ML19332E641

LER 89-002-01:on 890416,main Steamline Low Pressure Reactor Trip,Safety Injection & Main Steamline Isolation Occurred. Caused by Mgt & Procedural Deficiencies.Formal Policy on Use of Extra Operator During Startup developed.W/891129 Ltr

ML19332E641
Person / Time
Site:	Braidwood
Issue date:	11/22/1989
From:	Querio R, Wagner J COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BW-89-1129, LER-89-002-01, LER-89-2-1, NUDOCS 8912080161
Download: ML19332E641 (8)

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i/ )1 Braidwood Nuci;;r P w:r Sttion

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November 29,1989 t

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Dear Sir:

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The enclosed Licensee Event Report from Braidwood Generating Station is '.

u , ; being transmitted to you as a Supplemental Report to LER 89-002-00.

- This report is number 89-002-01; Docket No. 50-456. ,

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1.Very truly yours,  ;

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. R.- E. Querlo ,

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Enslosure:- Licensee Event Report'No. 89-002-01 4

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4 PART 1 TITLE W EVDrf OCCURRED

' Reactor Trip Safety Injection and Main 4~

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-Steamline Isolation during Plant Heatup 04/16/89 1640

'due to Management rand Procedural Deficiencies. myg ggg

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REASW FM SUPPLDG3ffAL REPET This supplemental report is being submitted to enhance the cause of

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1 PART 2-L Acc ProcE = smI. umn wP& dd.4,.2& dpem ,

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SUPPLDGNTAL REPORT APPROVED AnD Aurn =1 ZED rom DISTRIBifr1m / g/UM. // [ %> M

' STATIN MANAGER Date

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( * -1.!CENSEE EVENT REPORT (LER) g l Facil.ity' Name (1) Docket Number (2) .fage (3)  !

_praidwood Unit 1 el El 01 01 01 41 51 6 1lofl0l6 Title (4) Reactor Trip, Safety Injection and Main Steamline Isolation During Plant Heatup due to Management Deficiency Event Date (5) LER Nygibar (6) Report DAlt (7) Other Facilities Involved (8)

Year Sequential f/j/j/ Revision Month Day Year Facility Names Docket Number (s)

Month. Day Year /

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j_ _ Number /// _ Number None 01 Sl 01 01.,D] 1 1

-01 4 11 6 81 9 81 9

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01012 7 0l 1 111 21_2 Bl 9 01 51 Ol 01 01 l l l THIS REPORT IS SUBMITTED PUR$UANT TO THE REQUIREMENTS OF 10CFR iCheck one or more of the followina) (11)

MDOE (9) 73.7)(b) 3 20.402(b) __ 20.405(c) .1 50.73(a)(2)(iv)

POWER' __ 20.405(a)(1)(1) _ 50.36(c)(1) 50.73(a)(2)(v) ___ 73.71(c)

LEVEL'- ._._ 20.405(a)(1)(ll) __ 50.36(c)(2) ___ $0.73(a)(2)(vii) ___ Other (Specify (101 _0l 0 ._ 20.405(a)(1)(iii) _X. 50.73(a)(2)(1) 50.73(a)(2)(viii)(A) in Abstract

/ /,/ / /,/,/,/,/,/ /,/,// /,/,/,/,/,/ /,/,/,/ / / __ 20.405(a)(1)(iv) _., $0.73(a)(2)(ii) _ 50.73(a)(2)(viii)(B) below and in

• 20.405( a)(1 )(v) 50.73(a)(2)(iii) 50.73(a)(2)(x) Text)

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LIC{NJEE CONTACT FOR THIS LER (12)

Name: TELEdQNE NUPSER AREA CODE Jerald Waaner. Reaulalgry As1EAD.ge Ext. 2497 8l115 4l 51 Bl l 218101 COMPLETE ONE LINE FOR EACH COMPON N FAlltlRLQ{l(RISID_ IN THIS REPORT (13)

SYSTEH COMPONENT MANUFAC-- REPORTABLE CAUSE SYSTEH COMPONENT HANUFAC- REPORTABLE CAUSE TURER TO NPRDS TURER TO NPRDS_

l I l' I I I I I I I I I I I I I I I l-I l l I l l l l l SUPPLEMENTAL REPORT EXPECTED (141' Expected tienth l Dav l Year Submission

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lyes (If ves. complete EXPECTED SWISSION DATE) X l NO l ll ll ABSTRACT (Limit to 1400 spaces, i.e. approximately fif teen single-space typewritten lines) (16)

At 1450 on April 16, 1989 a plant-heatup was being monitored using a graph display on a Control Room monitor (CRT). At 1601 the Nuclear Station Operator (NS0) attempted to repair a failed recorder. At 1640 a Main Steamline (MS) Low pressure Reactor Trip. Safety Injection (SI), and MS Isolation occurred due to RCS

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t- -pr ssure being above 1930 (P-11) psig and MS pressure less than 640 psig. At 1646 the SI signal was reset.

At 1648 SI flow was terminated. This event was caused by management and procedural deficiencies. A formal policy on the use of the extra NSO during startup and heatup operations has been developed. The Plant Heatup procedure will be revised to add a hold point to verify that all Steam Generator pressures are greater than 640 prig before RCS pressure exceeds P-?). This event will be included in Reactivity Hanagement training I sessions. The CRT graph display will be modified to include an alarm for p-11. There was a previous cccurrence of inadvertent safety injection. This was due to testing the wrong channel during the performance of a surveillance. The corrective actions addressed both root and contributing causes for the event.

Prsvious corrective actions are not applicable.

2717A(112889)/2

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Form Rev 2.0 h LIEENSEE EVENT REPORT fLER) TEXT CONTINLIATION

_LER NutgER f 61 - Pane (3)

FACILITY,N4fE (1) DOCKET NUPSER (2)'

[ , Year. / Sequential ff/j// Revision

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armi n i e l's'l o 1 0: 1 o'l '41 51 6 819 eIo12 o11 el 2 or el s  ;

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c TEXTi Energy Industry, Identification System (EIIS) codes are identified in the text es [KX)

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l A. PLANT CONDITIONS PRIOR TO EVENT '

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Unitt ' Braidwood li =

-- Event Date: April 16,1989; Event Time: 1640; .

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.Modet! 3 - Hot Standby; Rx Power: 0%;

!RCS (AB). Temperature / Pressure 500 degrees-F/1935 psig_

8 DESCRIPTIONOFEVENT5 .

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, ;ThIre were no systems cricomponents inoperable at the beginning of the event which contributed to the severity of

the event.

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At 1450 on' April. 16, 1989 the'af ternoon shif t Nuclear Station Operator (NS0) (Licensed Reactor Operator) relieved 1

. the day shif t NS0_ on Unit-1. ' A plant heatup and pressurization were in progress'in accordance with IBwGP.100-1, .

Plant Heatup. 3Two pressure loops associated with 1A and IB steam generators (SG). (AB) were simultaneously in test

to facilitate Instrument Maintenance Department.(Ile) calibrations. Plant conditions at this time were:

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RCS Pressure: 1340 psig 4

, . RCS Temp: 460 degrees F SG Pressure = 431 psig ~ ~ ~ "

Pressurizer Pressure Control: Manual j_  !

!i Pressurizer Level Control: Manual SG Level Control: Manual The heatup and'pressertration'were being nonitored using a computer graph displayed on a Control Room monitor

'(CRI); 7 This graph _ displays the actual RCS Pressure and RCS Temperature over a green Target Value Line.

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Surveillances IBw0$ 4.9.2-1, Pressurizer . Temperature Limit Surveillance, and IBw0S 4.9.1.1-1, RCS . #

. Pressure / Temperature Limit Surveillance were in progress in accordance with IBwGP 100-1.

At 1530 the Unit 1 NSO observed that the IB RCS Cold leg RTD was providing erratic indication. He notified the Station Control Room Engineer (SCRE) and Technical Specification 3.3.3.5 Action Statement was entered.

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From 1601 to 1639: .The Unit 1 NSO observed that the chart recorder pen for the failed RTD was not inking. He Cttempted to repair the'non-inking pen. During repair attempts he spilled ink on the chart, his hands, and the h . Main Control Doard. During the process of cleaning up the spilled ink, the Unit 1 NSO periodically monitored the l + hiatup'and pressurization on the CRT. He was also periodically monitoring upper norrie temperature on another

' CRT, making the ' required adjustments to the.1CV121, Pressurizer Level Control Valve, (CV) (CB) and the IFWO34A, B, L T [C, and'D,.SG Level Control Valves, (FW) (SJ) for each SG.

' . . At H1639 the-Unit 1 NSO observed that the actual Pressure versus Temperature was deviating from target value on the

, ' W htatup and' pressurization display on the CRT. After noting that pressure was higher than desired for the L temperatures he went to his desk'to refer to procedure IBwGP 100-1.

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LICENSEE EVENT REPORT ILERI-TEXT CONTINUATION ' Form Rev 2.0 FACILITY NAPE (1). DOCKET NUDOCR (2) LER NUPSER (6) Pane (3) 0 Year- // Sequential / Revision

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f J 1_thlde r // Number sralt= :2 1 0is101010l41sf6 aI9 - 0l012 - 0l 1 01 J Of 01 6 TEXT: - Energy Industry Identification System (E!!S)-todes are identified in the text as (XX)

~B.- D O CRIPTION OF EVENT:. (Cont'd)

At-1640 a Main Steamline Low Pressure Reactor Trip, Safety Injection (51) (BQ), and _ Main Steam 14ne Isolation

occurred. _This was due to RCS pressure being above 1930 psig, the P-11 setpoint, in conjunction with Main

.Steamline (SB) pressure being less than 640 psig. Braidwood Emergency Procedure,18wtP-0, Reactor Trip or j l

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' Safety injection Unit 1, was entered. Injection of the cool Refueling Water Storage Tank (RWST) water

" resulted in an increase.in RCS pressure, a decrease in RCS temperature from 500 degrees F at the start of the .j cvent,' and a decrease in the Main Steamline pressure.

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At[1644 the' resultant insurge of relatively cooler RCS water into the Pressurizer caused the Pressurizer i

Liquid: Space water temperature to decrease from its initial value of 625 degrees F at the start of the event, i l 1At 1646 Braidwood Emergency Procedure, 19wEp ES 1.1, SI Termination Unit I was entered. The SI signal was l l

reset and termination of the SI flow was initiated.

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At 1648 the High Head Safety injection Isolation Valves were closed tennineting the safety injectio flow to the RCS. The Main Steamline. pressure reached a minimum value of 612 psig. I 1

1 At 1649 RCS pressure from,the SI achieved a maximum value of 2242 psig with a Main Steamline pressure of 612 f

.psig. ;This resulted in the Administrative Limit of 1600 psid to be exceeded by 30 psid.

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At 1650 the Pressurizer. Liquid Space temperature reached a minimum value of 518 degrees F. This indicated a

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' Pressurizer Liquid space.'cooldown of 107 degrees F in a 6 minute period which is in excess of the L -Administrative. Limit for cooldown of 100 degrees F in a one hour period. However, it was still well within ithe 200 degrees F in one hour limit allowed by the Technical Specifications. As a result of the termination

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i of -the 51 flow, the insurge flow to the pressurizer stopped and as a result the Pressurizer Liquid Space j

= water tempereture started to increase. l l

At'1651 the differential pressure between the RCS and the Main Steamlines decreased below 1600 psid with the RCS at 2155 psig and the Main Steamlines at 617 psig.

-At.1659 an Unusual Event was declared and terminated pursuant to the Generating Stations Emergency plan l

l; i(GSEP) Emergency Action Level (EAL) 2.g - ECCS initiation signal and resulf, ant injection to the vessel (Not

. spurious).

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' At' 1704 the Nuclear Accident Reporting System (NARS) notification was made to declare and tenninate the

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. Unusual Event. l 1

1 At 1726 the Pressurizer Liquid Space temperature reached a peak value of 623 degrees F. This indicated a l- prsssurizer Liquid Space temperature increase of 105 degrees F in 36 minutes which is in excess of the l' ' Technical Specification Limiting Condition for Operation (LCO) of 100 degrees F in any one hour period.

Stable. plant tonditions were achieved.

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LICENSEE EVENT REPORT fLER) TEXT CONTINUATION Tore Rev 2.0 DOCKET NUPSER (2) ~LER NUPBER f6) Pane (3 FACILITY NAPE (1) s Year! /// Sequential /// Revision fff fff

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raras.x.d1 o I s 1 0 1 o I o I di 51 6 e19 - oIo12 - ol 1 01'4 0F cl 6 Energy Industry Identification System (E!!$) codes are' identified in the text as (XX)

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8.1 DESCRIPTION

OF EVENT:" (Cont'd)~

, The' appropriate'NRC notification via the ENS phone system was;made at 1732 pursuant to:

O -10CFR50.72(b)(2)(iv) .- Any event that results or should have resulted in Emergency Core Cooling System discharge .into the reactor coolant system as a result of a valid signal.

10CFR50.72(a)(1)-- The declaration of. any of the Emergency Classes specified -in the. Licensee's approved ,

Emergency Plan.

10CFR50.72(c)(1)(11() - A termination of the Emergency Class.

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The NRC notification vl's the ENS phone system was also made incorrectly pursuant to 10CFR50.36(c)(1)(ii)(A) based on a deficient Administrative Procedure for identifying and classifying events. The $1 automatically

' actuated at* the : correct setpoint.f Therefore, this reporting requirement is' inappropriate.

At approximatelyl2200 while reviewing the Pressuriser Temperature Limit Surveillance, it was discovered that

,th) bestup of the Pressurl er Liquid Space was in excess of the Technical Specificetion Limit of 100 degrees

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'T in a one hour period. The_LCO action statement was entered. An engineering evaluation was initiated to d;termine the ef fects of the heatup on the Pressurizer in accordance with the Technical. Specifications. +

4 (At0556'on. April 17,'1989[the.OnsiteReviewoftheengineeringevaluationwascompleted. The evaluation

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concluded that the structural integrity of the Pressurizer was ecceptable for continued operation.

This event-is being reported pursuant to:

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10CFR50.73(a)(2)(iv).- Any event or condition that resulted in manual or automatic actuation of any 1: Engineered Safety Feature, including the Reactor Protection System.

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Any operation or condition prohibited by the plant's Technical Specifications.

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. Bdsed 'en che initta'l information associated with this event a "Braidwood Station Error Evaluation h - Presentation" was held to review this event with the personnel directly involved and their supervisor. The p 'ccrrective. actions addressing both root and contributing causes are detailed below. ,

ChCAUSEOFEVENT:

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Th2' root cause of the event was a Management deficiency. The responsibility for assessing, prioritizing, and

reassigning available personnel is a Management function. The SCRE is the Control Room Supervisor. He has the ultimate responsibility to ensure that Control Room work assignments are properly prioritized and that cdaquete personnel are assigned to tasks in progress. The SCRE was aware of the ink spill and cleanup iefforts but determined that prioritization or assigning additional personnel was not required. Cleaning up the spilled ink became a significant increase in the workload for the Unit 1 NSO.

L .The Operating Organizational scheme requires the HS0 position to share in the Management functions of prioritizing and personnel assignment for both control room and in plant work activities. The Unit 1 NSO centinued with cleanup efforts while monitoring the heatup. He did not request assistance.

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si sq 4 LICENSEE EVENT REPORT fLER) TEXT CONTINLl& TION form Rev 2.0 DOCKET NUPSER (2) . LER NUPBER f 61- Pane (31  ;

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TEXT; ' Energy, Industry Identificaticn System (Ells) codes are identified in the text as [XX) s

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h C. CAUSE 0F EVENT: ~(continued)-

'Bhththe'SCREandtheUnit1NS0hadtheresponsibilitytoeitherensurethat1or2availableNS0swas

• assigned to Unit I until .the cleanup was completed or assign the cleanup a priority that would not interfere with the monitoring of the heatup. The deficiency of both the SCRE and the Unit 1 NSO in this Management p ,

r;sponsibility created the error.

C:ntributing causes to this event were; '

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1. ' An' extra NSO was specifically added to aid the Unit 1 NSO during the plant heatup and pressurization evol uti on. - Since the Operating Department had no formal policy on the use of this extra NSO, the Shif t

Engineer (Licensed Senior Reactor Operator) assigned him to unrelated activities associated with ,

' Unit 2. The Unit 1 NSO perceived this reassignment of personnel as direction that he was to handle the-6 Unit'I heatup alone.'  ;

2.1 1(2)BwGP 100-1, Plant Heatup, contained no control point te verify that all SG pressures were greater .

L than 640 psig before RCS pressure exceeded the P-11 setpoint.

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I D. . SAFETY ANALYSIS:'

Thire was no effect on plant or public safety from this event as it occurred, as the plant was in hot standby

.cnd all plant equipment operated as designed.

All engineered safety features and the-reactor protection system, including manual reactor trip, were j' ' cp2rable to mitigate the consequences of this event.

E. ' CORRECTIVE ACTIONS:

Isenedia' te corrective actions: ~

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1. The $1 signal was reset and stable plant conditions were established.

2 An engineering evaluation of the structural integrity of the Pressurizer was performed for the pressure

-transient. The' evaluation concluded that the structural integrity of the Pressurizer was acceptable for-continued operation.

3.- Westinghouse has performed an analysis of the effects of this event on the structural integrity of the RCS. The analysis has concluded that the impact of this event on the structural integrity of RCS components is~ insignificant.

Based on the initial information associated with this event the personnel directly involved with this event participated in a "Braldwood Station Error Evaluation Presentation" to identify the root and contributing

.causes of.this event. Based on the conclusions of this presentation the following corrective actions will be taken:

1. Operating Department has developed and established a formal policy on the use of the extra NSO during startup and heatup operations.

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LICENSEE. EVENT REPORT (LER) TEXT' CONTINUATION Fore Rev 2 9._

FACRITY NME (1) DOCKET NUWER (2) LER NUM ER (6) Pane (3)

Year.: /// Sequential /// . Revision ,

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, eral" 1 - oIsIoIoIoI41516 e19 - o1012 - oI1 el 6 Or el 6 1 i TERT Energy Industry Identification System (E!!$) codes are identified in the text as [KK) -

J q E.;JCORRECTIVE ACTIONS: -(continued); 1

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. 2. . l1(2) DwGP-100-1. Plant Heatup will .be revised to establish a hold point to verify that all steam.

generator. pressures are greater than 640 psig before.RCS pressure exceeds P-11 setpoint. This will be.

Etracked to completion by Action Item 456-200-89-06103.

1 This event will be reviewed with Operating Department personnel as part of the training associated with

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L iReactivity Management. This will be tracked to completion by Action Item 456-200-89-06104.

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4. The heatup and pressu'rlastion computer graph display will be modified to include the setpoint for P-11.

This will be tracked to completion by Action Item 456-200-89-06105.

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F.n PREVIOU$ OCCURRENCES:

Th'ere was a previous occurrence of inadvertent- safety injection, DVR 20-1-88-019/LER 50-456-88-002. However,

th;t event was due to an Instrument Mechanic testing the wrong channel during the performance of a

, surveillance. The corrective actions were implemented addressing both root and contributing causes for this

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y -cvent. Previous corrective actions are not appitcable to this event.

G. ' COMPONENT' FAILURE DATA *

. This event was not the ' result of component failure, nor did any components fall as a result of this event.

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