ML20087A319
| ML20087A319 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/08/1969 |
| From: | Carson R, Dodds R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20086U000 | List:
|
| References | |
| FOIA-95-36 50-219-69-05, 50-219-69-5, NUDOCS 9508040271 | |
| Download: ML20087A319 (8) | |
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U. S. ATOMIC ENERGY COMMISSION RP.GION I DIVISION OF COMPLIANCE Report of Inspection i
CO Report No. 219/69-5
.. f Licensee:
JERSEY CENTRAL POWER & LIGHT COMPANY
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' Oyster Creek -1 License No. DPR-16 Category B Dates of Inspection:
June 4 and 5, 1969 Dates of Previous Inspection:
May 27 and 28, 1969 Inspected by :
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R. T. Dodds, Reactor Inspector
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$M Reviewed by N
R. T. Carlson, Senior Reactor Inspector
' Dh t e' Proprietary Information:
None SCCPE Type of Facility:
Boiling Water Reactor Power Level:
5 Mwt (for low power physics tests - 1600 Mwt full term license)
Location:
Lacey Township, Ocean County, New Jersey Type of Inspection:
Routine, announced I
Accompanying Personnel:
None Scope of Inspection:
Review startup test results, review facility records, and tour of the facility.
SUMMARY
Safety Items - No items of safety significance were identified during the visit.
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,. Noncompliance Items - None.
g Unusual Occurrences - The roof on the refueling building was damaged i
on April 3, 1969, when the exhaust dampero closed and the building was pressurized by the intake fans.
(Section K.2.)
Status of Previously Reported Problems -
t 1.
The south (B) main steam isolation valve outside contain-ment was disassembled for inspection.
The leakage observed during the containment leak rate test appears to have been through the seat of the pilot valve.
(Section K.l.)
Other Significant Items -
1.
In-sequence rod notch worths appear to be about as expected.
(Section F.2. )
2.
The inner mechanical seals on two of the primary water re.:irculation pumps had to be replaced because they had been improperly assembled at the factory and subsequently damaged daring operation of the pumps.
The seal on the remaining three pumps will be inspected and replaced as required.
(Section E.1.)
3.
The results of the open vessel full flow vibration tests indicate that core component vibration was well within the acceptable criteria limits by a factor of about four or-better.
(Section G.l. )
Management Interview - The results of the visit were discussed with Messrs. McCluskey and Hess.
Since no items of noncompliance or significant safety problems hr.d been identified, no commitments were proffered by the licensee.
L A special interview was held with Messrs. Ritter and Calcec.
- Later, another interview was held with Mr. Calcec.
The purpose of the first interview was to acquaint Messrs. Ritter and Calcec with the results of the inspection experience with regsrds to the startup test program.
at Oyster Creek.
During the discussion, Mr. Ritter emphasized that JC was the responsible party for (1) ensuring Compliance with the license and (2) ensuring that the results of startup test program adequately demonstrate that the nuclear characteristics of the plant coinply with the Technical Specifications and that the plant can be operated safely.
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The purpose of the interview with Mr. Calcec was to acquaint him with the AEC's Regulatory organization.
Particular emphasis was placed on the functions of DRL and Compliance.
The reporting requirements
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contained in the license and the Technical Specifications were also i
discussed, i
I DETAILS A.
Personnel Contacted:
1 Principal personnel contacted during the visit included the following:
Jersey Central Power and Light Company (JC)
G. Ritter, Vice ? resident G. Calcec, Manager of Generating Stations T. McCluskey, Station Superintendent D. Hetrick, Operations Supervisor D. Ross, Technical Supervisor (In-Training)
J. Sullivan, Engineer I
J. Carroll, Shift Foreman General Electric Company (GE)
W. Hess, Site Operations Manager W.
Bibb, operations Superintendent D. Diefenderfer, Principal Test Design and Analysis Engineer R. Green, Instrument Engineer B.
Administration and organization 1.
Chemical Supervisor Since Mr. Doyle, Chemical Supervisor, has terminated his employment, the firm of Shepart T. Powell has been providing JC with a Chemical Supervisor until a permanent replacement can be hired.
The interim Chemical Supervisor is Mr. Swend Daljard, Radiochemist.
Mr. McCluskey stated that the consultant had already had a definite input on the radio-chemistry program by " streamlining" many of the test procedures.
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Reactor Operations 4
j Mr. Hess stated that the startup test program was on schedule.
The operational sources were installed and the regular source range j
monitors tested by May 12, 1969.
Control rod sequences have been
.i checked.
In-sequence rod notch worths appear to be about as expected.
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However, complete rod worth measurements cannot be made until higher power operation has been authorized.
The open vessel full flow vibration tests were run.
Test results show that core component vibration was well within the acceptable criteria limits by a factor of four to one or better.
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Reactor operations were suspended following the open flow vibration test to allow the removal of low temperature strain gauges from fuel channels and guide tubes.
High temperature vibration instrumentation l
was being installed on the core shroud and steam separator at the time of the visit.
Once these have been installed and tested, the i
reactor head will be installed and preparations made for high l
temperature operation.
In accordance with the present license, no reactor operation will be permitted with the head on until the license has been amended to permit full power operation.
E.
Primary System t
1.
Recirculation Pumps According to Mr. McCluskey, the inner mechanical seals on two of the reactor recirculation coolant pumps had to be replaced because the seals had been improperly assembled at
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the factory and subsequently damaged during operation of the pumps.
The seals on the remaining three pumps were in the
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process of being inspected and will be replaced if necessary.
F.
Reactivity Control and Core Physics 1.
Nuclear Instrumentation The dunking chambers were removed and the regular in-core source range monitors (SRMs) connected and tested for satisfactory operation after the five permanent Am-Be sources had been installed.
The four in-core SRMs indicated 30-41 cps after the permanent sources were installed.
The review of the discriminator curves showed that signal to noise ratio for all channels was greater than 100.
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i A results of the SRM performance tests show that the SRMs were operating in conformance with the acceptance criteria.
t A rough power level calibration was made of the intermediate
. j range monitors (IRMs).
They were satisfactorily tested and set to scram the reactor at 5E 2.5 Mwt (limit - 5 Mwt).
The peak operating power appears to have been about 1 Mwt.
2.
Nuclear Characteristics The results of the shutdown margin tests and control rod sequence were l'irst reviewed and discussed with Mr. Diefenderfet (GE) and the with Mr. Ross (JC).
The reactor appears to be shut down by about 3.5% 4 k/k with all rods in and by at least 0.5% Ak/k (probably pl% AK) with the strongest rod withdrawn at a temperature of 850 F.
According to Mr. Diefenderfer, the rod worths appear to be as advertised in the dispersed rod patterns.
The following is a summary of notch worths of the control rods that have been obtained to date:*
Sequence "A"
Rod No.
Position Notch Worth - % Ak/k 26-23 08 - 10 0.0314 10 - 12 0.0503 12 - 14 0.0824 26-39 08 - 10 0.0120
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10 - 12 0.0209 12 14 0.0431 Sequence "B"
Rod No.
Position Notch Worth - % /sk/k 14-23 10 - 12 0.0425 12 - 14 0.0667 30-23 10 - 12 0.0493
- Test data limited by low power operation.
More information will be obtained during heatup and higher power operation.
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Core and Internals i
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Core Vibration Measurements i
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The results of the open vessel core vibration measurements 1
that were completed on May 23, 1969, were reviewed and j
discussed with Mr. Diefenderfer.
The purpose of the test
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was to measure the vibration amplitudes or strains of selected reactor internal components and by comparing the observed amplitude with an analytically established acceptance criteria i
investigate the vibration integrity of the systems.
The comparisons indicate that all measured values were well within the criteria limits by a factor of about four or better.
The operating conditions were chosen to simulate those encountered during actual power operations as far as coolant flow excitation forces are concerned.
The flow per pump loop was 40,000 gpm for four pump operation and 37,000 gpm j
per loop for five pump operation.
The maximum pump speed achieved was that corresponding to a generator output frequency of 50 Hz.
The coolant temperature increased from 850 F to j
1100 F.
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The vibration displacement of the separator assembly and the shroud were measured using differential transformers to sense the radial motion of ti'ose parts with respect to the vessel at 900 azimuth intervals.
The motion of each recirculation i
loop suction line was sensed with an accelerometer mounted to respond to motion along the vessel radial direction.
The vibration strain in the fuel channels and guide tubes (2 each) were measured using weldable strain gauges.
Two strain gauges were mounted on each tube to sense axial strain.
Four gauges were mounted on each of the fuel channels (2 at center and 2 at upper quarter) also to sense axial strain.
i The following table lists the maximum vibration displacements or strains observed during the test.
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, Acceptance Criteria Maximum Vibration Amplitude (From Calculational Models)
Location Displacement, mils Frequency. Hz Displacement, mils Frequency 3Hz I. Separator 1.5 7.7 340 4.1 I
100 10.2 l
310 14.6 Separator 4.0 4.6 Bolts Loose Condition
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(No Criteria Given)
Shroud 0.9 12 1240 14.6 114 24 60 29 1.0 6-8 Bolts Loose Condition (No Criteria Given)
Suction Line 11.0*
12.5*
40 5.5 Strain, uin/in Frequency, Hz Strain, uin/in Frequency, Hz Fuel Channels 4.0 4.0 652 5.4 8.0 10.7 652 10 8.0 16.0 1770 20.6 6.0 19.0 Guide Tubes 22.0 18.0 178.0 16.6 178.0 17.6 K.
Containment 1.
Main Steam Isolation Valves Mr. Hess described the work being done on the south (B) main steam isolation valve outside containment to enable it to meet the leakage rate acceptance criteria.
The valve was dismantled during the course of the visit.
Prior to this, the valve was cycled 180 times.
It was tested for leakage after every 5-10 cycles.
Typically, the leakage decreased from an initial rate of 54 SCFH to about 20-25 SCFH, however, two instances were noted where the leakage rates were 52 SCFH and 92 SCFH.
The leak rate was also measured before cycling by increasing the closing air pressure on the valve piston.
It decreased from 42 SCFH to 28 SCFH when the driving pressuro was increased j
- from 98 psig to 125 psig.
- Discussion with GE revealed that these results had been reviewed with I
the co*gnizant GE structural engineer (Mr. C. S. Duckwald) and found l
to be acceptable.
Test results also reviewed for CO by W. J. Foley, l
Parameter, Inc. (CO Consultant) and found to be acceptable.
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! l The inspector checked the-valve after it had been dismantled.
Two small " nicks" were observed in the main seat, however,
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they did not appear to be located at the mating surface 0 and the plug at 44 ).
The l
I (i.e., the seat is beveled at 45 seat of pilot valve was clean, but several " blotch" areas j
fj were noted that appeared to correspond with shallow pits -
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(estimated to be several mils deep) that were in the mating
'q surface of the plug.
This then appears to be the source of 7'I the leakage and will be corrected by lapping or " dressing" the pilot valve plug.
Photographic slides were taken of the valve plug and seat.
l They will be kept on file in the Regional office.
l 2.
Secondary Containment on April 3, 1969, (before issuance of license and before the second reactor building inleakage test), the tar paper weather cover on the roof of the reactor building was damaged when the ventilation exhaust dampers closed and the building pressurized by the intake fans.
Except for around the drains and the parapet, the incident did not tear the tar paper, just lifted it.
According to Mr. McCluskey,.this was the second time this had happened.
The roof has since been repaired by replacing the tar paper.
In addition, a plastic foam caulking was put between the joints of the sheet metal roofing (planks of interlocking I
aluminum called "O"-decking).
It was believed that this should keep pressure from being exerted on the tar paper again.
i The tar paper acts as a sealant against the weather.
To preclude this type of occurrence in the future, the i
dampers have been interlocked to shut off the fan if the dampers should close.
Also, two high pressure interlocks have been installed to trip the fan if the reactor building pressure exceeds 1.2 inches of water.
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