rx Power: 100%;

RCS (AB) Temperature / Pressure: NOT-NOP t

B.'. DESCRIPTION OF EVENT:

1 There were no systems or components inoperable at the beginning of the event which contributed to the severity of-the event.

l At approximately 1200 on October 5,1989 Nuclear Engineering Department (NED) notified Braidwood Station of an identified discrepancy with the Steam Generator Blowdown (50) (WI) system. The Updated Fin (L Safety Analysis ReportL (UFSAR) and the Westinghouse functional requirements specify $D isolation on the initfation of the-

~

l Auxillary feedwater (AF) (BA) system. Isolation of $0 minimites the AF flow requirements. The current 50 system I

design provides for isolation on Containment Phase A Isolation and High Energy Line Break Isolation. With the f

cxisting design, isolation of the SD system was not assured upon receipt of the following AF initiation signals.

1.

Steam Generator ($G) (AB) Lo Lo Water Level.

2.

Loss of Power to the Reactor Coolant Pumps.

3.

Manual System initiation.

The $D system was immediately isolated on Braidwood 2 and an initial evaluation was performed. Based on the results of this evaluation the following actions were taken:

1.

Temporary procedure changes were made to Braidwood Emergency Procedures (BwEP). These changes require Operator Verification / Performance of SD isolation concurrent with verification / performance of AF inttletion in the BwEP's. When these changes were implemented operation of the SD system was permitted.

NED ~ as requested to evaluate the af f ects of the evisting SD system design on the Accident Analysis of the 2.

w UFSAR.

3.

The event was conservatively determined te be a four hour reportable event pursuant to 10CFR50.72(b)(2)(lii).

L The appropriate NRC notification via the EN$ phone system was made at 1332 puisuant to 10CFR50.72(b)(2)(ill).

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12915m(103189)/3

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LICENitt EVENT REPORT (LER) TEXT CONTIIRAATION Fare Rev 2.0_

fACILIT'Y NAE (1)

DOCKET NUPSER (2)

LER NupeER (6)

Pane (3)

Year

///

Sequential //j Revision

/

f ff

///

Number

///

Number _

Braldwood 1 0 1 5 1 0 1 0 1 0_.I 41 51 6 8l9 01113 010 01 3 0F_

0l _4 f tXT :

Energy Industry Identification System (Ells) codes are identified in the text as (KK) 6.

DESCRIPTION OF EVENT

- (cont'd)

On October 25, 1989 Braidwood StatI~on received the itED evaluation of lee af fects of $0 system design on the

.UFSAR Accident Analysis.. The evaluation contained _a justification for interis operation (JIO) prepared by Westinghouse Electric Corporation. The evaluation identified three Accident Analysis scenarios that were impacted by current SD system design:

~

.1.

Loss of Normal Feedwater (LONF)

~

2.

Loss of Non-emergency AC Power to the Plant Auxl11eries (LOAC).

3.

Feedwater System Pipe Break (FSPB).

Current SD system design had no imoact on the remaining UFSAR Accident Analysis. The conclusions of the cvaluation are summarized below:

1.

'The Accident Analysis is affected by the lack of a SG Lo Lo Water Level $D isolation. The Accident Analysis is not af fected by a lack of SD isolation for any other type of AF initiation.

2.

While the plants were outside the Analysis presented in the UFSAR, at no time were the plants in a condition that posed a risk to the health and safety of the public.

3.

Calculations were performed using measured values from actual plant data. These calculations also took credit for manual SD isolation within 10 minutes for LONF and LOAC events or credit for reaching a Phase A initiating setpoint within 6 minutes for FSPB events. Based on these calculations it has been determined that the UFSAR acceptance crlieria was met and the conclusions of the UFSAR are valid.

'4 Braldwood Station is justified to continue operation. This justification specifies the following:

l a.

Revisions to the BwEP's to Verify / perform SD system isolation on Af initiation.

- e b.

The permanent modifications will be completed on a prudent schedule during the next outage of sufficient duration. The scheduling will allow for engineering design, material procurement, and outage scheduling.

The BwCP revisions have been made. This was completed on October 5, 1989. The Installation of the modifications are currently scheduled for the first refueling outage for Braidwood 2 and the second refueling cutage for Braidwood 1.

The JID's-detailed discussion of the effects on each accident scenario is included in the Safety Analysis section of Byron LER 89-009-00: Docket No. 50-454.

This event is being reported pursuant to 10CFR50.73(a)(2)(v) - any event or condition that alone could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.

2915m(110389)/4

LICEN$EE EVENT. REPORT iLER) TEXT CONTINuaT10N Fere Rev 2.3

,m ', -

- FACit!TY NAME (l)

DOCKET NUPSER (2)

LER NupBER f61 panal31 j '

Year

///

Sequential

///

Revision ff

//j/ff j//

Number Number

/

l

.Braidwood 1 01510l010141516 8l9 01113 0l0 01 4 0F Old TEXT Energy Industry identification $ystem (Ell $) codes are identified in the text as (XX)

C.

CAU$E OF EVENT:

l l

The root calse of tHs event was a preservice design ' deficiency. Isolation of $D was specified in the system I

functional requirer,ents but was not included in original design documents. The cause of the failure to e I

incorporate the 'pecified'$D isolation requirements into current system design is unknown. The investigation to determine the cause is still in progress. The investigation will be tracked to completion by action item 456 200-89-17101. Any significant information pertaining to the cause of this failure will be documented in a supplemental report.

D.

$AFE b ANALY$l$:

L l

This event had no effect on the safety,0f the plant or the public. Based on the NED and Westinghouse l

Evaluation. plant operation has always been within the acceptance criteria of the UFSAR.

l i

l-

'Under the worst case condition of sustained inadequate AF flow. the emergency procedures provide for either the establishment of feed to the $Gs from the normal feedwater system or cooldown and depressurization of the i

i RC$ to a point where the Residual Heat Removal System can be placed in service using redundant ECC$

components.

l l

E.

CORRECTIVE ACTIONS

l

$D was isolated on Braldwood 2, Braidwood I was in Mode 6 with $D and AF removed f rom service. Temporary

[

Procedure changes were made to the BwEP's., these changes require operator Verification / Performance of $0.

t i.

itolation on AF initiation events. Upon implementation of these changes, $0 system opetation was permitted.

The SD system will be modified to isolate per the system functional requirements. This Modification is

(

currently scheduled for the first refuel outage on Unit 2.

This is modification M20-2-89-031 and it will be l'

tracked to completion by action item 456-200-89-17102. The modification is currently scheduled for the I

second refuel outage on Unit 1.

This is modification M20-1-89-033 and it will be tracked to completion by ection item 456-200-89-17103.

t The UF$AR will be revised to reflect the correct functional requirements for $0 Isolation as th(y relate to the various AF initiation conditions. This will be tracked to completion by action ite 456-200-89-17104.

F.

PREVIOUS OCCURRENCES

s None G.

COMPONENT FAILURE DATA

J NIne l

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1 2915m(103189)/5