Text

Repts on Aos,Primary Containment Leakage & End of Cycle 3 Local Leak Rate Test
	ML20084D336
Person / Time
Site:	Dresden
Issue date:	05/09/1975
From:	Stephenson B; COMMONWEALTH EDISON CO.
To:	James Keppler; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
	299-75, 5827, 5927, NUDOCS 8304110096
	Download: ML20084D336 (28)
	v • d • e

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.' s . gd.) OnsFi itional Plaza. Chicapo, Ilknois \j 7 Address tieply to: Post Office Box 767

/. Chicago, lilinois 60690

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BRS Ltr. #299-75 Dresden Nuclear Power Station R. R. #1 Morris, Illinois 60450 May 9, 1975 Mr. James G. Keppler, Regional Director Directorate of Regulatory Operations-Region III p,a 3 7 U. S. Nuclear Regulatory Co::n:ission 799 Roosevelt Road Glen Ellyn, Illinois 60137

SUBJECT:

REPORT OF AENOR'%L OCCURRENCES, PER SECTION 6.6. A. M!D OF PRIMARY C0::TAIIMO.T LFXr' AGE PER SECTION 4.7. A.2.c 0F THE TECIUJICAL SPECIFICATIGIS DESDEt! UNIT 2 DID OF CYCLE 3 LOCAL LEAK RATE TEST SUIGMRY

References:

1) Regulatory Guide 1.16 Rev. 1 Appendix A

2) Notification of Region III NRC Regulatory Operations: See the table below.

3) Drawing Number: See the table below.

Report Number: See the table below.

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l 8304110096 750509 DR ADOCK 05000 COPY SEITI' REGION _

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+ b Report Region III Notification Drawing Report Occurrence -

Menber Telephone Telegram Number Date Date -

50-237/74-50 Fr. Johncon Mr. Koppler M-14 11/5/74 %

11/6/74 9 1400 11/6/74 0 1504 -

50-237/74-01 Mr. Johnson Vr. Keppler W14 11/7/74 a 11/8/74 C 1040 11/8/74 @ 1245

e9 50-237/74-62 Fr. Johnson Fr. Keppler M25 11/7/74 "

11/8/74 G 1040 11/8/74 0 1245 50-237/74-63 Mr. Johncon Fr. Keppler W25 11/8/74 11/15/74 O 1600 11/15/74 G1630 50-237/74-64 Fr. Johnson Fr. Keppler W25 11/11/74 11/14/74 @ 1545 11/15/74 0 0855 50-237/74-65 Fr. Johnson Vs. Keppler W29 11/13/74 11/14/74 0 1545 11/15/74 C 0855 ,

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50-237/74-68 Mr. Johncon Mr. Keppler W29 11/14/74 8 11/15/74 e 1600 11/15/74 0 1630

50-237/74-72 Mr. Johncon Fr. Keppler B-24 & 26 11/21/74 11/22/74 C 1600 11/22/74 G 1600 50-237/74-71 Fr. Knopf Mr. Keppler L51 11/29/74 i

12/2/74 0 1310 12/2/74 @ 1335 l 50-237/74-75 Fr. Johnson Fr. Kcppler 0.-6535A 12/20/74 12/12/74 12/12/74 8 1545 12/18/74 @ 0920 50-237/74-76 Fr. Johnson Fr. Keppler ~

12/13/74 G 1545 12/16/74 o B-24 & 26 12/20/74 12/12/74 k e

50-237/74-80 Fr. Johncon Fr. Keppler W12 12/23/74 12/24/74 4 1030 12/24/74 G 1140 d d

i 50-237/75-3 Mr. Johnson Fr. Keppler W26 1/20/75 1/21/75 G 1000 1/21/75 0 1120 1

50-237/75-6 Mr. Johncon Fr. Kcppler W29 1/22/75 1/23/75 G 1445 1/23/75 0 1530

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,. . Mr. Jamts G. Kep4 r May 9, 1975

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Focility: Dresden Nuclear Power Station, Morris, Illinois IDENTIFICATION OF OCCURRF2 ICE

'dhile performing local leak rate tests for Dresden Unit 2 during the recent refueling outage, as per section 4.7.A.2.e of the Technical Specification, the following tests exceeded the maximum allowable limit:

Report Test Volume Test System Max leakage Number Boundary Tvre SCFH G 48 PSIG Valves:

50-237/74-60 220-57A & 58A Primary isol Reactor feed 29 38 k valve piping 50-237/7660 220-57A & 62A Primary isol Reactor Feed 29 38 valve Piping Valves:

50-237/74-61 220-57B & 58B Primary isol Reactor Feed 29 38 valve Piping 50-237/74-61 220-57B & 62B Pri=ary isol Reactor Feed 29 38 valve Piping Valve Flanges:

50-237/74-62 1601-32D Double Pressure 29 38 or combined Gasketed Suppression leakage of

_ 58.763 50-237/74-62 1601-33D Double Pressure 29 38 or combined Gasketed Suppression leakage of

_ 58.763 Valves: 8502-501, 50-237/74-67 1601-21, 22, 55,&56 Primary Icol Pressure 29 38 Valve Supprocsion Valve flanges:

50-237/74-64 1601-32B Double Prcosure 29 38 or combined Gasketed Suppression leakage of 58.763 Valve flanges:

50-237/74-64 1601-33C Double Pressure 29 38 or combined Ganketed Suppression leakage of 58.763 50-237/74 64 1601-33E Double Pressure 29 38 or combined Gasketed Suppression leakage of 58.763 50-237/74-64 1601-33F Double Pressure 29 38 or combined Gasketed Suppression Icakage of 58.763 Valves:

50-237/74-65 1501-18A & 19A Primary Isol LPCI 29 38 valve 50-237/74-65 1501-20A & 38A Primary isol LPCI 29 38 valve r

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IDENTIFICATION OP OCCURRENCES (cont'd)

O R'e port Test Volume Test System Max leakage SCFH @

Number- Boundary Type 48 PSIG Valves:

50-237/74-68 1501-18B & 19B Primary isol LPCI 29 38 viv 50-237/74-68 1501-20s & 38B Primary isol LPCI 29 38 v1v 50-237/74-68 1501-27B & 28B Primary isol LPCI 29 38 v1'y Penetration:

50-237/74-72 X-202s Eleetrieal Primary 29 38 Pent Containment Valves:

50-237/74-71 2301-45 & 74 Primary isol HPCI 29 38 v1v Penetration:

50-237/74-75 X-105C Bellows g ry 29 38 seal Penetration:

50-237/74-76 X-123 Bellows Primary 29 38 seal Cont.

Valves:

50-237/74 80 203-1C & 2C MSIV Main steam 11 5 G 25 PSIG piping Valves:

50-237/75-3 205-2-7 & blind Primary isol Rx Head Cooling 29 38 flanco valve Valves:

50-237/75-6 1501-25B & 26B Primary isol LPCI 29 38 valve Due to the excessive leakage through the feedwater check valves 220-58A and 220-62A, report number 50-237/74-60 includes the fact that the total through leakage for all penetrations (exluding double gasketed seals) exceeded the Technical Specification limit of 178.29 SCFH at a pressure of 48 PSIG.

CONDITIONS PRIOR TO OCCURRENCE l

At the time of each occurrence, Dresden Unit 2 was shutdown for refueling. The position of the reactor mode switch was as follows:

Report Number Rx Mode Switch Report Number Rx Mode Switch Position Position 50-237/7h-60 Shutdown 50-237/74-72 Refuel I E 50-237/74-o1 Shutdown 50-237/74-71 Refuel

[ 50-237/74 62 Shutdown 50-237/74-75 Refuel

~

g;.037/74-67 Shutdown 50-237/74-76 Reibel 50-237/74-64 Refbel 50-237/74-80 Shutdown

^

50-237/74-65 Refuel 50-237/75-3 Shutdown So-237/74-e8 Refuel 50-237/75 6 Refuel

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,, Mr,. James G. Kopp . May 9, 1975

.Os DESCRIPPIOI! OP OCCURRDICE Report Number 50-237/"4-60:

I At 1615 hours0.0187 days 0.449 hours 0.00267 weeks 6.145075e-4 months on November 5, 1974 the feedwater check valves in "A" line were leak i rate tested by using the pressure decay method at 48 PSIG to detemine the "as j found" Icakage. The two check valves (220-58A & 62A) were tested independantly I

of each other; therefore, the maximum through leakage was the minimum of the individual leakages. The 220-57A manual valve was a boundary for both tests,

! but was assumed not to leak because of a waterhead present on the reactor i side. Upstream of each check valve leakagcc were verified. The 220-58A l and 220-62A had leakages of 96, 393 sCn1 and 5,835 sCMI respectively, see

] Appendix "A" for the final leakage rates.

l Report Number 50-237/74-61:

l At 0220 hours0.00255 days 0.0611 hours 3.637566e-4 weeks 8.371e-5 months on November 7, 1974 the feedwater check valves in "B" line were leak rate tested. The same method and valve configuration as used in the "A" line l tests were utiliacd. The 220-58B and 220-62B valves were found to have leakages of l 3,238 sCMI and 358.2 SCMi respectively, see Appendix "A" for the final leakage i rates.

, Report Number: 50-237/74-62:

l At 1520 'hsura on November 7,1974 the torus to drywell vacuum breakers, 1601-32D j and 1601-33D, were leak rate tcsted by using the pressure decay method at 48 PsIG j to determine the "as found" leakage. The vacuum breakers have a " double gasketed" l type seal. Each vacuum breaker test simultaneously checks:

I 1 One cover flange

2. Two shaft seals 3 Two shaft seal flanges

)

i since the volume tested is between the doubic concentric gaskets, the maxicum through leakage is assumed to be half the total leakage, if the leakage sourco (or sourew) cannot be found. Vacuum breakers 1601-32D and 1601-33D had total leakages of 48.859 SCni and 121 701 sCMI.respectively, yielding a total through leakage of 85 250 SCHi. See appendix "B" for the final leakage rates.

Report Number 50-237/74-67:

! At 1415 hours0.0164 days 0.393 hours 0.00234 weeks 5.384075e-4 months on November 8, 1974 the inerting and purge line for the drywell and 1

torus was leak rate tested by using the piesaura decay method at 48 PsIG to j determine the "as found" Icakasc. The A0-1601-55 valve was known to leak prior to the test; therefore, during operation, the 8503-500 manual valve was closed and out of service, thus providing primary containment. The total test i leakage was 40 910 SCMI. The maximum through leakage is 20.455 sCMI. see Appendix "A" for the final leakage rate. ;

Report Number 50-237/74-64:

At 1340 hours0.0155 days 0.372 hours 0.00222 weeks 5.0987e-4 months on November 11, 1974 the torus to drywell vacuum breakers 1601-32B,

1601-33C, 1601-33E, and 1601-33P were leak rato tested in the same manner as 1601-32D and 1601-33D. The respective individual leakages were 146.041 sCMI, 1

243.849 SCMI, 40.716 SCMI, and 52 349 SCHI. Combining the through leakage a

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Mr. Jamts G. K;pplcr -

-5a- May 9, 1975 O O l of all. twelve vacuum breaker yicids 358.487 SCMI. See appendix "B" for the final leak rates.

l Report Number 50-237/74-65:

j At 1445 hours0.0167 days 0.401 hours 0.00239 weeks 5.498225e-4 months on November 13, 1974, the isolation valves for the test line and suppression chamber spray on LFCI "A Loop" were leak rate tested by using the l pressure decay method at 48 PSIG to detemine the "as found" leakage. The

isolation valves and the leakages for these lines are MO-1501-20A & 38A/

l 537.769 SCHI and MO-1501-18A & 19A/952.101 SCMI respectively. The two tests were repeated, but there was an 80 PSIG waterhead present against the l M0-1501-38A and MO-1501-18A. The results were then 0.0 SCFH and 13.885 SCMI l respectively. Assumming the above leakages to be through MO-1501-20A and MO-1501-19A, the total through Icakages were 0.0 SCMI for valves MO-1501-20A

& 38A and 13.885 SCn1 for valves MO-1501-18A & 19A. See appendix "A" for the final leakage rates.

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Report Number 50-237/74-68:

l At 1330 hours0.0154 days 0.369 hours 0.0022 weeks 5.06065e-4 months on November 14, 1975, LPCI "B Loop" valves M0-1501-18B & 19B (suppression chamber spray), M0-1501-20B & 38B (tect line), and MO-1501-27B & 28B (containment spray) were leak rate tested by using the pressure decay method at 48 PSIG to detemine the "as found" Icakage. In each test, one valve leaked a great deal more than the other, as verified by sightglasses located in l

upstream vents. The valves found to be leaking was MO-1501-18B, 383, and 27B.

The test on valves M0-1501-27B & 28B was repeated with an 80 PSIG waterhead present on the other side of the MO-1501-27B valve. The result of this test was 2.462 SCMI. The total leakages and maximum through leakages are 7228.108/

3614.054 SCMI for M0-1501-18B & 19B, 987.132/493.567 SCni for M0-1501-20B &

38B, and 959 532/2.462 SCFH for MO-1501-27B & 28B. See appendix "A" for the final leakage rates for the above tests.

Report Number 50-237/74-72:

At 0915 hours0.0106 days 0.254 hours 0.00151 weeks 3.481575e-4 months en November 21, 1974 c1cotrical penetration X-2025 was being leak rate tested by using the pressure decay method at 48 PSIG to determine the "as found" leakage. As por procedure, nitrogen was used instead of service air.

The total penetration leakage was found to be 217.834 SCHI. Using a conic gun and

" snoop", no leaks were detected in the outer seal. After removing the inner cover and shield blocks, the source of the leakage was found to be in the drywell. Cracks were found in the epoxy used to form a seal where the cables come through the inner boundary. Since no leaks were detected in the outer seal, the through leakage is 0.0 SCMI. See appendix "C" for the final leak rate results.

! Report Number 50-237/74-71:

At 1730 hours0.02 days 0.481 hours 0.00286 weeks 6.58265e-4 months on November 29, 1974 the check valves in the HPCI condensate return line were leak rate tested by using the pressure decay method at 48 PSIG.

l The two valves in the line are 2301-45 (a check valve) and 2301-74 (a stop check l valve). The total test leakage was found to be 2,611 707 SCMI. Using l

the line to the 2368A pressure switch as an upstream vent, the 2301-45 valve was found to be leaking. Using the sonic gun at penetration X-317A (HPCI condensato return to torus), the 2301-74 valve was found to have no leaks. Since all the leakage was through the 2301-45 valve, the total through leakage was 0.0 SCMI. See appendix "A" for the final leakage rate results.

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! ,. Mr. James G. Kepp . May 9, 1975 l

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! Report Number 50-237/74-75: ,

For the " description of occurrence", see the letter from B. B. Stephenson to Mr. J. G. Koppler dated December 20, 1974. See appendix "D" for the final leakage rate.

Report Number 50-237/74-76:

I For the " description of occurrence", see the letter from Mr. B. B. Stephenson to Mr. J. G. Keppler dated December 20, 1974. See Appendix "D" for the i final leakage rate.

! Report Number 50-237/74-80:

'l At 1400 hours0.0162 days 0.389 hours 0.00231 weeks 5.327e-4 months on December 23, 1974 the main steam isolation valves on"C" steam line were local leak rate tested by using the pressure decay method j at 25 PSIG to determine the "as found" leakage. The vessel water level was at

a level below the steam parts; therefore, A0-203-1C and A0-203-2C were i tested simultaneously. Both valves had the air supply valved in to assist the springs in keeping the valves in the closed position. The first test yielded a total leakage of 159.433 SCMI c 48 PSIG (92.189 Scal 0 25 PSIG). Valve l A0-203-10 was cycled a few times and then the test was repeated, yielding a

, total 1cakage of 260.215 0 48 PSIG (150.441 Scat C 25 PSIG). On December 31, 1974 repairs on A0-203-1C were co=plete and the test was repeated a third time (vessel water level below steam ports). The third test ycilded a total leakage of 154.485 Scat c 48 PSIG (89.314 SCMI G 25 PSIG). Assuming that A0-203-1C was leak tight, then A0-203-2C had a leakage of 154.485 Scat c 48 PSIG

(89 314 SCMI C 25 PSIG). On March 12, 1975 repairs to A0-203-2C were complete and

! the test was again repeated (vessel water below steam ports). The total leakage i of this test was 0.0 SCRI.

Report Number 50-237/75-3:

At 1130 hours0.0131 days 0.314 hours 0.00187 weeks 4.29965e-4 months on January 20, 1975 the reactor head cooling 205-2-7 check valve

) was leak rate tested by the pressure decay method at 48 PSIG to determine the "as

found" Icakage. A flange was installed where the line normally connects -to the' reactor head to provide a downstream boundary for the test. The leak rate was found to be 53 779 SCai and was verified to all be through the 205-2-7 check valve by use of an upstream vent. In order to determine the through leakage, a second test was performed between MO-205-2-4 valve and the installed flange. This test includes the volume in the previous test. This

! test yielded a leak rhte of 0 310 SCHI. See appendix "A" for the final leakage rate.

Report Number 50-237/75-6:

At 1400 hours0.0162 days 0.389 hours 0.00231 weeks 5.327e-4 months on January 22, 1975 the LPCI A0-1501-25B was leak rate tested for the first time. A 48 PSIG pressure decay test was used to determine the "as

! found" leakage. The other valve used to form a boundary was the 1501-26B manual valve. This valve ties into the recirculation header and therefore had a water- l l head present. The resulting leakage of 56.632 SCMI was then consequently all through ;

the check valvc as verified by an upstream vent. A.second test was done using ,

j 1501-26B, MO-1501-22B and MO-1001-5B as boundaries. This volume includes the volume of the previous test. This tested yicided a leakage of 0.635 SCMI and therefore this is the maximum penetration through leakage. See Appendix A }

l for the final leakage rate.

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', Mr.. James G. Keppl . . May 9, 1975 DESIGNATION OF APPARENT CAUSE OF OCCURRENCE Report Numbers 50-237/74-60 & 61:

Inspection of the feedwater check valves subsequent to leak testing revealed the valves to be missing their "0" rings which cauce the valves to seal at low pressures. The missing "O" rings were made of a silicone based rubber called "Silartic". This material is rated for a maximum temperature of 450*F in an air environment. In a water environment, at the temperature of the feedwater system

( 340 F), this type of "0" ring dissociated into sand and carbon dioxide.

Report Numbers 50-237/74 62 & 64:

The torus to drywell vacuum breakers are a double gasketed typc seal and cach valve hac five test points which are as follows:

a) 1 double gaskoted flanged cover b) 2 double casketed shaft seal flanges c) 2 shaft seals In all the cases where the vacuum breakers exhibited any leakage at all, the leakage was through the shaft seals. This leakage was due to a lack of the waterless grease used as a packing. In the cases where excessive leakage was exhibited, the zerk fittings were found to be bad.

Report Number 50-237/74-67:

Prior to the refueling outage, valve A0-1601-55 cxhib h ed excessive leakage rates, therefore, the 8503-500 manual valve was cloced and taken out of service until the unit was shutdown for refueling. The A)-1601-22 valve exhibited come disc to seat leakage and some packing leakage. It could not be dctornined if the A0-1601-21 & 56 valves had any disc to seat leakage though they did not have any packing leaks. The 1601-56 valve was removed and the line was blind flanged.

The tect was repeated with a resulting leakage of 6.639 SCFH, which was within Technical Specification limits.

Report Numbers 50-237/74-65 & 68:

The following LPCI valves are all throttle control valves and are used to regulate flow:

a) MO-1501-18A & B b) MO-1501-38A & B .

The leak rate tests involving the above valves were all repeated, but prior to the tests, it was verified that the control room operator held the control switch in the "close" position long enough to completely close the valve. The leakage rate of the above tests were all within Technical Specification limito, therefore, during the first tests either the operator failed to completely close the valves or the limit switches were not properly ad, justed.

Failure of the MO-1501-27B valve was due to excessive seat to dice clearance and improper operation of the limitorque.

May 9, 1975 Er. ames G. Kepplc 8-Report Number 50-237/74 72:

Failure of the electrical penetration X-202S was due to cracks in the epoxy on the electrical inside seal. The epoxy is used for filling in the gaps between the cables and the pipe they are contained in. The cause of the cracks in unknown.

No abnormal environmental conditions were observed.

Report Number 50-237/74-71, 50-237/75-3 & 6:

The 2301-45, 205-2-7, and A0-1501-25B all had excessive discNo to other seat leakage defects due to dirt on the seats and small scratches in the discs.

could be found.

Report Numbers 50-237/74-75 & 76:

For the " designation of apparent cause of occurrence", see the letters from Mr. B. B. Stephenson to Vr. J. G. Koppler dated December 20, 1974 Report Number 50-237/74 80: 4 Ecth main steam isolation valves A0-203-1C & 2C were found to be dirty, and contain small cracks in the seats and mating disc surfaces. No major defects were indicated or extensive repairs required.

ANALYSIS OF OCCURRENCE The health and safety of the plant personr.el and the public were not jeopardiced as a result of any of the above mentioned Icak rate test failures.

Report Numbers 50-237/74-60 & 61:

Although the feedwater check valves exhibited large leak rates at 48 PSIG, they where considerded functional. If a feedwater line had broken during operation, the high differential pressure conditions created would have been more than adequate to cent the valves. If a low pressure condition existed and the line broke within secondary containment, the standby-cas treatment system would offer a sufficient barrior to the atmosphere. If the line break was in the turbine building, motor-operated valves were available to sufficiently isolate the Icakage, with the exception of two lengths of pipe approximately 12 feet cach between the motor-operated valves in the turbine building and the secondary containment. The probability of a line break here is very remote.

Report numbers 50-237/74-62 & 64:

Should an accident have occurred, the combined leakage from all the torus to drywell vacuum breakers,uhich are within the bounds of the secondary containment, would have been well within the capability of the standby-gas treatment system, to prevent direct release to the environment.

Report Number 50-237/74-67:

The A0-1601-55 valve had failed prior to the refueling outacc. The manual valves, 8502-501 and 8503-500, were closed and taken out of service. The system passed

4

May 9, 1975

. Mr. James G. Koppl .

the ' remaining tests until this end of cycle leak test.

Report Mumbers 50-237/74 65 & 68: l The Icakage through j There is no safety significance, even with a LPCI line break.

the inner cost valve in cach of the five cases is within the technical specification limits.

Report Number 50-237/74-72:

The Only the inner scal of c1cetrical penetration X-202S exhibited leakage.

outer seal was still providing primary containment within the technical specification limit. If the outer seal failed during an accident, all leakage would be into the secondary containment and within the capabilitics of the standby-gas treatment system.

Report Number 50-237/74-71:

}

There is no safety significanco associated with the 2301 45 failure because 2301-74 is a stop check valve and is nearer to the torus penetration. Even though the 2301-74 valve was closed manually during the Icak test, it scated properly and there is no reason why it should not do the same in the case of an accident causing reverse flow.

Report Numbers 50-237/74-75 & 76:

For the " analysis of occurrence" for each of these failures seo letters from Mr. B. B. Stephenson to Mr. J. G. Koppler dated December 20, 1974. -

Report Number 50-237/75-3:

There is no safety significance associated with the 205-2-7 failure because the penetration through leakage at valve MO-205-2 b was within the technical specifications Limit.

Report Number 50-237/75-6:

There is no safety significance associated with the A0-1501-25B failure because the penetration through Icakage at valves MO-1501-22n and MO-1001-5B was within the Technical Specification limit. The LPCI injection line is not affected by the failure as far as operability is concerned.

Report Number 50-237/74-80:

Although the "C" main steam line isolation valves exhibited an excessive leakage rate at 48 PSIG, they were still considered functional. During operation the differential pressure conditions created following an isolation would be more than adequate to seat the valves. For a steam line break outside the drywell, I the leakage through the valves is insignificant compared to the initial blowdown values as described in the FSAR. For breaks innido the drywell, the leakage through the valves would be contained by the main stop valves, turbino control valves, and bypass valves.

(- m

May 9, 1975 Mr. James G. Ktppler_ ,

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CORRECTIVE ACTIOUS Report Numbers 50-237/74-60 & 61:

rings made of ;

Repairs of the feeduater check valves included installing new "0" a carbon fluoride compound called "viton". This material will withstand radiation as well as the environmental conditions of the 350~ F water to which it vill subjceted. Other repairs included lapping the valve discs. The valves were reassc= bled and tested. See Appendix "A" for the final leakage rates.

Report Humbers 50-237/74 62 & 64:

The vacuum breakers with excessive leakage had new erk fittings installed on the shaft seals. Both chaft seals on all the vacuum breakers were repacked with waterless grease. For the final leakage rate of all the breakers see Appendix "B".

Report Number 50-237/74-67:

The torus to drywell-purce, vent line was the only " butterfly valve" leak test to fail. In order to reduce this leakage and prevent any of the other butterfly valves from developing leaks, all were sent back to the manufacturer for rebuilding the seats. This included the following valves:

A0-1601-20A & B l0-1601-23

. A0-1601-56 A0-1601-21 A0-1601-24 A0-1601-60 A0-1601-22 A0-1601-55 A0-1601-63 The net seat installed was Resiloscal "',l" EPT Rubber. Soc Appendix "it" for the final leakage rates.

Report Number 50-237/74 65 & 6S:

LPCI valves MC-1501-18A & B and MO-1501-38A & B were found to have no prob 1 cms with leakage ence they were closed all the way. In order to ensure that the control room operator doesn't get a prematu*e " closed" signal, the limit suitches were checked and adjusted as required. See appendix "A" for the final leakage rates.

See the special note at the end of this letter concerning valve MO-1501-275 Report Number 50-237/74-72:

Electrical penetration X-202S was the only penetration of this type to exceed the technical specification limit. The repair of this penetration required dir-ection and supervision by General Electric Company. To prevent future leaks, as long as special equipacnt was required for X-202S, four other cicetrical penetrations uere scaled. These are X-203B, X-2043, X-200A, and X-2020. A special G.E. Co procedure uas used to insert more epoxy, thus scaling any existing cracks. See appendix "C" for the final leakage rates.

Report Numbers 50-?37/74-71 & 80 50-237/75-3 & 6 Valves 2302 45, A0-20;-1C & 2C, 205-2-7, and A0-1501-25B were all disassembled l and no major defects were found. All that was required to return these valves l l

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Mr. 'Jaraco G. Kopp1b V May 9, 1975 to normal condition was a seat cleaning and valve lapping. See Appendix "A" for the final leakage rates of valves 2301 45, 205-2-7, and A0-1501-25B. See Appendix "E" for the final leakage rate of valves A0-203-1C & 2C.

Report Number 50-237/74-75:

See the letter from Mr. B. B. Stephenson to Mr. J. G. Koppler dated December 20, 1974 Report Number 50-237/74-76:

The original bellows on penetration X-123 were re:aoved and cent back to the tanufacturer, Pathway Bellown Inc. They made up a new bellows seal by welding new bellows to the existing end plates. The new bellows ocal was then installed as a single unit. The process pipe was out at an existing weld between the two fluid heads. Thic allowed the bellows seal to be innta11ed without increasing the nu .ber of welds used in the original configuration. After all welding was complete, the velds were radiographed and the penetration was cueeecafully hydrostatie tested. See appendix "D" for the final leakage rate.

FAI11!BE DATA l

Valve Ntr.ber Identi fication Hintory 220-58A 18" check valve No previous D-2 failuren Crnne 97! DA failure (6/7h) 220-58B Same an above 2 Previous D-2 Failurco (6-71/6-72) D3 failure (6/74) 220 62A Same as above No previous D-2 failures D-3 failure (6/74) 22t '2B Sane as above 1 previeun D-2 Failure (6/72) D3 failure (6/7h) 1601-32A-F Check valve-Atwood No previous D-2 failuren

& & Morrill 601fBSPL No previouc D-3 failures 1601-3 3-P 1601-21 18" Eutterfly valve 1 previous D-2 failure H. Pratt 2 FII (6/72) No previous D-3 Operator: Mi'ller Hyd A 81 failure 1601-22 Same an above Same an above 1601-55 4" butterfly valve- H.Pratt No previous D-2 failuroc l 2FII Operator: I*lller Hyd No previous D-3 failurec ,

J-53 l l

1601-56 18" Butterfly valve H. Pratt 1 provicus D-2 Failure ,

2FII Operatcr: Miller Hvd (6/72) No previous '

A 81 D-3 failure i

, May 9, 1975 Mr. Jam::n G. Keppler, .

~

3 b d FAILURE DATA (cont'd) l l

l Identification IIintory Valve !! umber 6" Globe valve Crane- lio previous D-2 failures 1501-18A & B No previous N 3 failures 151 xR Operator:

limitoraue SIE00 14" Globe valve Crane- Same as above 1501-38A & B 7183v Operator: limitorque S!.N3 16" Gate valve Cranc- Same as above 1501-27B 33$- x R Operator: limitorque S' E0 12" electrical pent. type 1 previous D-2 failure I Penetration X-202S 4 pont. (6/71 X-202BB)No previous (

D-3 failtu'en ,

2301-45 16" check valve Miccion 1 previous D-2 failure l

Duo Check model B (6/71) lio previous .

l D 'i failure Bellown X-105c 36" Pathtmys Tandem Unllows No previoun failures l

l l Pellown X-123 12" Palhwayc-Single tellows Snme an above l

203-1C & 2C 20" Crane Y-Pattern Globo No previous D-2 failurcs valvo 2 previous D-3 failurca (6/74 20%1n 20%2n) 205-2-7 2}" D:o Check style B No previouc D-2 failures Minnion Mfc. No previoun D-3 failures 1501-25B 16" tectable check First time tested Atwood & Morrill due to svatem nod Hote: As per the telephone conversation with Mr. P. Johncon on Pay 6, 1975 at 1530 hours0.0177 days 0.425 hours 0.00253 weeks 5.82165e-4 months , this report will be cubmitted with the follouing local leak rate testa still pending.

l e

May 9, 1975 Mr. James G. Keppl {

Test Boundary System Penetration 'l Valvos 91-95 & 04 CRD Return X-144 Valves 301-90 & 09 CRD Return X-144 Valves 1501-27B & 28B LPCI X-145 Ecuiment hatch Primary Coitainc.ent X-100 Drywell head flante Irimary Containment Prior to Dresden Unit 2 startup, these tests will be completed and a supplementary report issued as soon as possible.

Sincerely,

' y Cr.B. B. Stephenson Superintendent EBS: r=p File /!!RC

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LOCAL LEAK RATE TESTS PERFORMED DURit:G THE U!1tT 2 REFUElit G *CUTAGE OF /77V- h n i G i i >
/J4/.k 25 . 71' /rejh,,
i TYPE OF PE4ETRAT10:1: .
!!41T I AL lillTIAL THRU FitAL Fl:AL .'h s.U i LEAK RATE . LEAKAGE LEAK RATE LEAMAGE [
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In'di. cates waterhead present on one side of valve
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