05000254/LER-1986-001, Responds to NRC Re Violations Noted in Insp Repts 50-254/86-06 & 50-265/86-06.Corrective Actions:Response to Repetitive Local Leak Rate Failures Provided in Encl Rev 1 to LER 86-001

Responds to NRC Re Violations Noted in Insp Repts 50-254/86-06 & 50-265/86-06.Corrective Actions:Response to Repetitive Local Leak Rate Failures Provided in Encl Rev 1 to LER 86-001

ML20203F583
Person / Time
Site:	Quad Cities
Issue date:	07/24/1986
From:	Farrar D COMMONWEALTH EDISON CO.
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
1894K, NUDOCS 8607310121
Download: ML20203F583 (1)

LER-1986-001, Responds to NRC Re Violations Noted in Insp Repts 50-254/86-06 & 50-265/86-06.Corrective Actions:Response to Repetitive Local Leak Rate Failures Provided in Encl Rev 1 to LER 86-001

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(x)
2541986001R00 - NRC Website

text

X l

+. -

g

(*.

4 r) Colninonwealth tidison One First National Plaza. Chicago,12nois s

Address Reply to: Post Omce Box 767 Chicago, Illinoa 60690 - 0767 July 24, 1986 l

Mr. James G. Keppler Regional Administrator U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137 Subject: Quad Cities Station Units 1 and 2 Response to Inspection Report Nos.

50-254/86-006 and 50-265/86-006 NRC Docket Nos. 50-254 and 50-265 Reference: April 18, 1986 letter from C. J. paperiello to Cordell Reed.

Dear Mr. Keppler:

The referenced letter transmitted the subject Quad Cities Inspection Report which addressed Integrated and Local Leak Rate Testing. Although no non-compliances were idt;ntified during this inspection, the referenced letter requested we provid* a response addressing our program for correcting j

repetitive local leak rate teilures. Our response is provided in the attachment in the form of Revision 1 to Licensee Event Report 86-001. This report describes our corrective action efforts.

i If you have any questions regarding this response, please contact i

this office.

Very tru, yours, L

& V-

'A:

D. L. Farrar Director of Nuclear Licensing im Attachment cc: NRC Resident Inspector - Quad Cities 860724 1894K 8607310121 PDR ADOCK 05000254 PDR JUL 2 51986 G

'7 t

$l

LICENSEE EVENT REFORT (LER)

Pacility Name (1)

Dicket Numbir (2)

Paon (3) lof l8 OUAO-CITIES. NUCLEAR POWER STATION. UNIT 1 0!510101of21514 1

0 Title (4) Leak Rate From All Valves and Penetrations on Unit One in Excess of Technical specification Limit Event Date (5)

LER Number (6)

Recort Date (7)

Other Facilities Involved (8)

///j Revision Month Day Year Facility Names'I Docket Numberfs)

/j/ sequential

/

Month Day Year Year g///

p//j I

Number

/

Number Ol 51 01 01 Of I l 011 016 816 816

~~~

0 1 0 {1

~~~

0 l 1 017 019 816 01510101Of I l THIs REPORT !s SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR fCheck one or more of the followino) fil) 20.402(b) 20.405(c) 50.73(a)(2)(tv) 73.71(b)

POWER 20.405(a)(1)(1) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c) 22.405(a)(1)(11) 50.36(c)(2) 50.73(a)(2)(v11)

Other (specify LEVEL

!O 20.405(a)(1)(tit) 50.73(al(2)(1) 50.73(a)(2)(viii)(A) in Abstract below (101 0

0

//////////////////////////

20.405(a)(1)(iv)

_X_ 50.73(a)(2)(11) 50.73(a)(2)(viii)(B) and in Text)

//////////////////////////

20.405(a)(1)(v) 50.73(a)(2)(111) 50.73(a)(2)(x)

LICENSEE CONTACT FOR THIS LER (12)

Name TELEPHONE NUMBER AREA CODE Nicos P. Diarindakis. Technical Staff Enoineer Ert. 2158 1 10l 9 615141-l2121 COMPLETE ONE LINE FOR EACH COM T FAILURE DESCRIBED IN THIS REPORT (13)

CAusE

system COMP 0NEMT MANuFAC-REPORTABLE

CAusE

SYSTEM COMPONENT MANUFAC-REPORTABLE TURER TO NPRDi_

TURER TO NPROS X

JlM l P!E IN Cl3 11 10 Y

B

$ lJ l Ils IV Cl6 18 14 B

SIJ l IIS IV Cf6 18 14 X

C lE l Ils IV Cl6 18 14 SUPPLEMENTAL REPORT EXPECTED (141 Expected Month l Day 1 Year submission X lyes (If vesmc.gm_21ete EXPECTED SUBMISSJQN_DATE)

X l N0 l

l l l

ABSTRACT (Limtt to 1400 spaces.1.e. approsimately fif teen single-space typewritten lines) (16)

On January 6, 1986, while performing Local Leak Rate Testing, the measured combined leakage rate for all Valves and Penetrations, except Main Steam Isolation Valves, was found to leak in excess of 293.75 SCFH (.060 La) which is allowed by the plant Technical Specifications.

Unit One was-shutdown for the end of cycle eight refueling and maintenance outage.

This report documents the repairs made to valves and penetrations with unacceptable leak rates and the final results of the local Leak Rate Testing program.

This report is submitted to you in accordance with the requirements of 10 CFR 50.73 (a)(2)(li), which requires the reporting of any event or condition that resulted in the condition of the nuclear power plant, including its principle safety barrier, being seriously degraded.

n gh 9 h' 0

0393H

\\

NAC Fere. 3ege UA NUCLEAllD KEOULATORY COnsasa0540N LICENSEE EVENT REPORT (LER) FAILURE CONTINUATION

'" = Do"*

• 3'a-o*

E MPIRES S/31/85 F ACILITY esAmst til DOCKET NunISER (3)

LER NtJnISER tel PAGE133 Quad Cities Nuclear Power Station vEAR "R';;',

"'a*O Unit One 01510IoIo121514 816 01011 0 11 012 OF 0l8 COMPLETE ONE LINE FOR EACM COGdPONENT FAILURE DESCR18ED IN THIS REPORT 113)

MAN C-RE RT R

R A LE MANyC.

CAU$E Sv8 TEM COMPONENT CJ.USE SY$ TEM COMPONENT y qpR s g

X ClE I Il S IV Cf6I814 Y

l l l l l l l X

vIA IIlSIV Pl31410 Y

l I l l 1 l l X

VlA IIIS I V P131 4l0 y

l l l l l l l X

Wl K IIIS l V Cl6I814 Y

l l l l l l l X

WI K IIlS I V Cl 61 814 Y

l I l I

'l l l Y

Wl v IIIS ! V clA I gla Y

l l I l l 1 l v

n l.T IIIS l y cl 61814 Y

l I l l l l l v

BlJ IIIS 1 V cl 619lA y

l I I I I l l B

BlJ IIISlV M l3 I 610 v

i l I l l l l X

VlA IIISlV Bl 31114 Y

l l l l l l l 1

l l I I I I I

I I I I I I I

I I l i I I I

I I I I I I I

I I I I I I I

I l l l I I I

I i i i I I I

I I I I I I I

I I I I I I I

I I I I I l l

I I I I I I I

I I l l l l 1

I I I I I I I

l i I I I I I

I I I I I I l

l I I I l l I

l I I I I I i

i i l I I I I

I I I I I I I

I I I I I I l

l l l l 1 l l

I l l I I I i

i l i I I I I I

I I I I I I I

I I I I I I I I i 1 1 1 I

I I I I l i I

l i I I I I I

I I I I I l l

l t !

I I I i

l l

I I I I I I I

I I I I I I TC FORM 3648

LICENSEE EVENT fEPORT fLER) TEXT CONTINUATION

' FACILITY NU4 (1)

DOCKET NUMBER (2)

LER NUMBER (6)

Pace (31

,/p/p

//

Sequential /j/j/

Year

/

Revision j/

//

Number

/

Number Quad Cities Unit 1 0 15l0 10 l0 l 21 51 4 8l6 Ol0l1 0 l 1 013 0F 0!8 TEXT

PLANT AND SYSTEM IDENTIFICATION

General Electric - Bolling Hater Reactor - 2511 MHt rated core thermal power.

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

IDENTIFICATION OF OCCURRENCE:

Leak rate froa all valves and penetrations in excess of Technical Specification limit.

Discovery Date:

1-6-86 Report Date:

7-9-86 This report was initiated by Deviation Report D-4-1-86-5 CONDITIONS PRIOR TO OCCURRENCE:

SHUTDOWN Mode (l) - Rx Power 0% - Unit Load 0 MWe SHUTDOWN Mode (l) - In this position, a reactor scram is initiated power to the control rod drives is removed and the reactor protection trip systems have been deenergized for 10 seconds prior to permissive for manual reset.

DESCRIPTION OF OCCURRENCE:

At 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, on January 6, 1986, Unit One was shutdown for end of cycle eight refueling and maintenance outage. While performing refueling outage Local Leak Rate Testing, the measured combined leakage rate for all Penetrations and Valves, except Main Steam Isolation Valves, was found to leak in excess of 293.75 SCFH (0.60 La)-

The following valves required repairs or adjustments (RA's).

Note that some of the RA's were not due to excessive leakage, but were the result of preventative maintenance.

The valve leakage before and after the RA's and an explanation of the work performed is provided in Table 1.

For valves where the RA's were initiated due to local leak rate test (LLRT) results, notes are shown in the comment section with details provided in the corrective action section of this report.

This report is being submitted to comply with the requirements of 10 CFR 50.73(a)(2)(11), which requires reporting of any event or condition that resulted in the condition of the nuclear power plant, including its principle safety barrier, being seriously degraded.

0393H

LICENSEE EVENT REPORT fLER) TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER f61 Pace (3)

Year

///

s qui.ntial

///

RIvisitn p,/p

//

Number

///

Number Quad Cities Unit 1 0l 5 10 l0 10l 21 51 4 816 0l0 l 1 0 l1 014 0F Ol8 TEXT APPARENT CAUSE OF OCCURRENCE:

The first step to a good corrective action or maintenance program is to determine why the valve in question leaked.

The answer to that question is not always obvious t: hen dealing with valves that are sometimes quite large or when the air leakages are small but require repair due to regulatory limitations. At Quad Cities, we believe that we have a good program for diagnosing valve problems and facilitating repairs through the use of Station Procedure QMP 800-18 and the checklist QMP 800-S15. When any safety related and/or primary containment isolation valve is disassembled, a Quality Control inspector performs a thorough inspection of the valve in order to determine the root cause of the valve leakage (or any other problems mandating the repair). An additional inspection is performed during re-assembly of the valve.

He believe that this method of diagnostics and control on these types of repairs meet or exceed any prevailing standard within the industry.

In addition, Quad Cities maintains on file the LLRT results for every primary containment isolation valve and penetration dating back to plant startup and trends those results.

The station's willingness to repair valves or penetrations that exhibit low, but equipment specific high or increasing leakages over past LLRT results, demonstrates a sincere effort to meet the requirements of 10 CFR 50, Appendix J.

Because of the stringent testing requirements of the above regulation anri problems encountered industry-wide in meeting those requirements, the corrective action portion of this report has been prepared to identify " chronic" problems experienced at Quad-Cities. Actions taken in the past and future plans are discussed.

The specific action taken this refuel outage on all valves with RA's due to LLRT leakage is given below in Table The note numbers can be referenced back to Table I to identify the valves.

0393H

LICENSEE EVENT REPORT (LER) ? EXT CONTINUATION

' FACILITY NAME (1) 00cKET NUMBER (3)

LER NUMBER f6)

Paae f31 p/,p/

sequential

/

/,p/

p//

Revision Year

///

Number

//

Number Quad Cittes Unit 1 01510 1010 1 21 Sl 4 8l6 0 1011 0 l 1 015 0F 018 TEXT

CORRECTIVE ACTION

The immediate action taken for many of the RA's is sufficient corrective action because the leakages involved were small and no pattern of chronic failure exists.

The items of special concern, however, are valves that have a history of excessive leakage and/or large leakage rates.

These problems are identified as follows:

1)

All feedwater check valves (220-58A, B; 220-62A, B).

2)

Drywell floor drain sump and equipment drain sump isolation valves (A0 2001-3, 4, 15, and 16).

3)

HPCI Steam Exhaust Check Valve (2301-45).

4)

Drywell Head (X-4).

5)

Drywell Purge Butterfly Valves (1601-23, 24, 60).

The above problems will be discussed in detail here concerning future corrective actions required to prevent further recurrence.

1)

ALL FEEDWATER CHECK VALVES (220-58A, B; 220-62A, B)

The failure of these valves to give good LLRT results is well documented at Quad Cities and at other stations throughout the industry.

While modifications have been performed to reduce the potential of valve leakage (e.g.

modifications to the disc / seat assembly seals and hold down clamps),

the primary problem continues to be that these valves are intended to isolate a high pressure water line and we are testing them with low pressure air.

The test method does not include a wav to firmly seat the disc prior to testing.

The testing does not simulate either normal operating or accident conditions that would act to seat these valves, and normally the feedwater lines would not act as a leakage path because they are water filled.

While other stations, with NRC approval, have attempted to use water ar.d/or water / air mixtures to seat the valves prior to tcsting with air, Quad Cities has not found this t-shnique to be effective.

The quantity of water that can be introduced into a 18-inch line through a 1-inch test tap does not seem to affect closure of the valve, and at times can be counter-productive by washing rust and dirt into the seat.

The water velocity that can be developed seems inadequate to either move the disc or keep the surface free of crud.

While the station continues in its efforts to develop a better maintenance program and test procedure for this valve, we believe that the problem is to a great extent generic with these particular valves.

Unfortunately, recent industry experience with a newly designed dual seat valve offered by Anchor Darling Corp. has not been totally successful as documented in NRC IE Bulletins.

0393H

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION YACILITY CAME (1)

DOCKET NUMBER (2)

LER NUMBER f6) pace (3)

//

sequential

///

Revision

,/p/p

,/,/,

Number.

Year

//

Number

/

Quad Cities Unit 1 0l5 1 0 l 0 l 0 1 21 51 4 8 l6 O l0 11 0 11 Ol6 0F 018 TEXT The station has initiated an Action Item Request (AIR NO. 85-12) to Station Nuclear Engineering Department (SNED) to investigate this problem and determine a solution.

2)

ORYWELL DRAIN SUMP VALVES (A0 2001-3, 4, 15, 16)

In the past, these valves have had trouble passing local leak rate tests.

It was discovered that the position of these valves was the cause of these failures.

The original design had the valves installed in the inverted position and the air operators were positioned below the valves.

This affected the ability of the air operator to tightly seat the valve which adversely affected the ability of these valves to perform well during local leak rate test.

To correct this problem Modifications M-4-l(2e-83-7 (floor drains) and M-4-1(2)-83-19 (equipment drains) were performed to upright the valves and operators. Modification M-4-1-83-7 was performed this refueling outage.

This modification uprighted vavles A0-2001-3 and 4.

Inis corrective action should prevent recurrence of LLRT failure.

The failure of A0-2001-16 was due to gasket and packing degredation.

This is not considered chronic problem.

No further corrective action is required.

3)

HPCI STEAM EXHAUST VALVE (2301-45)

There is an existing AIR to SNED to investigate a replacement for this valve.

The station does not feel that the Mission Duo Check valve is the best choice for this application.

Further engineering will be performed to determine if a more suitable valve is available, or if modifications to the existing valves will make them more reliable.

4)

DRYWELL HEAD (X 4)

Quad Cities One has had repeated LLRT failures of the drywell head o-rings.

The drywell head o-rings were made by J-Bar Inc., compound No. 2405.

Nuclear Services Technical has recommended that 0-rings made by Garlock (#8364) be installed. Another Commonwealth Edison Station has been using the Garlock material and have had success with their drywell head local leak rate tests.

Garlock o-rings were installed per the recommendation by Nuclear Services Technical.

Also, special care was taken during re-installation of the drywell head to ensure proper alignment.

0393H l

LICENSEE EVENT REPORT (LER) T[XT CONTINUATION

'ACILITY NAME (1)

DOCKET CUMBER (2)

LER NUMBER (6) paae (3)

F Year

///

sequential

///

Revision

,///

Number p/p/p Number

/

Quad Cities Unit 1 0 l 5 10 l0 10l 21 SI 4 8 I6 010 l1 O l1 017 0F 018 TEXT 5)

DRYWELL PURGE BUTTERFLY VALVES As was stated in the previous section, these valves have experienced two previous LLRT failures.

The causes of the previous failures and of the most recent failures are not considered to be due to a chronic problem, however the station has noted and is concerned that the "as-left" leakage of these valves (over the past six years) has been increasing.

The valves listed below will be disassembled during the next refuel outage for seat inspection and repair or replacement as necessary:

A0 1601-23 A0 1601-24 A0 1601-60 ANALYSIS OF OCCURRENCE:

The consequence of this occurrence is that it was necessary to repair a number of containment isolation valves to bring the ccmbined measured leak rate below the Technical Specification limit prior to resuming power operaton.

Exceeding the Technical Specification limit does not pose any significant risks or hazards to public safety because the total leakage determined by Type B and C tests does not, in any way, represent a probable leakage from the containment under accident conditions.

There are a number of factors which prevent totaling Type 3 and C test results to obtain a probable containment leakage.

First, many of the Type C tests are performed by pressurizing the volume between isolation valves in series.

While the Local Leak Rate Test (LLRT) does give the total leakage for both valves, the maximum (torst case) leakage one would expect from the containments could occur when both valves leak equally.

Therefore, the probable containment leakage would be no more than half of the LLRT total for both valves, and, in fact, the leakage could be zero if all the LLRT measured leakage was through only one of the valves.

Second, a number of Type C tests are performed on valves in series with other individually tested isolation valves.

In this situation, the worst case probable containment leakage would be the minimum of the two LLRT results, not the total of the two.

Third, there are also cases where the test boundary for the Type C test consists of three or more isolation valves.

In this situation, if the LLRT result shows that only one valve repair is required, the LLRT result following the repair would be the torst possible leakge for any other valve on the boundary.

Thus, the "as left" LLRT result would also be the " worst case" leakage from the containment prior to the repair.

Fourth, Type B tests, which tests penetrations and double gasketed seals, test two sealing surfaces, one from the pressurized volume to the Primary Containment and another from the pressurized volume to the Secondary Containment.

in this case, the " worst case" leakage would be half of the LLRT result.

0393H

LICENSEE EVENT REPORT (LERI TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (3)

LER NUMBER f61 Pace f3)

,/p, Sequential

//

Revision

//

/p///

Year Num0 E

///

Nym_her Ouad cities Unit 1 015 10 l0 10 l 2l 51 4 8 l 6 O l0 l 1 0 l1 018 0F 018 TEXT The " worst case" total leakage path calculation as described is still not a true measure of expected leakge during accident conditions.

For example, a number of Type C tests are performed on systems which would, under most accident scenarios, be filled with water and pressurized (e.g. Reactor Feedwater and RHRS).

These valves, thile they may represent a substantial portion of the total measured leakge for Type B and C testing, would contribute nothing to a radiological release under most accident conditions.

In addition to Primary Containment, other engineering safeguards are designed to mitigate the consequences of a radiological release during accident conditions.

These systems are the Emergency Core Cooling System (ECCS), the Emergency Diesel Generators [EK], the Secondary Containment, the Standby Gas Testment System [BH],

and the Off-Gas [WF) " hold-up" piping and chimney.

FAILURE DATA:

Previous leak rate failures are documented in LERs 254/84-002 and 265/85-007.

l l

l l

l 0393H/01862

TABLE 1 LEAKAGE (SCFH)

DESCRIPTION

VALVE NO.

COMPONENT DESCRIPTION AS FOUND AS LEFT COMMENTS "B" Reactor Feedwater CV-220-58B Crane Tilting Disc CV 750.1 0.0 Note 1 Table 2 "B" Reactor Feedwater CV-220-62B Same as Above 1937.0 0.0 Note 2 Table 2 1

Table 2 j

Cleanup Suction MO-1201-2 6 inch Crane Gate Would Not Pressuirze 21.45(c)

Note 3 (783-UL)

Cleanup Suction M0-1201-5 Same as Above Would Not Pressuirze 21.45(c)

Note 3 Table 2 Drycell/ Torus Purge A0-1601-23 Pratt 18 inch Butterfly 207.0 (c) 63.0(c)

Note 4 Table 2 Exhaust (D 1200G) s Drywell/ Torus Purge A0-1601-60 Same as Above 207.0(c) 63.0(c)

Note 4 Table 2 Exhaust j

Drywell Floor Drain A0-2001-3 3 inch Crane Gate 75.13 9.0(c)

Note 5 Table 2 Sump (D1200G)

Drycell Floor Drain A0-2001-4 Same as Above 18.0 9.0(c)

Note 6 Table 2 Sump Drywell Equipment A0-2001-16 Same as Above Would Not Pressurize 0.9(c)

Note 7 Table 2 Drain Sump HPCl Steam Supply M0-2301-4 10 inch Crane Gate 151.9 (c) 22.46 (c)

Note 8 Table 2 (783-U)

IIPCI Steam Supply M0-2301-5 Same as Above 151.9 (c) 22.46 Note 8 Table 2 HPCI Steam Exhaust CV-2301-45 24 inch Mission Duo Check Would Not Pressurize 0.0 Note 9 Table 2 Oxygen Analyzer A0-8801D Blaw-Know 3/4 inch Globe 12.5 0.0 Note 10 Table 2 Valve TIP Ball Valve 733-1 General Pneumatics Corp 0.5 0.9 Retested After (608 KWJ06-3)

Preventative Maintenance i

ih80H

TABLE 1 (continu:d)

LEAKAGE (SCFH)

DESCRIPTION

VALVE NO.

COMPONENT DESCRIPTION AS FOUND AS LEFT CO M NTS l

TIP Ball Valve 733-2 Same as Above 0.0 0.1 Same as Above TIP Ball Valve 733-3 Same as Above 0.25 0.8 Same as Above TIP Ball Valve 733-4 Same as Above 0.0 0.9 Same as Above TIP Ball Valve 733-5 Same as Above 0.0 0.0 Same as Above ACAD System CV-2599-23A Hancock Swing Check 0.3 2.3 Retested After (1-5580W-1-XNC062)

Maintenance Not Due to Leakage Dry = ell Head X-4 Chicago Bridge and Iron Would Not Pressurize 0.0 Note 11 Table 2 Dry: ell Head X-4 Chicago Bridge and Iron 1.4 0.0 Retested After i

Preventative Maintenance NOTE:

"(c)" denotes the combined leakage of all the valves in the test boundary.

l l

4 i

1 0580H

TABLE 2 NOTE NO.

DISCUSSION 1

Work Request No. Q47066.

Repairs to feedwater check valve (CV-220-588) included installing new pins and new bushings.

The condition of the old pins and bushings were causing the valve to seat' improperly.

Leakage History:

Feedwater Check Valve'(CV-220-588) 12-31-70 2.49 SCFH 4-25-74 888.20 SCFH 6-25-74 2.99 SCFH 1-8-76 3026.0 SCFH 1-12-76 5.38 SCFH 3-8-76 13.44 SCFH 4-2-77 10.4 SCFH 1-20-79 4.4 SCFH 9-5-80 5.39 SCFH 9-8-82 Could not pressurize 12-3-82 24.64 SCFH 3-26-84 Could not pressurize 6-21-84 0.0 SCFH 1-14-86 750.1 SCFH 3-5-86 0.0 SCFH Conclusion:

The feedwater check valves are large, 18" check valves on the feedwater lines and have an erratic test history.

The main reason for this is that the valve does not seat when tested with 48 PSIG of air. All feedwater check valves are considered a chronic problem.

Reference the corrective action section of this report.

0581H

TABLE 2 (Continued)

NOTE NO.

DISCUSSION 2

Work Request No. Q47052. Some small pieces of scale and crud were discovered on the seating surface. Valve was cleaned and re-assembled.

Leakage History:

Feedwater Check Valve (CV-220-628) 12-31-70 5.63 SCFH 4-25-74 1332.00 SCFH 6-25-74 11.10 SCFH 1-10-76 3040.00 SCFH 2-4-76 14.46 SCFH 4-1-77 2.22 SCFH 1-20-79 2558 SCFH 1-26-79 4.4 SCFH 10-14-80 17.76 SCFH 09-08-82 104.5 SCFH 12-08-82 23.9 SCFH 3-26-84 563.8 SCFH 5-24-84 0.0 SCFH 1-13-86 1937.0 SCFH 2-14-86 0.0 SCFH Conclusions:

Same as for CV-220-588 0581H

i TABLE 2 (Continued)

NOTE NO.

DISCUSSION 3

Work Request No. Q47878.

The inspection of the cleanup suction (M0-1201-2) valve showed that the stellite on the disc had cracks.

The old stellite was machined off and new stellite was welded on.

Work Request No. Q46529. Valve 1201-5 was repacked due leakage visually observed during plant operation.

Leakage History: Volume Boundary M0-1201-2 and M0-1201-5 12-31-70 0.033 SCFH 7-14-74 9.10 SCFH 1-26-76 2.48 SCFH 4-5-77 5.75 SCFH 2-1-79 0.0 SCFH 10-31-80 5.75 SCFH 09-30-82 14.87 SCFH 12-08-82 1.8 SCFH 5.16-84 5.0 SCFH 1-28-86 Would Not Pressurize 3-15-86 21.45 SCFH Conclusions: The leakage history of these valves do not indicate a chronic problem with leakage.

No further action is required.

l l

l 0581H

.=

TABLE 2 (Continued)

NOTE NO.

DISCUSSION 4

Work Request No. 048234.

Repairs on A0-1601-60 consisted of an adjustment of the valve disc.

Work Request No. Q48235.

Repairs on A0-1601-23 consisted of an adjustment of the valve disc.

Drywell/ Torus Purge Exhaust Leakage History: Volume Boundary - A0-1601-23, 24, 60, 61, 62, 63 12-31-70 0.38 SCFH 4-30-73 5.50 SCFH 4-22-74 5.50 SCFH 1-5-76 5.40 SCFH 4-12-77 4.5 SCFH 2-16-79 45 SCFH 2-18-79 14.2 SCFH 11-2-80 9.0 SCFH 10-17-82 9.0 SCFH 3-23-84 81.0 SCFH 3-27-84 27.0 SCFH 2-19-86 207.0 SCFH 3-19-86 63.0 SCFH Conclusions: There has been one other LLRT failure due to valve disc alignment on these valves.

There has been one LLRT failure due to a failed packing.

These failures are not considered to be due to a chronic problem.

Reference the corrective action section of this repcrt for additional infomation.

i 1

0581H

TABLE 2 (Continued)

NOTE NO.

DISCUSSION 5

Work Request No. Q47924. A0-2001-3 was disassembled and the valve disc and seats were lapped.

Drywell Floor Drain Sump Valves Leakage History: Volume Boundary - A0-2001-3, 4 12-31-70 0.497 SCFH 4-12-74 4.00 SCFH 1-5-76 8.85 SCFH 3-24-77 2.6 SCFH 1-30-79 26.0 SCFH 2-2-79 1.85 SCFH 9-23-80 0.65 SCFH 10-20-82 1.1 SCFH 3-17-83 0.0 SCFH 4-2-84 75.6 SCFH 7-20-84 0.3 SCFH 2-26-86 3-75.13 SCFH, 4 = 18.0 SCFH 3-21-86 9.00 SCFH After Mod

• Conclusion:

The leakage history of valves A0 2001-3, 4, 15 and 16 have shown some leakage problems in the past.

Reference the corrective action section of this report.

• Modification is addressed in corrective action section of this report.

'0581H

- - - ~

TABLE 2 (Continued)

NOTE NO.

DISCUSSION 6

Work Request No. Q48558. A0 2001-4 was disassembled and the valve disc was lapped.

Drywell Floor Drain Sump Valves Leakage History: Volume Boundary - A0-2001-3, 4 12-31-70 0.497 SCFH 4-12-74 4.00 SCFH 1-5-76 8.85 SCFH 3-24-77 2.6 SCFH 1-30-79 26.0 SCFH 2-2-79 1.85 SCFH 9-23-80 0.65 SCFH 10-20-82 1.1 SCFH 3-17-83 0.0 SCFH 4-2-84 75.6 SCF,.

7-20-84 0.3 SCFH 2-26-86 3 - 75.13 SCFH, 4 - 18.0 SCFH 3-21-86 9.00 SCFH After Mod

• Conclusion: The leakage history of valves A0 2001-3, 4, 15, and 16 have shown some leakage problems in the past.

Reference the corrective action section of this report.

• Modification is addressed in the corrective action section of this report.

1 l

0581H

TABLE 2 (Continued) 4 NOTE NO.

DISCUSSION 7

Work Request No. Q48039. A0 2001-16 was disassembled and the stem packing and bonnet gasket were replaced.

Drywell Equipment Orain Sump Valves Leakage History: Volume Boundary A0 2001-15, 16 12-31-70 0.522 SCFH 5-3-74 1.34 SCFH 1-5-76 10.37 SCFH 3-24-77 4.65 SCFH 2-2-79 2.85 SCFH 9-23-80 4.75 SCFH 10-20-82 0.9 SCFH 4-10-84 16.2 SCFH 7-21-84 3.2 SCFH After Mod

• 2-28-86

> 60.0 SCFH 3-17-86 0.9 SCFH Conclusions: Stem packing and bonnet gasket failures are not' considered chronic problems with these valves.

Reference the corrective action sction of this report for corrective action that has already been taken on these valves.

• Modification is addressed in the corrective action section of this report.

i t

l l

l 0581H L

TABLE 2 (Continued)

NOTE NO.

DISCUSSION 8

Work Request No. Q44649. M0 23J1-4 was repacked.

Work Request No. Q48189. M0 2301-5 was repacked.

HPCI Steam Supply Valves Leakage History: Volume Boundary M0 2301-4, 5 12-31-70 0.15 SCFH 5-24-74 1.43 SCFH 1-15-76 4.03 SCFH 3-20-77 3.47 SCFH 1-19-79 0.0 SCFH 8-31-80 5.185 SCF:1 9-6-82 3.46 SCFH 3-7-84 1.2 SCFH 7-21-84 0.0 1-7-86 151.9 SCFH 3-17-86 22.46 SCFH Conclusions:

These valves do not have a chronic problem.

N6 further action is required.

1 l

l l

0581H l

[

~

TABLE 2 (Continued)

NOTE NO.

DISCUSSION 9

Work Request No. Q46386 and Modification M-1-85-68.

This valve is a Mission Duo Check on the HPCI turbine exhaust.

The inspection on this Valve showed damage to the rubber seat.

The valve was new in July, 1984.

The valve.was replaced this outage with a very similar valve.

Leakage History:

12-31-70 0.83 SCFH 4-16-74 4.03 SCFH 1-4-76 84.42 SCFH 1-4-76 12.86 SCFH 3-22-77 16.3 SCFH 1-19-79 190.3 SCFH 2-5-79 0.0 SCFH 8-31-80 4.02 SCFH 09-07-82 Could Not Pressurize 12-07-72 0.80 SCFH 3-12-84 Unable to Pressurize 7-13-74 4.0 SCFH 1-9-86 Could Not Pressurize 3-14-86 0.0 SCFH Conclusions: The life expectancy of this vavle appears to be approximately 1 or 2 cycles.

Because of the severe damage found during valve inspections and problems encountered in the industry with these valves used for this application, this is considered a chronic problem.

Reference the corrective action section of this report.

0581H

TABLE 2 (Continued)

NOTE NO.

DISCUSSION 10 Work Request No. 48040. A0 88010 was repacked.

Leakage History: 02 Analyzer Valve A0 8801-0 6-27-84 0.095 SCFH 1-4-76 1.77 SCFH 3-31-77 1.1 SCFH 2-3-79 0.1 SCFH 9-9-80 0.4 SCFH 9-13-82 0.2 SCFH 3-14-84 1.5 SCFH 2-27-86 12.5 SCFH 3-19-86 0.0 SCFH Conclusions: This valve does not have a chronic problem.

No further action is required.

A 0581H

TABLE 2 (Continued)

NOTE NO.

DISCUSSION 11 Work Request No.

45207.

The drywell head 0-rings were replaced with a Garlock Material.

Inspection showed o-ring degredation and potential head flange mis-alignment.

Leakage History:

12-31-70 0.0 SCFH 7-15-74 4.73 SCFH 4-18-75 0.00 SCFH 1-3-76 0.00 SCFH 5-7-77 0.00 SCFH 1-18-79 30 SCFH 2-18-79 0.0 SCFH 12-17-80 0.0 SCFH 09-06-82 Could Not Pressuirze 12-15-82 0.0 SCFH 3-7-84 30 SCFH 7-23-84 0.0 SCFH 1-6-86 Could Not Pressurize 3-15-86 0.0 SCFH Conclusions: The leakage history of the drywell head shows repeated failures.

This is considered a chronic problem.

Reference the corrective action section of this report.

4 0581H we

r SHEET 7 EVENT SUXMARY REV.1 4/86 A"ND VR Number PGK Of-1-E4rQQS CAESE CODES Ldst generation Reactor trip NRC violation, level ___

Cost > $25,000 ESF actuation GSEP event, class _____

Hazard or Spill z

NRC reportable Tech Spec LCO Personnel injury X

LER Potential or future loss Component PSE Failure mode type Denartment i

X f C. h1\\

Mr Qt M!M f

DJear. ettn 4M)

X lC-MI MI 9l M in I

ouf b-P aLNmenf X

Bread.

)< ash l

I C-Mf MI

)\\

MfM I

Licensed? L or blank Type Level Detail code DeDartment A

I i

i A

i i

A l

i i

l Type Detail Code i

I Deoartment

} D 1

s ys I

rep <epr compwaf sksnn l

. E l l

I I

.'y i

J l

I Type

, D etail code bI f

i Type of deficiency Detail code I

Procedure tvoe f) i I

)

l I

)

l

- Detail code i

TYDe Decartment 3

l l

I l

t I

I i

k f

I l

I I

f Commonwealth Edison ouad Cities Nuclear Power Station 22710 206 Avenue Nonh Corcova. Illinois 61242 Telephome 309/65 & 2241 RLB-86-102 July 10, 1986 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C.

20555

Reference:

Quad Cities Nuclear Power Station Docket Number 50-254, DPR-29, Unit One Enclosed please find Licensee Event Report (LER)86-001, Revision 01, for Quad Cities Nuclear Power Station.

This report is submitted to you in accordance with the requirements of the Code of Federal Regulations, Title 10, Part 50.73 (a)(2)(11), which requires reporting of any event or condition that resulted in the condition of the nurlear power plant, including its principle safety barrier, being sericusly degraded.

The original Licensee Event Report (LER)86-001 stated that the Local Leak Rate Testing (LLRT) program had found leakage in excess of Technical Specification limits, but did not provide a complete summary pending completion of the testing program and corrective actions.

This report addresses all valves and penetrations that had repairs performed to reduce the leakage total to within the Technical Specification limit.

Respectfully, COMMONWEALTH EDISON COMPANY Quad Cities Nuclear Power Station bb R. L. Bax Station Manager RLB/MSK/dak Encl.

cc:

J. Hojnarowski A. Madison INP0 Records Center NRC Region III 0581H