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l-April 9, 1970 J. P. O'Reilly, Chief, Reactor Inspection.& Enforcement Br.,
Division of Compliance, Headquarters i
INQUIRY MEMORANDUM JERSEY CEFFRAL POWER & LIGHT COMPANY (OYSTER CREEK 1), 219/70-D CONTROL ROD DRIVE MALFUNCTION During a telephone call to Oyster Creek 1 by the assigned CO:I inspector on April 8,1970, Station Superintendent, Tom McCluskey stated that following a "LW Pressure.in Main Steam Line" scram I
at 10:16 a.m. on April 7,1970, three control rod drives (coordinates 14-35, 18-35 and 42-27) were observed to be-each
.a at notch positions 02 (6 inches out).
All other drives were-determined to be at 00 notch position (fully inserted).
The scram was stated to have been' normal in all other respects; the reactor was shutd wn.
Other pertinent facts stated were:
1.
The.three drives were subsequently inserted to the 00 notch position by the console operator.
2.
The three. drives were scram tested-once and all times were determined to be within the technical specification require-ments; the drives operated normally during the scram l test.
(These drives were not in the group of drives monitored by i
the automatic scram recorder at the time of the scram.)
The following scram test data was given o
Scram Insertion Times (Seconds)
Coordinate 10%
50%
90%
i 14-35 0.40 1.32 2.41 18-35 0.39 1.25 2.28 42-27 0.40 1.25 2.26 NgEE:
The reactor pressure was stated to have been~600 psig S
for the above-test.
These drives are not listed in the. group f
of eight drives with inner filters.
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3.
Gerteral Electric personnel at the site and San Jose reviewed the matter, It is postulated that two things may have contributed to cause the problem (a) dirt in the last buffer orifice hole, or (b) leakage around the e, top piston seal.
It is postulated that the drives actually went past the 00 notch position on the scram, then settled out to the 02 notch position.
4.
It was stated that no scram reset took place until approx-imately 15 minutes following the scram.
5.
Continued close surveillance will be maintained for similar type problems in the eient of any future scrams.
The plant operations review coesnittee is further evaluating other appropriate surveillance requirements.
6.
The reactor subsequently was made critical at 5:31 p.m. on April 7,1970, and is now operating at full power.
The assigned inspector plans to review the mate,ars in detail during the next regular inspection visit.
Mr. McCluskey's position is that this matter is not reportable in accordance with the terms of the license.
CO:I has requested a written report concerning this control rod incident.
R. T. Carlson, Senior Reactor Inspector cc E. G. Case, DRS R. S. Boyd, DRL (2)
- 8. Levine, DRL D. J. Skovholt, DRL (2)
L. Kornblith, Jr., CO Regional Directors, CO REG File
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3944 1
l April 17, 1970 I
J. P. O'Reilly, Chief, Reactor Inspection & Enforcement Br.,
L Division of compliance, Headquarters INQUIRY MEMORANDUM JERSEY CENTRAL POWER & LIGHT COMPANY (OYSTER CREEK 1), 219/70-E CONTROL ROD DRIVE MALFUNCTION Mr. Tom McCluskey, Station Superintendent, telephoned Region I.
at 10:30 a.m. on April 17, 1970, and stated that on April 14, 1970, at 5:22 a.m. the reactor scrammed from a reactor low level.
j One control rod drive 18-35 stopped at notch position 02 (6 inches out).
He stated that the drive was monitored during the scram I
by the control rod drive recorder and the scram insertion times conformed to the requirements of the technical specifications.
The cause of the malfunction is believed to be stop piston seal leakage.
Other pertinent facts stated were:
1.
This is the same drive that similarly malfunctioned during a scram on April 7, 1970.*
This. drive and the two others that had previously malfunctioned during the April 7th scram were subsequently scram tested several times and all three drives stopped at the 02 notch position during the test, however, conformed to required technical specification insertion times.
2.
It was stated that control rod drive seal leakage was generally increasing with other drives.
Fifty-five (55) drives were stated to have seal leakage of sufficient magnitude to present a problem during rod withdrawals.
This was stated to be an increase of 17 drives in one week.
(Note - Seal leakage does not prevent the scram. function).
All drives are being exercised daily (one notch in, one notch out) to prove operability.
3.
A second control rod drive has been declared inoperable (six is the maximum number of inoperable drives allowed by technical specifications for continued ooeration).
- see Inquiry Memorandum, 219/70-D, l
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4.
The scram was stated to have resulted from testing and adjustments being performed on steam turbine controls to correct a steam flow oscillation.
5.
The reactor was started up following the scram, and the turbine was back on line at 7:28 p.m. on April 14, 1970.
The reactor is now operating at full power.
6.
An extended outage (approximately one month) is tentatively scheduled for Monday, April 20, 1970.
It was stated that control rod drives would be resnoved, inspected and repaired during the outage.
It is planned to install 10 mil inner filters on each control rod drive during the outage.
Control rod drive seals will be replaced.
Mr. McCluskey stated that both stop piston and collet piston seals are among the leaking drive seals to be replaced.
He stated that GE suspects that seal springs mey havre failed.
The licenses plans to submit a written report to DRL on both the April 7 and 14 occurrences.
The assigned inspector is planning a. sitet inspection early in
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the shutdown to review the entire matter.
R. T. Carlson Senior Reactor Inspector con E. G. Case, DRS R. S. Boyd, DRL (2)
- 8. Levine, DRL j
D. J. Skovholt, DRL (2) 1 Regional Directors, CO L. Kornblith, Jr., CO RBG File
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3944 April 30, 1970 J. P. O'Reilly, Chief, Reactor Inspection & Enforcement Br.,
Division of Compliance, Headquarters INQUIRY MEMORANDUM JERSEY CENTRAL POWER & LIGHT COMPANY (0YSTER CREEK 1), 219/70-F LINEAR INDICATIONS IN WELD OVERLAY ON CORE SPRAY N0ZZLE SAFE END Region I received a telecon from T. McCluskey, Station Superintendent, Oyster Creek 1, on April 29, 1970, in which he related the following facts regarding linear indications observed in the veld overlay on the north core spray system reactor vessel nozzle safe end:
1.
In light of the corrent safe end problem at Nine Mile Point, and because of the opportunity currently available (0C-1 in extended shutdown to investigate and correct control rod drive problems),
JC decided to examine the accessible surfaces (exterior) of the safe ends on their two core spray system reactor vessel nozzles.
2.
During the examination (perforned on April 26, 1970), several (2-3) linear indications were noted in the veld overlay on the north nozzle safe end.
3.
These indications were observed on the bottom half of the safe end at the 6-7 o' clock position looking at the vessel. More specifically, they are located in the overlay where it tapers into the field weld joining the safe end to the 6" x 8" reducer.
(See Figure I) 4.
The longest indication is + 3/8" long. The indications are not in any particular pattern.
5.
A " slight indication" was also noted on the south core spray nozzle safe end but this one was removed with bne light grinding.
6.
Per Mr. McCluskey, the indications on thenorth nozzle safe end have every appearance of being weld metal type defocts. He stated that they made ene light grind on the indications before he stopped the work to allow for a more thorough examination of the conditions. The indications remained following this grinding.
7.
Representatives of GE (San Jose), MPR (JC consultants) and CPU are scheduled to be at the site on April 30, 1970, to pursue this matter further.
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Mr. McCluskey stated that he was communicating this information to CO because of the known Regulatory concern in this area.
l Following telecon consultation with you on April 29, 1970, JC was informed that arrangements have been made for A. Holt. DRS, and G. W. Reinmuth, CO, j
to also be present at the site on April 30, 1970, to participate in the i
further examination of these condFtions'.'
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R. T. Carlson, Senior Reactor Inspector Rnclosure:
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l cc: E. G. Case, DRS R. S. Boyd, DRL (2)
S. Levine, DRL
-D. J. Skovholt, DRL (2) i L. Kornblith, Jr., CO Regional Directors, CO REG File t
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OYSTER CREEK 1 CO INQUIRY-MEMORANDUM 219/70-E' i
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.l NOTES.
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1.
Per McCluskey, original intent at time of effecting repairs to reactor vessel safe ends (Amendment 43) was to remove original (sensitized) safe end in its entirety. This was not feasible; 1
therefore, it was replaced only in part.
2.
Per McCluskey, 308L weld overlay was to have been limited to the remaining length of sensitized safe end only; whereas, this i
investigation revealed it to have been applied to replacement j
safe end as well.
s FIGURE I Description of North Core Spray Nozzle Safe End (As Interpreted from 4/29/70 Telecon)
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