ML20084K291
| ML20084K291 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/18/1974 |
| From: | Kalivianakis N COMMONWEALTH EDISON CO. |
| To: | Oleary J US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20084K292 | List: |
| References | |
| AO-50-254-74-8A, NJK-74-163, NUDOCS 8305190259 | |
| Download: ML20084K291 (5) | |
Text
J m
Comkwealth Edison b
Quad-Citi;s Generating St tion Cordova. Ilknois 61242' Post Office Box 216 Telephone 309/654-2241 s
NJK-7 163 N
J
$L g,69gr Cf,L July 18, 1974 0 $h
- p W
s Mr. John F. O ' Leary, Director Directorate of Licensing Regulation Q w'.,f U. S. Atomic Energy Conunission Washington, D. C.
20545
Reference:
Quad-Cities Nuclear Power Station, Docket No.
50-254, DPR-29, Appendix A, Sections 1.0.A,2, 4.7.A.2. i.(2)(b), 4.7.A.2.1 (2)(c), and 6.6.B. I.a
Dear Mr. O ' Leary:
Enclosed please find Abnormal Occurrence Report Numbers A0-50-254/74-8a through 81 for Quad-Cities Nuclear Power Station. These occurrences were reported to Region ill, Directorate of Regulatory Operations by telephone, and to you and Region lit,
Directorate of Regulatory Operation by telegram on the following dates:
A0 254/74-8a C and D Main Steam Line April 1, 1974 isolation Valves A0 254/74-86 0xygen Analyzer Valve April 10, 1974 A0 8802B A0 254/74-8c Drywell Equipment Drain April 12, 1974 Sump Discharge Valves A0-1 -2001 -15 and AO 2001 -
16 A0-50-254/74-8d RCIC turbine Exhaust Check April 17, 1974 Valve I-1301-41 A0 254/74-8e Feedwater Check Valves l-April 25, 1974 220-58A, 588, 62A, and 62B A0 254-74-8f Drywell Personnel Access May 17, 1974 A i r-Lock A0 254/74-89 Drywell and Torus Purge May 21,1974 Valves A0 254/74-8h Reactor Water Cleanup Dis-July 5, 1974 charge Valve H0-1-1201-80 A0 254/74-81 Reactor Water Cleanup Iso-July 9,l1974 lation Valve H0-1-1201-5 L/
/
8305190259 740718 1
gDRADOCK 05000254
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PDR l
b d ^11 COPY ENT REGION 7 2
M r. John F. O 'Lccr July 18, 1974 On July 17, 1974, local leak rate testing in accordance with Technical Specification 4.7.A.2.h was completed on Unit 1.
The following satisfactory results were obte.ined:
TECH SPEC TOTAL LEAKAGE ITEMS LEAKAGE LIMIT (SCFH)
(SCFH)
Double-Gasketed Seals 36.72 7 031 Testable Penetrations 110.16 3.35 (with isolation valves)
Any One Penetration or 18.36 12.86 Isotation Valve (worst case)
Any One Main Steam Iso-11.5 10.95 lation Valve (worst case)
Drywell Personnel Air-18.36 0
Lock Containment Isolation 110.16 93.175 Valves (with testable penetration)
(total)
This composite report of the detaFis of the repairs necessary to correct abnormal leakages is submitted to you in accordance with our agreement of April 4,1974, and is submitted in accordance with the requirements of Technical Specification 6.6.B.I.a.
Very truly yours, COMMONWEALTH EDISON COMPANY QU
-C ITl NUCLEAR POWER STATION
/
s
~VW N
. Kalivianakis S
tion Superintendent i
NJK/LFG/Jeh cc: Region ill, Directorate of Regulatory Operations J. S. Abel I
l l
9 9
REPORT NUM8ER: A0 254/74-Qa REPORT DATE: July 18, 1974 OCCURRENCE DATE: April 1, 1974 FACI LITY: Quad-Citles Nuclear Power Station Cc,rdcsa, I llinois 61242 IDENTIFICATION OF OCCURRENCE:
Main steam isolation valve local leak rate test results showed excessive leakage on steam lines C and D.
CON 0lT10NS PRIOR 'T0 OCCURRENCE:
Reactor in the Hot Shutdown Condition, zero reactor pressure, MSIV's closed, in process of slowly raising reactor water level prior to vessel head removal.
DESCRIPTION OF OCCURRENCE:
On April 1,1974, local leak rate testing was being performed on the Unit One main steam isolation valves (MSIV 's). Prior to the test, the reactor pressure was re-duced to zero and all eight MSIV's were placed in the closed position using the j
control room switches. The reactor water temperature was slightly above 200 F at j
the time of the leak rate test.
The leakage measurements were taken by pressurizing the volume enclosed by the in-board and outboard MSlV's, through a pressure test connection, to a pressure of 25 psig. Leakage measured through the "A" steam line valves was determined to be 5.5 SCFH; therefore, the Technical Specification limit of 11.5 SCFH was met for both valves in that particular steam line.
When an attempt was made to pressurize the "B" steam line volume between the inboard and outboard valves, an obstruction in the pressure test connection prevented air i
from entering the volume. It was decided at the time to go on to the other steam lines, and re-test the "B" steam line at a later date. A work request was initiated to clear the pressure test connection so that a test could be performed.
The "C" steam line was pressurized between the inboard and outboard MSlV's, and the leakage was measured to be_168 SCFH at 25 psig test pressure. A similar test on the "D" steam line yielded a leak rate of 48 SCFH. Both of these leakage values exceeded the limit of 11.5 SCFH for any one MSIV.
On May 13 the reactor water level was increased to flange level, and re-testing of the MSIV's conenenced so as to determine which valves on the "C" and "D" steam lines were actually leaking excessively. Also, testing was required of both of the "8" steam line isolation valves. Since the head of water inside the inboard MSIV's was greater than 25 psig equivalent pressure, the leakage to be measured through the MSIV's with the reactor water level at the flange would be through the outboard MSIV 's only. Leak rate measurements were made at a 25 psig test pressure witti the following results:
A0 254/74-Ba July 18, 1974 y,Akyi LEAKAGE (SCFH) 1-203-2s 10.95 1-203-2C 216.0 1-203-20 34.7 Valves 2C and 2D exceeded the 11.5 SCFH limit, and work requests were initiated lunadiately to repair these valves. The leakage through 28 was acceptable, and no further action was taken with regards to this valve.
On May 14 the reactor water level was lowered to below the steam inlets to the vessel. Leak rate measurements were then re-taken on the B, C, and D steam lines to determine the total leakage through both MSIV's on each lloe. This total minus l
the leakage measured through the outboard valves would equal the leakage through l
the inboard valves. The leakages through the inboard M51V's was determined to be as follows:
VALVE LEAKAGE (SCFH) 1-203-18 5.18 1 -203-IC 0.0 1-203-10 34.7 Valves 18 and IC showed acceptable leakage valves, but ID exceeded the leakage limit. A work request was initiated to repair the valve in the same manner as valves 2C and 20 To summarize, valves A0-l-203-ID, 20, and 2C exceeded the Technical Specification leakage rate criteria. All other MSIV's had acceptable leakage values and no further corrective action was deemed necessary.
l DESIGNATION OF APPARENT.CAUSE OF OCCURRENCE:
Upon disassembly of the main steam isolation valves which experienced excessive leakage, minor surface defects were noticed on the main valves' pilot stem. The pilot stem surfaces on the ID and 20 valves showed significantly less defects compared to that of the 2C valve.
A Crane Valve Company representative was present during the valve repairs. The cause of the leakage was attributed to these surface defects. It is postulated I
that hard valve closure during Group 1 isolation conditions may have caused the main valve pilot stem surfaces to become defective.
The main valve seating surfaces were closely examined and no excessive corrosion or surface marks were observed. Therefore, it has been determined that the major cause of the excessive leakages on the three main steam valves was the presence of surface defects on the valve pilot stems.
ANALYSIS OF OCCURRENCE:
In the case of the 2C main steam isolation valve, although its leakage measured 216 SCFM, the leak rate through the inboard valve on the C steam line was deter-mined to be 0.0 SCFH. Therefore, the total leakage through this particular steam line from the primary containment would be minimal under accident conditons.
r p
A0-50-254/74-8a U July 18, 1974 The total leakage possible through the D steam line would be 34.7 SCFH in the event of a steam line break outside the primary containment. This number is less than the composite leakage that would take place through the other three steam lines if the leakage through sach were equal to the maximum valve of 11.5 SCFH.
The maximum total leakage through the A, B, and C main steam lines as determined by the local leak tests would be 10.68 SCFH, and this value is less than the limit for anv one valve.
The excessive leakage rates did not in any way render the MSlV's inoperable, nor was the ability of the valves to perform their design function affected. U pon receipt of a Group i isolation signal, these valves would have shut in the re-quired time, and performed the isolation function.
During normal plant operation, the MSIV's are open. Thus, the excessive leakage measured whlie the valves were closed and while the unit was shutdown, did not affect the safe operation of the plant, nor did the leakages present jeopardize the health and safety of the public.
CORRECTIVE ACTION:
All three main steam isolation valves were disassembled completely. The main valve pilot stems were machined to remove all defects. The valve seating surfaces were lapped under the observation of a vendor representative.
Following re-assembly of each Individual valve, a leak test was performed at a test pressure of 25 psig. The following satisfactory test results were obtained:
VALVE DATE LEAKAGE (SCFH)
A0 203-2C 7/14/74 8.64 A0 203-I D 6/03/74 5.76 A0 203-2D 7/ 08/74 0.0 FAILURE DATA:
1 The MSIV's had not been tested for leakage since the pre-operetional tests on November 27 and 28,1970. The leakage rates through valves A0-1-203-ID, 2D, and 2C were 8.2 SCFH, 0.0 SCFH, and 2.0 SCFH, respectively at that time.
Unit 2 MSIV's 2-203-18 and 28 were leak tested on April 27, 1974, during an in-vestigation of missing parts on electromatic relief valve 2-203-3E. The measured leakage was 52.4 SCFH, and repairs were initiated.
in this case, the inboard valve had two deep gashes on the pilot stem which was the major cause of the leak-age. The pilot stems were machined and seats lapped on both valves. A subsequent leak rate test was performed and the measured leakage was 3.25 SCFH.
No other failure data on Quad-Cities Unit 1 or Unit 2 MSIV leakages has been ob-tained.
.