ML19344D453

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Updated 79-027/03L-1:on 791126,six Occurrences Identified Whereby Excessive Leakages Were Measured W/Respect to Primary Containment Isolation Valves.Caused by Component Failure & Seating Problems
ML19344D453
Person / Time
Site: Quad Cities Constellation icon.png
Issue date: 03/04/1980
From: Kan S
COMMONWEALTH EDISON CO.
To:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML19344D449 List:
References
LER-79-027-03L, LER-79-27-3L, NUDOCS 8003120413
Download: ML19344D453 (12)
v • d • e


Text

NRC FORM 364 U.S. NUCLEAR REGULATORY COMMISSION (7.nl UPDATE' REPORT -

  • PREVIOUS REPORT DATE: 12-24-79 LICENSEE EVENT REPORT CONTROL 8 LOCK: l l l l l l lh (PLEASE PRINT CR TYPE ALL REQUIRED INFORMATioN) 1 6

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7 8 74 REPORT DATE EVENT DESCRIPTION AND PROSA8LE CONSEQUENCES h gl While performing the Main Steam isolation Valve local leak rate test, valves I gl A0 2-203-1A, A0 2-203-2A, A0 2-203-18, and A0 2-203-2B were found to have leak rates 1 fgy-] l in excess of the 11.5 SCFH allowable limit as specified by Technical Specification i gl 4.7.A.2.i.c. The excessive leak rate did not render the MSIV's inoperable. The l 0 6 l to ta l leak rate possible through the A and B steam lines would have been 30.0 SCFH l

, f3T 1 I and 24.2 SCFH respectively had a steam line break occurred. The A0 2-203-1C, 2C, 10, l 1018: I and 2D MSIV's had satisfactory leak rate results. l E CODE sV8C E COMPONENT CODE SU8 0'E s E lOl91 l C l D l@ y@ l Bl@ l V l A l L lV lE lX l@ y@ W @ 20 7 8 9 to 11 12 13 18 19 SEQUENTIAL OCCURRENCE REPORT REVISION LER/RO & VENT YEAR REPORT NO. COO 8 TYPE N O.

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04 JJ 40 41 43 44 47 CAUSE GESCRIPTION AND CORRECTIVE ACTIONS h ,

13 l 0 l l The A0 2-203-IA, 2A, and IB valves were found to have roughened surfaces on the main 1 yl seats. A slight amount of erosion was also found on the pilot disc of the A0 2-203-1A1

, , l valve. The stroke was extended to repair the A0 2-203-28 valve. The other three l l . valves were relapped and a new pilot disc was installed on the A0 2-203-IA valve. l

, 4 l The combined corrected leak rate for the 2-203-1A and 2A valves was 1.2 SCFH and for l 7 8 80 SJye 2-203-1B and 2B valves was 3A SCgf,ggo, l status sPOwER OTHER status (3/ oisCOvERv oisCOvERY oesCRiPT:ON @

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1. LER NUMBER - LER/R0 79-27/03L-1 II. LICENSEE NAME: Commonwealth Edison Company i ll. FACILITY NAME: Unit Two IV. DOCKET NUMBER: 050-265 V. EVENT DESCRIPTION:

On November 26, 1979, local leak rate testing of the Main Steam isolation valves, procedure QTS 100-3, revealed that valves A0 2-203-l A, A0 2-203-2A, A0 2-203-IB and A0 2-203-28 had excessive leakage. The leak rates are 30.05 SCFH, 32 3 SCFH, 47 2 SCFH and 24.2 SCFH, respectively. These leak rates exceed the 11.5 SCFH allowable limit as specified by Technical Specification 4.7.A.2.i.c. All other MSIV's were leak tested satisfactorily. On November 27, 1979, the stroke of 2B valve was extended and the leak rate was found to be 8.1 SCFH, which was within the limit.

VI. PROBABLE CONSEQUENCES OF THE OCCURRENCE:

In the event of a steam line break outside the primary containment, the total leakage possible through the "A" and "B" steam lines would be 30.0 SCFH and 24.25 SCFH respectively. The excessive leakage rate did not in.any way render the MSIV's inoperable.

Upon receipt of a Group i isolation signal, these valves would have closed in the required times and performed the isolation function. Therefore, the safety implications of this occurrence are minimal.

Vll. CAUSE:

Upon disassembly of the A0 2-203-IA main steam isolation valve, an uneven surface on the main seat was observed. A slight amount of erosion was' also found on the pilot disc.

The A0 2-203-1B and 2A main steam isolation valves were also found to have roughened surfaces on the main seats.

Vill. CORRECTIVE ACTION:

Main steam isolation valve A0 2-203-1A was disassembled; a new pilot disc was installed, and the main seat was relapped.

Valves A0 2-203-1B and 2A were also relapped to eliminate the uneven surfaces on the main seats. After reassembly, valves A0 2-203-1 A, IB, and 2A were retested and found to have a i satisfactory leak rate of 1.2 SCFH for "A" steam line and 3 5 SCFH for "B" steam line. These valves are manufactured by Crane Company.

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E' @o'frE REPORT- -

' PREVIOUS REPORT DATE: 11-24-J7 LICENSEE EVENT REPORT CONTROL BLOCK: l 1

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[TTT1 I While performing the main steam line drain local leak rate test, valve M0 2-220-2 was l ITTTII f und to have a leak rate of 26.0 SCFH. This was in excess of the 18.36 SCFH I

[TT q l allowable limit as specified by Technical Specification 4.7.A.2.i. The in-line I i o ,5 ; ; redundant isolation valve, M0 2-220-1, was found to have an acceptable leak rate valuel o 8 l Subsequently the ability to isolate the main steam line drain was not adversely I gl affected. l I O ls t i I E COOE SU C E COMPONENT CODE SUSCODE SU E

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47 CAUSE DESCRIPTION AND CCRRECTIVE ACTIONS h lil0ll A small amount of steam cutting on the valve seating surface was determined to be l gg the cause of this occurrence. The valve seating surface was lapped and a second l l

,,,,,g leak rate test was performed. The corrected leak rate was 8.45 StrH.

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1. LER NUMBER: LER/R0 79-27/03L-1 II. LICENSEE NAME: Commonwealth Edison Company Quad-Cities Nuclear Power Station lit. FACILITY NAME: Unit Two IV. DOCKET NUMBER: 050-265 V. EVENT DESCRIPTION:

On November 25, 1979 Unit Two was shutdown for a refueling and maintenance outage. While local leak rate testing the main steam drain line valves on November 26, the through-seat leakage of valve M0-2-220-2 was found to be excessive. The as found leak rate was 26.0 SCFH which exceeded the Technical Specification Appendix J limit of 18.36 SCFH. A pressure decay method was used to determine this leak rate using procedure QTS 100-1. This valve is manufactured by the Crane Valve Company.

VI. PROBABLE CONSEQUENCES OF THE OCCURRENCE:

The main steam line drain valves are part of Group One isolation protection system, which is designed to minimize leakage of steam from the drywell should a main steam line rupture.

Because the two valves in series isolate the main steam line drain, the leakage from the drywell would have remained minimal.

The redundant MO 2-220-1 valve had an as-found leak rate of 1.72 SCFH. Further, these valves are usually closed during power operation and are opened during startup and shutdown to remove condensation from the main steam lines.

Vll. CAUSE:

Upon disassembly of the M0 2-220-2 valve, the seating surface showed minor signs of steam cutting.

Vill. CORRECTIVE ACTION:

The seating surface of the valve was lapped. The stem packing was also replaced as part of routine maintenance. The final combined leak rate measurement was 8.45 SCFH. This valve has had a seat repair cnce previously.

4

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[TF51 l valves MO-2-1001-34B, MO-2-1001-363, and M0-2-1001-378 was found to leak 182 SCFH, l ITTT1 I which was in excess of the allowable limit for any one valve as specified in Section l o 5 l 4.7.A.2.i of the Technical Specifications. Further investigation revealed that the l

[ypig-} l major part of the leakage was through valve M0-2-1001-368. The upstream M0-2-1001-34B l

@l valve successfully passed the local leak rate test. l 10181 l l 7 8 9 40 C E CODE SUSC E COMPONENT CODE SUSC00'E SU E M

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CAUSE OESCRIPTION ANO CORRECTIVE ACTIONS h li l0 l l The valve disc was found to be slightly warped and corrosion was found on the valve l gl seat. The disc guide pin was also bent. The valve was taken apart and the disc and l

, , l seat were cleaned and remachined to provide proper fit. The guide pin was straightenecj

, 3 l and the torque switch on the limitorque operator was adjusted. The volume was l

, , l retested and the corrected leak rate was 3.0 SCFH. l 7 8 9 Bo STA S  % POWE R OTHER STATUS Ott O RY DISCOVERY DESCRIPTION

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T. Hafera 309-654-2241 ext. 176 o NAME OF PnEPARER PHONE: $

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1. LER NUMBER: R0-79-27/04L-0
11. LICENSEE NAME: Commonwealth Edison Company Quad-Cities Nuclear Power Station Ill. FACILITY NAME: Unit Two IV. DOCKET NUMBER: 050-265 V. EVENT DESCRIPTION:

On February 4, 1980, while performing Local Leak Rate Test, procedure QTS 100-12, the volume bounded by valves M0-2-1001-34B, MO-2-1001-36B, and M0-2-1001-37B was found to leak in excess of the allowable limit for any one valve as specified in Section 4.7.A.2.1 of the Technical Specifications. The leak rate was found to be approximately 182 SCFH.

To determine which valves were leaking, each valve was tightened down manually using the hand wheel while the test volume remained pressurized. From this i t was found that manually tightening the M0-2-1001-368 valve reduced the leakage to an allowable limit. However, because the valve needed to be tightened excessively to reduce the leakage, it was concluded that the valve would not seat properly even under the maximum torque that could be supplied by the valv- motor. Work Request No. QO3352 was issued to repair the valve.

VI. PROBABLE CONSEQUENCES OF THE OCCURRENCE:

Because the redundant RHR "A" containment cooling loop was fully operable at all times, and the associated valves were found to have acceptable leakage, the overall safety of the system was never affected. The ability of the RHR system to provide adequate reactor and containment cooling during an accident as designed without the use of one of the cooling loops and other additional components has been documented in the FSAR in section 5.2 3 3 Because the MO-2-1001-34B valve located upstream of the M0-1001-36B valve had an apparent acceptable leak rate, proper line isolation would have occurred if necessary.

V11. CAUSE:

The cause of the failure of the MO-2-1001-36B valve has been designated as equipment failure. The valve disc was found to be slightly warped and corrosion was found on the valve seat.

The disc guide pin was also bent. These factors contributed to'the inability of the valve to sest properly. The M0 1001-36B valve is a 14 inch, 390 pound, motor operated gl6be type valve that is manufactured by Crane Company.

Vill. CORRECTIVE ACTION:

The valve was taken apart-and the valve disc and seat were cleaned and remachined to provide a proper fit. The guide pin was removed., straightened and rewelded into place in the valve. The torque switch on the limitorque operator was adjusted and the valve was retested. The final leak rate was found to be 3 0 SCFH which is less than the allowable 18.36 SCFH leak rate in the Technical Specifications.

There have been two previous occurrences in 1976 and 1977 related to this voume, but were both related to the M0-1-1001-37A valve.

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OATE 7 8 60 64 OOCKET NUMBER t,d EVENT QATE EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h

[TTT1l While performing the RCIC turbine steam supply valve local leak rate test, valve l ITTTIl MO 2-1301-16 was found to have a leak rate of 76.7 SCFH. This was in excess of the l ITT~i l i 18 36 SCFH allowable limit as specified by Technical Specification 4.7. A.2.i. [

gl Redundant in-line valve, M0 2-1301-17, was found to have a leak rate of 5.6 SCFH, l

[g 7;;-) l which was within acceptable leakage limits. Thus, the steam line. leakage was minimal.l ITITl I l 10181l l 7 8 9 80

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44 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS h i o l The valve body seat was found to be out of alignment such that the valve seating l 3 i l surfaces did not match up. When lapping of the valve seats did not prove successful, l t 2 l the valve was replaced. The newly installed M0 2-1301-16 valve was tested and the l l l

W, 3 l . corrected leak rate was 0.9 SCFH. -

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NRC FCCM 366 U. S. NUCLEAR REGULATORY COMMISSl!N (7 77)

  • UPDATE
  • REPORT-PREVIOUS REPORT DATE: 12-24-79 LICENSEE EVENT REPORT CONTROL SLOCX: l 1

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7 8 80 61 COCKET NUMSER 68 69 EVENT OATE REPORT DATE EVENT DESCRIPTION AND PROBA8LE CONSEQUENCES h

< ITTTI I A local leak rate test of the drywell-suooression chamber vent exhaust valves was l ITTTll performed in accordance with procedure QTS 100-30. The as-found leak rate from the l gl volume bounded by valves A0 2-1601-23, 24, 60, 61, 62, and 63 was found to be 27.0 I iO ;s; ; SCFH. This was in excess of the 18.36 SCFH allowable limit for any one valve as I lo is i i specified by Technical Specification 4.7.A.2.i. I f0TIl l I, 10181 I l 7 8 9 88 E ODE S 8C E COMPONENT CODE su8 o'E SU E g lSlDl@ ] @ IB l@ l V l A l L lV lE lX l@ l Bl@ l Ll@ 20

, 7 8 9 10 11 12 13 18 19 SEOUENTIAL OCCURRENCE REPORT REul5ICN LE R/R0 EVENT YEAR REPORT NO. CODE TYPE .

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37 40 41 4 4 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS h li l0 l l The valve disc on the A0 2-1601-23 valve was found to be off-center. The disc was I i

gl re-centered. In addition, the valve seats on A0 2-1601-61 and 62 were lapOed. After I i ,12 I I these repairs the volume bounded by valves A0 2-1601-23, 24, 60, 61, 62, and 63 was I gl retested and the combined corrected leak rate was 18.0 SCFH. l i 4 l l 7 8 9 80

$A  % POWER OTHER STATUS IS RY Ol500VERY DESCRIPTION g ] @ l0l0l0lgl NA l lBlgl Local Leak Rate Test l A TlVITV CO TENT RELEASED OF RELEASE AMOUNT OF ACTivtTY LOCATION OF RELEASE N4 NA ITTil8 9W @ IZ l@l 7 10 13 44 l l 45 80 l

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7 8 9 10 68 69 80 Z NAME OF PREPARER PHONE:

l. LER NUMBER: LER/RO-79-27/03L-1
11. _ LICENSEE NAME: Commonwealth Edison Company Quad-Cities Nuclear Power Station Ill. FACILITY NAME: Unit Two IV. DOCKET NUMBER: 050-265 V. EVENT DESCRIPTION:

On December 5, 1979, a local leak rate test of valves A0 2-1601-23, 24, 60, 61, 62, and 63, revealed an excessive leak rate of 27 SCFH. Technical Specification 4.7.A.2.! requires a leak rate for any one isolation valve of less than 18.36 SCFH.

These valves are located on the drywell-suppression chamber vent system.

VI. PROBABLE CONSEQUENCES OF THE OCCURRENCE:

During testing of this drywell-suppressi-a chamber vent volume, it was determined that the major tior of the leakage was attributable to the A0 2-1601-23 valve. Since A0 2-1601-63 and A0 2-1601-24 did not exhibit signs of excessive leakage, and are the second isolation valves in the vent lines, the total leakage through.these vent lines was less than 18.36 SCFH, and safe plant operation was not affected by this occurrence.

Vll. CAUSE:

The cause of this occurrence was component failure. When disassembled the A0 2-1601-23 valve disc was fcond to be off-center. The A0 2-1601-23 valve is an 18 inch butterfly valve manufactured by the Henry Pratt Co. The A0 2-1601-61 and 62 valves are 2 inch globe valves manufactured by the Crane Company.

Vill. CORRECTIVE ACTION:

The disc for A0 2-1601-23 was recentered. The valve seats on A0 2-1601-61 and 62 were lapped. After these repairs, a successful local leak rate test was performed. The measured leak rate was 18.0 SCFH.

4

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""I[0DITEREPORT- -

' PREVIOUS REPORT DATE 12-24-79 LICENSEE EVENT REPORT (PLEASE PRINT OR TYPE ALL REQUIRED INFoRMATION)

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REPORT OATE EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h ITTTl I While performing local leak rate testing on the A feedwater line. the volumes I

O ,3, j required to test check valves CV 2-220-SdA and CV 2-220-62A could not be pressurized. l

[T77l l On November 27. 1979 leak rate tests were performed on the B feedwater line with l ITTTl I resultant leakages of 0 SCFH for check valve CV 2-220-588 and 406.8 SCFH for check l lO ls 1 l valve CV 2-220-62B. (Technical Specification 4.7. A.2.1) l IO l 71 l I F5171 [__ f 7 8 9 80 E CODE SUSC E COMPONENT CODE SUSCOO'E SU E FOTi'l 7 8 IC IH l@ W@ l B l@ IV llaA lL IV lE IX l @ l C l @ [_0_l 9 to 11 12 18 19 20 SEQUENTIAL OCCURRENCE REPORT REVISION LE R/RO EVENT YE A R REPOR T NO. CODE TYPE N O.

@ ,afg u l217l 9l l-l l0l2l7l 24

_1/l l 01 3 l 28

[L_l l-,-j l1l 32 22 23 26 27 23 .30 31 KN AC O ONP NT M HOURS 22 S8 iT FOR1 8. SUPPLIE MANUFACTURER I22 Bl@l24 Zl@ d@

35 lasZ l@ l 0 l 0 l .0 l 0 l 2i 7 40

[ Nljg 4i ,

l Y lg 42 l42N l@ lC 44 l6 l6 15l@

47 "CAUSE DESCRIPTION AND CORRECTIVE ACTIONS i O l Corrosion on the valve seats and worn Kairaz 0-rings were found to be the cause. I i : l The valves were cleaned, reworked, and new Kalraz 0-rings were installed. T5e valves I

, 2 l were retested and the corrected leak rates were 7.8 SCFH for CV 2-220-58A, 0.0 SCFH l gl for CV 2-220-62A, and 14.9 SCFH for CV 2-220-62B. l i 4 l l 7 8 9 80 ST $  % POWER OTHER STATUS ISCO RY OtSCOVERY OESCRIPTION y W g l0l0l0lgl NA l lB[gl Local Leak Rate Test l l ACTIVITY CO TENT RELE ASED OP RELEASE AMOUNT OF ACTIVITY LOCATION OF RELEASE ,

7 i 6 8 9 l Zl@ to Ql 1:

NA 44 l l 45 NA 80 l l PERSONNEL EXPOSURES NUM8ER TYPE CESCRIPTION LLLII I I o l@l Z l@l NA l

,ERSON~El'iN; UKES

~uM s A DESCRi,TiO~@

  • g l0l0l0l@l NA l 7 8 9 11 12 80

' ^"^

'#5S PE PT ON ' " " @ l 8 I Z l@LSC 7 8 9 l

10 NA 80 I  ;

1 NRC USE ONLY ,

RIPR ,

2 O L3@Ol NA l lllllllllllllj 7 8 9 10 68 69 80 5 l J. Hoeller 309-654-2241 ext 171 o  !

NAME OF PREPARER PHONE: $ l I

NRC FORM && U.S. NUCLEAR REGULATORY COMMISSION n.n.: . UPDATE

  • REPORT-j- PREVIOUS REPORT DATE: 12-24-79 LICENSEE EVENT REPORT CONT.70L 8 LOCK: l 1

l l l l l lh 6

(PLEASE P*.lNT CR TYPE ALL REQUIRED INFoRMATsON) l I68l@

l

[ 10111

? 8 9 l l lUCENSEE L l QCOOEl A l 0 l 2 l@15 0luCENSE 14 0l 0l -l 0l 0l 0 l - l 0 l 0LICENSE NuMeER 45 26 l 0 l@l TYPE JJ4 l 57 1CAT l 1 l 1 l@

CON'T 101i1 "*'

S uRg l L l@l 0 l 5 l0 l0 l0 l2 l6 l 5 l@l 1 l 210 l 3 l 719 l@l 0 l 310 69 74 l 4 l 81080 l@

REPORT DATE 7 6 60 61 GoCKET NUMBER 68 EVENT DATE 7%

EVENT DESCRIPTION ANO PROBABLE CONSEQUENCES h 1032; l While performing local leak rate testing, the containment purge volume bounded by )

yl valves A0 2-1601-21, 22, 55, and 56 was found to have a leak rate of 135.1 SCFH. Thist

g o 3, g l was in excess of the 18.36 SCFH allowable limit for any one valve as specified by J o 3 g Technical Speci fica tion 4.7. A.2. i . The safety implications of this event were l gl minimal since valves A0 2-160-22, 55, and 56 were within acceptable leakage limits. l FaTi1 1 The volume would have isolated properly when needed. l I

Lo_L8.] I 88 y a9 E oOE SUSC E COMPONENT CODE SU8 CODE SU E ITTTl 7 8 l S l A l@ d@ l Bl@ l V l A l L lV l E lX l@ l Bl@ 20W @

9 10 13 12 13 14 19

,,, SEQUENTIAL OCCU R R ENCE REPORT REVISION EVENT YE AR REPORT NO. CCOE TYPE N O.

LER

@ ,ag/RO l 7l 9l [---J l 0l 2l 7l l/l l 01 31

8

[_L 30_,

J l-l 31 l 1l 32 21 22 23 24 26 27 29 A N A ope P T MET HOURS 22 38 iT POR 8. SUPPLIE MAN A TURER (Ejg[34,_,

V Z

,jg lJS Zlg lM Zl@ l0]0l0l 37 4 l gg 48 l Y* g 42 l N lg 43 l4 C l6 l6 l547l@

CAUSE DESCRIPTION AND CORRECTIVE ACTIONS gl Most of tl e leakage was determined to be through valve A0 2-1601-21. The valve disc j gl was found to be slightly out of alignment with the seat. The valve shaft was I

, , l adjusted to provide correct alignment and a second leak rate test .was performed on l

, 3 l the volume. The corrected leak rate was found to be 14.5 SCFH. l l

U, L4,,,,1 1 I 7 8 9 80 SYA  % POWER OTHER STATUS I C RY DtSCOVERY CESCRIPTION l NA Local Leak Rate Test l i 5 l H l@ l 0 l 0 l 0 l@l l lBl@l l ACTivlTY CO TENT (

RELEAsEc OP RELe Ast AMOUNT OP ACTIVITY LOCATION 08 RELEASE l 1 6 NA l l NA l 7 8 9 g l Zto [gl11 44 45 83 PERSONNEL EXPOSURES esuM8ER TYPE DESCRIPTION l.1,,,lf,] l 0l 0 l 0 l@l Z j@l NA l

' ' ' PERSONNE'L ' iN;UWiES " "

NUMBER OE5CRIPTION NA i R l0l0l0lgl l 7 8 9 11 12 80 LOSS 08 OR DAMAGE TO PACIL*TY Q TYPE DESCRIPTION  %./

9 NA l 7 8 9 l Zl@l10 80 DESCRIPTION 2 O ISS381 L

NA l lllllllllllll}

7 8 9 10 68 69 80 7.

D. Wykoff 309-654-2241 ext. 180o NAME OF PREPARER PHONE: i l

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