05000254/LER-1997-014-01, :on 970416,target Rock Safety Relief Valve Removed from Unit 2 During Q2R13 & Unit 1 During Q1R14 Were Not Tested within 12 Months.Caused by Defective Procedure. Revised Maintenance Procedure

:on 970416,target Rock Safety Relief Valve Removed from Unit 2 During Q2R13 & Unit 1 During Q1R14 Were Not Tested within 12 Months.Caused by Defective Procedure. Revised Maintenance Procedure

ML20141G784
Person / Time
Site:	Quad Cities
Issue date:	05/15/1997
From:	Peterson C COMMONWEALTH EDISON CO.
To:
Shared Package
ML20141G775	List: 05000254/LER-1997-014, Forwards LER 97-014-00 Per 10CFR50.73(a)(2)(i)(B).Listed Commitment Made by Ltr 05000254/LER-1997-014-01, :on 970416,target Rock Safety Relief Valve Removed from Unit 2 During Q2R13 & Unit 1 During Q1R14 Were Not Tested within 12 Months.Caused by Defective Procedure. Revised Maintenance Procedure
References
LER-97-014-01, LER-97-14-1, NUDOCS 9705220493
Download: ML20141G784 (4)

LER-1997-014, on 970416,target Rock Safety Relief Valve Removed from Unit 2 During Q2R13 & Unit 1 During Q1R14 Were Not Tested within 12 Months.Caused by Defective Procedure. Revised Maintenance Procedure

Event date:
Report date:
2541997014R01 - NRC Website

text

LICENSEE EVENT REPORT (LER)

Form Rev. 2.0 Patility N_me (1)

Docket Number G)

Page (3)

Quad Cities Unit One 0l5l0l0l0l2l5l4 1 l of l 0 l 4 Title (4)

Tug:t Rock Safety Relief Valves Removed From Unit 2 During Q2R13 and From Unit i During QlR14 Did Not Receive As-Found Set Point Testing Within 12 Months Following Removal As Required By ASME/ ANSI OM-1987 Part i Due To A Defective Procedure Which Did Not Provide Clear Guidance of Test Requirements.

Event Date (5)

LER Number (6)

Report Date (7)

Other Facilities Involved (8)

Month Day Year Year Sequential Revision Month Day Year Facility Docket Numberts)

Number Number Names Quad Cities Unit 2 0l5l0l0l0l2l6l5 0l4 1l6 9l7 9l7 ol1l4 0l0 0l5 1

5 9

7 0l5l0lol0l l

l OPERATING THIS REPORT IS SUBM; a a12D PU GUANT TO T iE REQU REA ENTS OF 10CFR MODE (9)

(Check one or more of the following) (Ili l-Ul 20.402(b) 20.405(c) 50.73(a)C)(iv) 73.71(b) 0- U2 POWER Unit 1 - 100 20.405(a)(1)0) 50.36(c)(1) 50.73(a)G)(v)

- 73.71(c)

LEVEL Unit 2 - 0

- 20 405(a)(1)Oi) 50.36(c)C) 50.73(a)C)(vii)

Other (Specify (10) l l

20.405(a)(1)0ii)

T50.73(a)G)0) 50.73(a)G)(vtii)(A) m Abstreet 20.405(a)(1)0v) 50.73(a)G)0i) 50.73(a)G)(viii)(B) below and in

- 20.405(a)(1)(v) 50.73's)G)0ii) 50.73(a)C)(x)

Text)

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LICENSEE CONTACI FOR THis LER (12)

N /.M E TELEPHONE NUMBER AREA CODE Chules PJterson, Regulatory Affairs Manager, ext. 3609 3

0l9 6l5l4l-l2l2l4l1 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SY5 TEM COMPONINT MANUFACTURER REPORTABLE

CAUSE

SYSTEM CUMPONENT MANUFACTURER REPORTABLE TO NPRDS 1D NPRDS I

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I SUP >LEMENTAL REPORT EXPECTED (14)

Erpected Month Day Year Subaussion lYES Of yes. compkse EXPECTED SUBMISSION DATE) 70 Date (15) l l

l ABSTRACT (1.mna to 1400 spaces. s.o., approumsicly fitwen smgk opwe typewrmen Ames) (16)

ABSTRACT:

A Target Rock Safety Relief Valve (TRSRV), that was removed from the 2-0203-3A position during the Q2R13 refueling outage on 060195, had not been set pressure tested within 12 months. Also a TRSRV that was removed from the 1-0203-3A position during the Q1R14 refueling outage on 031596 had not been set pressure tested within 12 months. This is contrary to ASME/ ANSI OM-1987, Part 1, Paragraph 1.3.3.1(c)(2) and Technical Specification (TS) Section 4.0.E.

The root cause of the event is a Defective Procedure, which failed to provide clear instructions for ensuring prompt testing in the Target Rock Removal and Installation Procedure (QCMM 0203-31).

Corrective actions include revising the maintenance procedure responsible for the removal of the TRSRV to ensure they are shipped to the testing facility in a timely manner and making improved use of the Nuclear Tracking System to verify the testing is performed within the required time frame.

The Safety Analysis concluded that despite a small deviation above the TS set point tolerances, there would not have been a violation of reactor safety limits on maximum ASME overpressurization, Anticipated Transient Without Scram overpressurization, Reactor thermal limits (Minimum Critical Power Ratio), or loss Of Coolant Accident fuel limits.

LER254197W14.WPF 9705220493 97C3J5 PDR ADOCK 05000254 S

PDR

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Form Rev. 2.0 FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3) 0 Year Sequennal Revision Number Number Quad Cities Unit One 0l5l0l0l0l2l5l4 9l7 0l1l4 0l0 2 lOFl 0 l 4 TEXT Energy industry idennfica6on System (EBS) codes are idenufied in the text as (XXj

PLANT AND SYSTEM IDENTIFICATION

General Electric - Boiling Water Reactor - 2511 MWt rated core thermal power.

EVENT IDENTIFICATION:

Target Rock Safety Relief Valves removed from Unit 2 during Q2R13 and from Unit I during QlR14 did not receive as-found set point testing within 12 months following removal as required by ASME/ ANSI OM-1987 Part I due to a Defective Procedure which failed to provide clear instructions for ensuring prompt testing of Target Rock Safety and Relief Valves.

A.

CONDITIONS PRIOR TO EVENT

Unit:

1 Event Date:

041697 Event Time: 0800 Reactor Mode:

1 Mode Name: Run Power Level: 100%

Power Operation - Mode switch in the RUN position with average reactor coolant temperature at any temperature.

i Unit: 2 Event Date:

041697 Event Time: 0800 Reactor Mode: 0 Mode Name: Refuel Power Level: 00%

None - No Fuel in the Reactor Vessel B.

DESCRIPTION OF EVENT

Technical Specification (TS) 4.0.E.1 states: " Inservice Inspection of ASME Code Class 1,2 and 3 components and Inservice Testing (IST) of ASME Code Class 1,2 and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10CFR Part 50, Sections 50.55a(g) and 50.55a(f), respectively, except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i) or 50.55a(f)(6)(1),

respectively."

The applicable testing Code for Safety / Relief Valves for Quad Cities Station is ASME/ ANSI OM-1987 Part 1.

Paragraph 1.3.3.1(c)(2) of this Code states: "0wners that satisfy testing requirements by installing a full complement of pretested valves in place of valves that had been in service shall set pressure test the valves which were removed within 12 months of removal from the system."

Contrary to the above, Target Rock Safety Relief Valve (TRSRV)[JE][RV) (Serial Number 224) that was removed from the 2-0203-3A position and replaced during the Q2R13 refueling outage on 060195 under Work Request (WR) 940100493 was not set pressure tested within 12 months of removal.

In addition, TRSRV (Serial Number 225) that was removed from the 1-0203-3A position and replaced during the QlR14 refueling outage on 031596 under WR 950085020 also was not set pressure tested within 12 months of ree <al.

The two TRSRVs along with the TRSRV (Serial Number 171) removed from Unit 2 during the current Q2R14 refueling outage, were sent to Wyle Laboratories for set point testing.

LER254\\97\\014.w?F

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Form Rcn. '2.0 FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

Year Sequential Revision Number Number Quad Cities Unit one 0l5lol0l0l2l5l4 9l7 0l1l4 0l0 3 lOFl 0 l 4' TEXT Engy Indur.<y idenuficanon system (EllSi codes are idanufied m the text as [XXj TRSRV Serial Number 224 was as-found set point tested on 042497 and was found to have an as-found set point of 1148 pounds per square inch gage (psig). This value was outside of the TS 1% acceptance band (1123.7 psig to 1146.4 psig).

TRSRV Serial Number 225 was as-found set point tested on 042597 and was found have an as-found set point of 1095 psig. This value was outside of the TS 1% acceptance band (1123.7 psig to 1146.4 psig) and the ASME OM-1987, Part 1, 3% acceptance band (1101.0 psig to 1169.1 psig).

No sample expansion was necessary since 100% of the sample group for each Unit (consisting of 1 valve) had been tested.

TRSRV Serial Number 171 was as-found set point tested on 042597 and was found to have an as-found set point of 1134 psig. This value was within the TS t 1% acceptance band of 1123.7 psig to 1146.4 psig.

The root cause investigation and safety analysis of the set point failures for TRSRVs serial numbers 224 and 225 are being evaluated separately under Problem Identification Forms (PIFs) 97-2204 and 97-2205.

C.

CAUSE OF THE EVENT

The root cause of the event is due to defective procedures which failed in providing clear instructions for ensuring prompt testing in the Target Rock Removal and Installation Procedure (QCMM 0203-31).

,,ithough there is a provision in the Station's Electronic Work Control System (EWCS) that ensures that the TRSRV is replaced once every refueling outage, a formal trigger and tracking mechanism was not in place to ensure set point testing is performed within the required time frame.

The investigation also revealed that similar trigger and tracking mechanisms were also not included in QCMM 0203-01, " Main Steam Safety Valve Removal and Installation" or in QCMM 0203-04, " Main Steam Valve Testing".

The Main Steam Safety Valves (MSSV), however, were tested within the required time frame (prior to startup) for these valves during the Q2R13, QlR14, and current Q2R14 outages. The Inservice Testing (IST) Coordinator ensured that these valves were tested prior to startup; however the IST Coordinator did not have any provisions made to ensure the tests would have been performed in his absence.

D.

SAFETY ANALYSIS OF EVENT:

The Safety Analysis concluded that despite a small deviation above the TS set point tolerances, there would not have been a violation of reactor safety limits on maximum ASME overpressurization, Anticipated Transient Without Scram (ATWS) overpressurization, reactor thermal limits (Minimum Critical Power Ratio), and Loss Of Coolant Accident fuel limits. Also any as-found set-points below the TS set point tolerances would provide a net benefit to protect the reactor vessel and fuel if analyzed for the ASME and ATWS overpressurization limits.

LER254T97\\014.WPF

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION Form Rev. 2.0 FACILrrY NAME (I)

DOCKET NUMBER (2)

LER NUMBER (6)

PAOE (3)

Year

- Sequennal Revision Number Number Quad Cities Unit One 0j5l0l0l0l2l5l4 9l7 0l1l4 0l0 4 lOFl 0 l 4 i

TEXT Energy Industry idennficanon System (EUS) coc es are id,anufied in the text as [XX)

)

I i

i E.

CORRECTIVE ACTIONS

CORRECTIVE ACTIONS COMPLETED:

1.

This event was discussed with applicable department personnel.

2.

An EWCS Predefine was implemented to provide a Work Request each refueling outage to track the TRSRV and MSSV testing.

3 CORRECTIVE ACTIONS TO BE COMPLETED:

i 1.

Procedures which govern removal / installation ar.d inspection / repair of TRSRVs and MSSVs will be revised to include all ASME/ ANSI OM-1987 Part 1 Code requirements.

(NTS # 2541809701401)

F.

PREVIOUS EVENTS:

1.

PIF 96-221 (LEVEL III - PIR 2-96-007)

"U-2 DIESEL GENERATOR FUEL OIL TRANSFER RELIEF VALVE INLET CHECK VALVE (2-5299-3) INTERNALS WERE NEVER REMOVED AND THE i

REQUIRED INSPECTION WAS NOT PERFORMED DUE TO INADEQUATE CHANGE MANAGEMENT."

2.

PIF 96-1346 ' LEVEL III - PIR 2-96-023)

" UNIT 2 RX RECIRC SAMPLE (2-0220-45) VALVE IS AN LLRT REQUIRED VALVE, THE AS-FOUND LLRT WAS NOT PERFORMED PRIOR TO UNPACKING VALVE PER QCTP 130-1 ATT. D, REF NWR 960030616-01 DUE TO AN INADEQUATE PROCEDURE.

3.

PIF 96-2325 (LEVEL III - PIR 1-96-167)

"RHR MOV l-1001-47 AND 50 WERE NOT SEAT LEAKAGE TESTED AS REQUIRED PER THE 0Ha-1988 PART 10 DUE TO INADEQUATE CHANGE MANAGEMENT."

4.

PIF 96-2866 (LEVEL III - PIR 2-96-051)

"SB0 DG REMOTE / LOCAL /PLC BYPASS EMERGENCY START TEST HAS NOT BEEN PERFORMED FOR UNIT 2 DUE TO INADEQUATE MANAGERIAL METHODS."

5.

PIF 96-3057 - (LEVEL III - TREND PIR l-96-021)

"12 PIFS HAVE BEEN WRITTEN IN 1996 FOR MISSED SURVEILLANCES/ INSPECTIONS DUE TO INADEQUATE CHANGE MANAGEMENT."

6.

PIF 97-1811 - (LEVEL III - PIR 1-97-032)

"EWCS SURVEILLANCE SCHEDULING PROBLEMS DUE TO THE INTERFACE BETWEEN THE EWCS AND P2 COMPUTERS BEING INADEQUATE."

G.

COMPONENT FAILURE DATA

No Component Failures occurred that resulted in this report.

LER234T9"A014.WPF