ML20044E419

LER 93-011-00:on 930423,manual Scram Initiated.Caused by Disconnected Linkage on Valve Positioner on Heater Drain Valve Due to Loose Jam Nut.Tailgate Session Will Be Held W/ Instrument Maint Dept Re Jam nuts.W/930521 Ltr

ML20044E419
Person / Time
Site:	LaSalle
Issue date:	05/21/1993
From:	Gronek M, Spedl G COMMONWEALTH EDISON CO.
To:	Office of Nuclear Reactor Regulation
References
LER-93-011-01, LER-93-11-1, NUDOCS 9305240332
Download: ML20044E419 (6)

Text

__-______ _- _____ ________________________ ________ __-____ _ __ - - _-__ __ _ _ _ _ - _ _ _ _ _ ___ -

S.

Commonweilth Edison ss LaSalle County Nuclear Station 2601 N. 21st. Rd.

Marseilles,liiinois 61341 Telephone 815/357-6761 May 21, 1993 l

l Director of Nuclear Reactor Regulation l U.S. Nuclear Regulatory Commission '

Mail Station P1-137 Hashington, D.C. 20555

Dear Sir:

Licensee Event Report #93-011-00, Docket #050-373 is being submitted to your office in accordance with 10CFR50.73(a)(2)(iv).

f G. F.Y Spedlko<-

Station Manager LaSalle County Station GFS/mg/grv Enclosure xc: Nuclear Licensing Administrator NRC Resident Inspector NRC Region III Administrator INPO - Records Center IDNS Resident Inspector 240053 930s24o332 DR ADOCK 05000373 9aos22 PDR

/[kI V

LICENSEE EVENT REPORT (LER)

Fem Rev 2.0 [

Facility Name (1) Docket Number (2) Pace (3) t (A$alle County Station Unit 1 0 15 10 10 10 13 17 13 1lofl0l4 i Title (4)  !

t Manual Scram dup to Disconnected Linkaoe on Valve Positioner on a Heater Drain Valve '

Eeent Date (5) LER Number (6) Report Date (7) Other Facilities Involved (B)  !

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

/// Number /// Number None 0 15 10 10 10 l I I

~ ~

01 4 21 3 91 3 91 3 0l1 l1 0l0 0l5 21 1 91 3 015101010l l I }

THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR ERAUNG (Check one or mere of the followino) (11) 1 20.402(b) _ 20.405(c) _X_ 50.73(a)(2)(iv) _ 73.71(b) i POWER _ 20.405(a)(1)(i) _ 50.36(c)(1) _ 50.73(a)(2)(v) _ 73.71(c)

LEVEL _ 20.405(a)(1)(ii) _ 50.36(c)(2) _ 50.73(a)(2)(vii) _ Other (Specify ,

(10) 0l9l 9 _ 20.405(a)(1)(iii) _ 50.73(a)(2)(1) _ 50.73(a)(2)(viii)(A) in Abstract

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

//////' //////////////////'//

ff fff f ff ff

_ 20.405(a)(1)(v) 50.73(a)(2)(iii) _ 50.73(a)(2)(x) Text)  !

LICENSEE CONTACT FOR THIS LER (12) ,

Name TELEPHONE NUMBER  :

AREA CODE ,

Mike Gronek. Technical Staff Eneineer. Entension 2447 8l115 3 15 17 l -16 17 16 l COMPLETE ONE LINE FOR EACH CO PON NT FAILURE DESCRIBED IN THIS REPORT (13) f CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE  ;

TURER TO NPRDS TURER TO NPRDS '

X l i I I I I I l I l ! I l I [

l l I I I I I I I I I I I I I l SUPPLEMENTAL REPORT EXPECTED (14) Expected Month I Day I Year Submission l lyes (If ves, complete EXPECTED SUBMISSION DATE) X l NO Date (15)  ; l9 l, ,

ABSTRACT (Limit to 1400 spaces, i.e. approxim.tely fif teen single-space typewritten lines) (16) j On April 23,1993 Unit 1 was in Operational Condition 1(F.un) at 99 percent power. At 0833 hours0.00964 days <br />0.231 hours <br />0.00138 weeks <br />3.169565e-4 months <br />, the "12A" {

Low Pressure Feedwater Heater (LP Heater) High Level Alarm was received in the Control Room, followed by auto isolation of the "A" LP heater string. The loss of feedwater he:'r r(s) procedure was entered and reactor power reduction to 900 We was immediately started.

i t

At 0835, the "12C" and "12B" LP heaters alarmed on high level followed by auto isolation of the "C" and "B" strings. The LP heater bypass valve,1CB007, was opened. In anticipation of a low reactor water level f condition and the possibility of entering Reactor Instability Region, the Shif t Engineer directed a manual [

scram.  !

i I

Cause of the event was a disconnected feedback linkage on the 1HD065A valve phsitioner due to a loose jam l

nut. The exact cause of the jam nut not being tight is not known. 8 Instrument Maintenance training is being changed and a tailgate session will be conducted to include the lessons learned from this event. Air supply for the 11 and 12 heaters emergency drain valves has been '

isolated as an interim measure. A query will be made to the industry for similar events associated with this i e type and model control valve positioner. '

This event is reportable pursuant to 10CFR50.73(a)(2)(iv) due to the manual scram.

t

LTCENSEE EVENT REPORT (LER) VEXT CONTINUATION Form Rev 2.0  !

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) Pace (3)  !

. . Year Revision -

- {/{/{/Sequential Number //

l{/(( Number

• tasalle County Station Unit 1 0l5101010131713 913 -

0I111 - 0 I0 01 2 0F 01 4 +

TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [XX)

PLAVT AND SYSTEM IDENTIFICATION:

General Electric - Boiling Water Reactor Energy Industry Identification System (EIIS) codes are identified in th text as [XX). t A. CONDITION PRIOR TO EVENT  !

Unit (s): 1 Event Date: 04/23/93 Event Time: 0833 Heurs Reactor Mode (s): 1 Mode (s) Name: Run Power Level (s): _o2%

i B. DESCRIPTION OF EVENT At 0833 on April 23, 1993, with Unit 1 in Operational Condition 1 (RUN) at 97% power, the "12A" Low -

Pressure Feedwater Heater (LP Heater)(HD)(SM) alarmed for high level (+3"). Innediately following the ,

alarm, heater level increased to the trip level of +4" (for a period of 10 seconds or more), which ,

resulted in the auto closure of the "A" LP heater string isolation valves 1CB005A and 1CB006A, and the "A" string heater drain pump forward valve 1HD045A. LaSalle Operation Abnormal (LOA-FW-01), " Loss of a Feedwater Heater (s)" procedure was entered and a reactor power reduction to 900 MWe was immediately started.

At 0835, the "12C" and "12B" heaters level increased to the alarm point and finally the trip point, resulting in auto closure of 1CB005B/C, 1CB006B/C, and 1HD04SB/C. In addition, the Nuclear Station .

Operator (NSO - Licensed Reactor Operator) started the CRAM array control rod insertica when _the LP  !

heater strings isolated due to the heater level reaching the trip point. '

At 0836, Operations opened the heater string bypass valve,1CB007 in accordance with LOA-FW-01.

In anticipation of a low reactor water level condition resulting from the loss of LP heater strings and the possibility of entering the Reactor Instability Region, the Shif t Engineer (SE - Licensed Senior  ;

Reactor Operator) directed the Unit 1 NSO to manually scram the reactor at 0837, and LaSalle General Procedure (LGP) 3-2, " Reactor Scram", was entered.

The 11 and 12 heaters are operated with the emergency drain valves full open and the normal drain valves closed due to inadequate flow capacities for the 11 heater normal drain v\lves because of original ,

design problems.

Following the scram, the following equipment abnormalities were identified.

t

1. The Scram Discharge Volume Vent Valve 1C11-F380 closed but had dual light indication due to a #

misaligned limit switch. The redundent valve 1Cll-F388 operated properly.

I

LICENSEE EVENT CEPORT (LER) VEXT CONT 7NUATION Fem Rev 2.0 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMPER (6) Pace (3)

Year /// Sequential ff/j// Revision fg/

// Number /U Number leSalle County Station Unit 1 01510 l 010 l 31713 9I3 -

0I1 l1 ~

0 10 Of 3 Or 01 4 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as [XX]

B. DESCRIPTION OF EVENT CONTINUED

2. The A feedwater heater string isolation valves 1CB005A and 1CB006A tripped their respective feed breakers upon closing. The Limitorque torque switches rotated past the open contact point. This caused the breakers to trip on thermal overload.

3. The Unit 1 Station Air Compressor (SAC) failed to load and the Unit 2 SAC was started. A work request was written to check the inlet valve and associated controllers.

4. The 1A Circulating Water Pump tripped on the bus transfer from the Unit Aux transformer to the system aux transf ormer following the turbine / generator trip. Two other circulating water pumps were on at the time and did not trip. This is not unusual due to the motor's design and no problems were identified upon investigation of trip.

C. APPARENT CAUSE OF EVENT .

Investigation revealed that the feedback linkage on the 12A Heater Emergency Drain Valve lHD065A positioner unthreaded and came apart. A jam nut, located on the feedback linkage, was found loose. A loose jam nut would allow the linkage to vibrate apart. The exact cause of the jam nut not being tight is not known. Without feedback force and control air, the valve positioner provided a full closed air signal to the top of the lHD065A actuator, thus causing the valve to close. With lHD65A full closed, level in the *12A" heater began to increase until level switch ILS-HD069 actuato thus causing valves 1CB005A and 1CB006A to close as designed. Af ter the "A" string isolated, the other strings isolated due to the increase in condensate flow through then, which increased the extraction steam condensing rate and overloading the heater shell side drains. Af ter the isolation, the unit was manually scrammed on the direction of the SE.

The exact cause of the nut being loose could not be determined. There were three postulated scenarios for this condition to exist:

a. Installation deficiency

b. Vibration, and

c. Work in the area.

D. SAFETY ANALYSIS OF EVENT The safety consequences of this event were minimal. The actions taken bf 0perating Personnel during this event were proper. All Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC)(BN) were available but not required during this event. All required actuations occurred as expected. Loss of Feedwater Heaters is an analyzed plant transient, as noted in the Updated Final Safety Analysis Report (UFSAR), which is included in Operator Initial and Requalification Training. The four minor equipment abnormalities did not impact safety on the operators ability to respond to the

• transient.

" ^ - ~ ~ - ^

. I LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Form Rev 2.0 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) Pace (3) <

Year fj/j Sequential /jj// Revision

//

j///

/// Number Number t.aSalle County Station Unit 1 01510l010131713 9I3 -

0I111 -

0 IO Of 4 0F 01 4 TEXT Energy Industry Identification System (EIIS) codes are identified in the text as (XX)

, E. CORRECTIVE ACTIONS

1. Operating entered LOA-FW-01, " Loss of Feedwater Heater (s)", in response to the perturbation to regain control of the transient, including opening the low pressure heater bypass valve ICB007, and ' inserting the CRAM arrays.

2. In anticipation of low reactor water level due to the loss of feedwater and the possibility of entering the Reactor Instability Region, the reactor was manually scrammed and LGP 3-2, " Reactor Scram", was entered.

3. A physical inspection was performed and it was discovered that the IHD065A was full closed and that the feedback linkage on the valve positioner had come loose.

4. An inspection of the other Unit 1 feedwater heater valves revealed similar loose jam nuts on four valve feedback linkage assemblies. These were subsequently tightened by Instrument Maintenance Personnel.

The cause of the loose jam nuts was investigated. Work was performed on two of the valves in January 1993, but no work had been conducted on the other two valves since initial startup of Unit lin 1982.

5. The Instrument Maintenance Training Program, Module 15 " Classroom and Laboratory", is being changed to include the lessons learned f rom this event. This will be tracked to completion by Action Item Record (AIR) #373-180-93-0030901. *

6. A Tailgate Session will be held with the Instrument Maintenance Department to discuss the importance of jam nuts and other retaining devices that can be observed during routine activities. This will be tracked to completion by AIR #373-180-93-0030902. ,

7. The air supply has been isolated as an interim measure for the Unit 1/ Unit 211/21 and 12/22 feedwater heater emergency drain valves. By isolating the air, positioner or booster relay failure would not cause the emergency valves to close.

8. A query will be made on the Nuclear Network for similar events associated with Masonellan, Model

1. 7401-702, Control Valve Positioners. The purpose of this inquiry is to identify industry f ailure information that may be applicable to this event. This will be tracked to completion by AIR

1. 373-180-93-0030903.

9. The limit switches were adjusted for the 1Cll-F380 per Work Request L22456.

10. The torque switches for the ICB005A and ICB006A will be tested during LlR06 by the MOV group to determine cause of problem. This will be tracked by AIR 373-180-93-0030904.

F. PREVIOUS EVENTS None.

4 p

G. COMPONENT FAILURE DATA r

KANUFACTURER NOMENCLATURE MODEL NUMBER MFG PART NUMBER Masoneilan Control Valve 7401 '702 r

EVENT

SUMMARY

suest ' l DVR Number  ;

AaND QL L,.95.Q

                                                                                                                          ~

CAESE CODES l

- Lost generation Reactor trip i NRC violation, level ___

                  ~                                                                                    -

Co s t > $25,000 ESF actuation GSEP event, class ____.1

                  ~

Hazard or Spill NRC reportable Tech Spec LC0 Z Personnel injury Z

                                                                                                       ~

LER Potential or future loss! Component 7,;g, P,5,Ede SALP functional area __; type i Department 1 k l

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