IR 05000219/1989028
| ML20006A982 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/17/1990 |
| From: | Collins E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20006A981 | List: |
| References | |
| 50-219-89-28, NUDOCS 9001310106 | |
| Download: ML20006A982 (20) | |
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Report No.
50-219/89-28 Docket No.
50-219 License No.
OPR-16 Priority --
Category C Licensee:
GPU Nuclear Corporation 1 Upper Pond Road Parsippany, New Jersey 07054 Facility Name: Oyster Creek Nuclear Generating Station Inspection Conducted:
November 5, 1989, - December 2, 19p Participating Inspectors:
M. Banerjee, Resident Inspector E. Collins, Senior Resident Inspector D. Lew, Resident Inspector Approved By:
go I kMb ifn f'lo Date mo E. Collins, Acting Chief
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Reactor Projects Section 4B Inspection Summary:
Inspection November 5 - December 2,1989 (Report No. 50-219/89-28)
Areas Inspected: The inspection consisted of 160 hours0.00185 days <br />0.0444 hours <br />2.645503e-4 weeks <br />6.088e-5 months <br /> by resident inspectors.
The areas inspected included observation and review of plant operational events (1.0), Notification of Unusual Event (2.0), Hydrogen /0xygen Monitoring System modification (3.0), a control rod uncoupling event (5.0), increased hydrogen leakage from the main generator (7.0), surveillance observation (8.0),
maintenance observation (9.0), and previously opened inspection findings (11.0).
Results: Overall, the plant was operated in a safe manner.
Two significant powcr reductions occurred during the period. One power reduction occurred as a result of environmental concerns which were raised when the dilution plant was lost. Another power reduction occurred as a result of low intake water levels.
A Notifification of Unusual Event was declared for the low intake water level.
The inspectors identified that the licensee does not include the containment leak rate test connection in their Appendix J program. This issue 'was lef t unresolved pending the licensee's evaluation of the need for type C testing of this valve.
Eight previously opened items were closed.
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TABLE OF CONTENTS P,ait 1.0 Plant Operations Revi n (71707)!................
1.1 Chronology of Operation Events...............
1.2 Control Room Tours...............
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1.3 Facility Tours.......................
2.0 _ Unusual Event (93702)......................
3.0 Hydrogen /0xygen Monitoring System Modification (71707).....
4.0 Post Accident Sample Sy> tem (71707, 93702).......
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5.0 Uncoupled Control Rod (71707, 93702)...............
6.0 Loss of MCC 1A22 Event (93702)...'...........-...
.9 7.0 Hydrogen Leakage from the Main Generator (71707, 93702)...,
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8.0 Monthly Surveillance Observation (61720)...,...... -..
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9.0 Monthly Maintenance Observation (62703)
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10.0 Observation of Physical Security (71707),....... -
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11.0 Previously Opened items (92701, 92702)............-,
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12.0 Inspection Hours Summary (71707)................
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13.0 Exit Meeting and Unresolved Items (30703)............
ATTACHMENTS Attachment I:
List of Personnel Contacted
- Say that these numbers refer to inspection procedures (modules).
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DETAILS 1.0 Plant Operational Review 1.1 Chronology of Operational Events
At the beginning of this inspection period the plant was operating at 200*4 power.
The plant had just completed its fortieth day of cor.tinuous operation with the turbine on line. Two technical specification action statements were in effect.
T5e "B" channel of
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the Hydrogen /0xygen Monitoring System was out of service because one of~the containment isolation valves located inside containment failed shut. A modification was being implemented to provide an alternate path for the "B" channel to sample the containment atmosphere.
Technical specifications allow continued plant operation for 30 days with one channel out of service.
The "B" channel was out of service since 10/26/89. The Standby Gas Treatment System 1 (SGTS) was out of service since 11/4/89 because of a faulty differential pressure gauge.
Parts were on order to replace the gauge.
Technical specifications allow continued plant operation for seven days with one SGTS out of service.
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11/7/89 The faulty differential pressure gauge was replaced on
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SGTS 1.
A system operability test was performed satisfactorily; and SGTS 1 was declared operable.
The seven day technical
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specification action statement was terminated.
11/12/89 While performing weekly Surveillance Procedure
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617.4.002, " Control Rod Drive Exercise and Flow Test / Inservice Test Cooling Water procedure", control rod 06-23 became uncoupled.
The control rod was recoupled; and, a coupling check satisfactorily performed.
Details of this event are described in paragraph 5.0.
11/15/89 While performing Station Procedure 607.4.004,
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" Containment Spray and Emergency Service Water System 1 Pump Operability and Inservice Test", the differential pressure for emergency service water (ESW) pump 52A exceeded its high action limit.
ESW System I was declared inoperable. Technical specifications allow plant operation to continue for seven days with this system out of service.
11/16/89 The dilution pumps tripped because the "A" phase of
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the power supply to the dilution plant became grounded.
All dilution pumps were out of service.
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11/17/89 Reactor power was reduced to approximately 40 percent
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because of environmental concerns that operation without dilution pumps would cause excessive temperature differences between the intake and discharge canals. With reactor power reduced to 40 percent ard no dilution pumps operating, the
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differential temperature between the intake and discharge canals was equivalent to that which would be observed if reactor power were 100 percent and dilution pumps were operating, Plant Engineering evaluated the Inservice Test data for ESW pump 52A. The original ESW pump 52A baseline was determined to be incorrect because the data were taken when the "ESW System 1 to Service Water System" check valve was leaking. The valve has since been replaced.
New baseline data was collected; and, ESW System I was declared operable.
The seven day technical specification action statement was terminated.
11/19/89 The grounded cable to the dilution plant was replaced;
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and it was returned to service.
The plant was returned to 100 percent power.
Modifications to the "B" channel of the Hydrogen /0xygen Manitoring System were completed.
A syste::i operability test was performed satisfactorily; and, the
"B" channel was declared operable. The 30 day technical specification action statement was terminated.
Details of the review of '.his modification are described in paragraph 3.0.
11/21/89 An Unusual Event was declared as a result of low
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intake water level. The low levels were caused by sustained high winds from the west. To mitigate the worsening low water level condition, reactor power was reduced to approximately.40 percent which allowed dilution and circulation water pumps to be secured.
When intake water level began to recover, reactor power was increased.
Details of this event are described in paragraph 2.0.
11/22/89 After intake level had reraered and stabilized, the
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Unusual Event was terminated.
11/24/89 Reactor power ascension was completed. The reactor
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plant was returned to 100 percent power.
11/28/89 Reactor power was decreased to accommodate the
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calibration of the controller for auxiliary flash tank level control valve, V-4-103.
Reactor power had to be reduced to minimize the rate of water input to the auxiliary flash tank while V-4-103 was out of service.
Minimum reactor power was 88 percent.
11/29/89 Reactor power was returned to 100 percent.
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Panel IM-175 was returned to motor control center.1A22, its normal power supply.
Because normal reactor coolant system leakage detection capabilities were lost during the evolution,
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alternate means of detection were utilized. The plant entered a seven day technical specification action statement during this evolution. The action statement was terminated after 20 minutes when the evolution was completed.
1.2 Control Room Tours Routine tours of the control room were conducted by the inspectors during which time the following documents were reviewed:
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Control Room and Group Shif t Supervisor's Logs;
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Technical Specification Log;
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Control Room and Shift Supervisor's Turnover Check lists;
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Reactor Building and Turbine Building Tour Sheets;
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Equipment Control Logs;
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Standing Orders; and,
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Operational Memos and Directives.
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No unacceptable conditions were identified.
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1.3 Facility Tours
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Routine tours of the facility were conducted by the inspectors to make an assessment of the equipment conditions, personnel safety, and procedural adherence and regulatory requirements. The following areas were among those inspected:
Turbine Building
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Vital Switchgear Rooms
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Cable Spreading Room
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Diesel Generater Building
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Reactor Building
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New Radwaste Building
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Old Radwaste Building
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The following additional items were observed or verified:
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Fire Protection:
Randomly selected fire extinguishers were accessible and
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inspected on schedule.
Fire doors were unobstructed and in their proper position.
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Ignition sources and combustible matsrials were controlled
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in accordance with the licensee's approved procedures.
Appropriate fire watches or fire patrols were stationed
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when equipment was out of service.
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Equipment Control:
Jumper and equipment mark-ups did not conflict with
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technical specification requirements, j
Conditions requiring the use of jumpers received the prompt
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attention of the licensee.
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Vital Instrumentation:
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Selected instruments appeared functional and demonstrated
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parameters within Technical Specification Limiting
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Conditions for Operation, d.
Housekeeping:
Plant housekeeping and cleanliness were in accordance with
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approved licensee programs.
Minor housekeeping deficiencies which were identified were promptly corrected by the licensee. No other unacceptable conditions were identified.
2.0 Unusual Event On 11/21/89, the licensee declared a Notification of Unusual Event (UE) as a result of low intake water level. Sustained winds greater than'40 miles per hour from the west caused the intake level to drop below one foot below mean sea level (MSL).
Oyster Creek's emergency classifications require a UE to be declared if intake level drops to negative one foot.
To mitigate the worsening low intake water level condition, the licensee reduced reactor power to approximately 40 percent. With power reduced to 40 percent, the licensee was able to secure two of four condenser circulation pumps and all three dilution pumps. This action minimized the amount of water which was diverted away from the intake structure.
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Minimum intake water level had dropped to just above the negative two feet mark. Oyster Creek's emergency classifications would require an Alert to be declared if intake water level were to drop to negative two feet.
The intake water levels stabilized and started to slowly trend up later in the day. Although intake water level was slightly above negative one foot, the licensee elected not to terminate the UE. This decision was i
made to avoid a situation where a UE would have to be redeclared if the intake level dropped below negative one foot again. An action plan was formulated which specified that the UE would be de-escalated when both the north and south intake water level readings were greater than or equal to 2,25 psig with four condenser circulation pumps and two dilution water pumps operating. An intake level at MSL would correspond to 2.25 psig.
At 2105 on 11/22/89, the north and south intake readings were 2.5 and 2.8
psig respectively.
The Unusual Event was terminated.
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The licensee's power ascension to 100 percent was complicated by environ-mental concerns. As a result of warm water, thousands of fish reside in i
the discharge canal.
The licensee's environmental group determined that
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l the discharge etnal temperature must be maintained above 50 degrees to prevent a potential fish kill.
The licensee adjusted power to control
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discharge canal temperature to prevent a fish kill and determined that prior to starting a dilution pump the condenser discharge temperature must be above 63 degrees F.
Power ascension was completed without a fish kill.
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The licensee raised a question of the promptness of the declaration of the.
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UE.
During the few hours prior to the declaration, the intake level had fallen below negative one foot.
The Group Shift Supervisor (GSS) did not declare a UE because clogging of the intake grates contributed to the low level. As the grates were raked and cleaned, the intake level recovered above negative one foot. A UE was declared by the GSS when it was evident to him that level was decreasing and would remain below one foot regardless of whether or not the grates were raked and cleaned. Based
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upon their review of these circumstancer., the licensee concluded that the GSS response and declaration of the UE was reasonable.
The inspector had no questions on operator response.
During this event, the licensee identified that there was an eror in the i
Containment Spray System Procedure.
The procedure showed that tae
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pressure corresponding to an intake water level of negative three feet (level at which a Site Area Emergency is declared) was 1.40 psig. The actual pressure should be 0.94.
The licensee intends to correct this typographical error.
The licensee's response to this event was good. Their action plan to terminated the Unusual Event and return the plant to 100 percent power was comprehensive and well planned.
No unacceptable conditions were identifie.
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3.0 Hydrogen /0xygen Monitor System Modification The "B" channel of the Hydrogen /0xygen (H2/02) Monitor System was removed
from service on 10/26/89 for corrective maintenance to the oxygen analyzer portion. However, upon completion of the corrective maintenance, the system could not be returned to service because the drywell inlet sample line isolation valve, V-38-41, located inside the drywell could not be opened. The licensee prepared a plant modification to provide for an
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alternate drywell inlet sample line to the monitor. The licensee also
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entered a 30 day technical specification action statement beginning 10/26/89 due to the inoperable "B" channel.
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The modification included disconnecting the existing sample supply and return lines at the outboard isolation valves. The supply line to the
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monitor was then connected to the existing return line.
Sample return to the drywell was rerouted via penetration x-67, an existing penetration used for integrated leak rate test (ILRT) purposes.
This penetration
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consists of a manual isolation valve and a screwed cap outside the drywell. The modification removed the cap and added two new solenoid t
operated isolation valves in series beyond the manual valve. Valve position indication was provided in the control room for these two new valves. Tubing was run from the valves to the monitor for sample return.
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The licensee considers this modification as a temporary configuration change and intends to return the system to its original configuration in the next refueling outage.
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s The inspectors reviewed the safety evaluation prepared by the licensee on this modification.
The safety evaluation addressed the sample point adequacy, representativeness of the sample with regard to the bulk drywell condition and adequacy of containment isolation.
No unacceptable conditions regarding the adequacy of the safety evaluation were identified.
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The inspector noted the manual isolation valves on the ILRT penetrations (V-38-92 and V-38-95) were not included in the Appendix J type C testing program.
No exemption from Appendix J exists for these valves.
The licensee indicated these valves were probably treated as vent / drain / test line connections. Acceptability of not leak testing the ILRT line isola-tion valves per Appendix J type C requirements is an unresolved item pend-ing review of the licensee's evaluation of the need for such a test.
(50-219/89-28-01).
After the modification was completed, an independent field walkdown performed by Technical Functions identified a cable separation deficiency,
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and a material non-conformance report (MNCR) was generated to resolve the issue.
The deficiency was corrected before the "B" channel of the H2/02 Monitoring System was returned to service on 11/19/89.
1he licensee's critique of this deficiency concluded that the supervisors and craft doing the work did not read the drawings properly. An incomplete package at the onset of the job, subsequent revisions to the modification package, and difficult to read drawings that led to varied interpretations s
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also contributed to the installation deficiency.
In addition, as portions of the job package were completed, Qualtiy Control (QC) walkdown was performed immediately following or concurrent with the walkdown performed by construction. Also discussed during the critique were the pressures associated with rapidly completirig the job.
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The inspector observed the critique meeting and reviewed the resulting corrective actions.
No unacceptable conditions were identified.
4.0 Post Accident Sample System Inspection Report 50-219/89-27 documented an event where the reactor coolant sample valve, V-551-198, for the Post Accident Sample System (PASS) was found shut. As a result of this event, two unresolved items were identified. One item (unresolved item 89-27-02) was left unresolved pending the licensee's completion of their critique on the event and NRC review of the corrective actions.
The other item (unrescived item 89-27-03) was left unresolved pending NRC review of the licensee's datermination of reportability.
During this inspection period, the licensee's critique on this event was completed. The root cause was determined to be inadequate task planning by Startup and Test Department (S/UT).
During the functional test of the Electrochemical Corrosion prevention Monitoring System modification on June 12, 1989, V-551-198 was shut. The test package was developed while V-551-198 was under Startup and Test jurisdiction. When it was rereviewed for implementation in June,'Startup and Test failed to identify that the test procedure required S/UT to manipulate an Operations Department valve.
The critique also identified that the system prints which were used to troubleshoot this system were uncontrolled and, therefore, not properly updated. This deficiency hampered the troubleshooting efforts when a reactor coolant sample could not be taken.
This deficiency, however, did not contribute to the inadvertent shutting of V-551-198, The critique identified several long term corrective actions to be implemented. These corrective actions include: (1) review of Startup and Test Procedures to ensure adequate preparation and review of valve lineups for functional test procedures, (2) review of Chemistry Department drawirigs to ensure that the drawings needed are controlled and staged in necessary locations, and (3) incorporation of the critique into the required reading program for Startup and Test, and Operations Departments.
The inspector reviewed the critique and determined the corrective actions to this event were appropriate.
Unresolved item 89-27-02 is closed.
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The inspector reviewed the licensee's determination of the reportability of this event.
The licensee stated that for analyzed accidents as
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described in the Updated Final Safety Analysis Report (FSAR), a sample which would be representative of reactor water chemistry could be taken without exceeding the time limits and individual exposures specified in
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the FSAR.
In regard to the question of obtaining a sample during an anticipated transient without scram (ATWS) event, tht: licensee stated that with their implementation of ATWS mitigating systems, the radiation levels inside the reactor building would be low enough for personnel access. A reactor coolant sample could be taken through the ncrmal sample path without exceeding the times and individual exposures specified in the FSAR.
The inspector concluded that the licensee's determination of reportability was appropriate.
Unresolved item 89-27-03 is closed.
5.0 Uncoupled Control Rod On 11/12/89, during routine control rod exercise tests, control rod 06-23 failed its coupling check. The control rod was inserted to position 46 then withdrawn to perform another coupling check.
Th's coupling check was satisf actory; however, during the next five attempts, the control rod failed its coupling checks three times. Additionally, the " full out" backlighting was not received during the checks.
A similar problem with coupling checks occurred during the las; operating cycle with control rod 30-23. The licensee had met witn the control rod drive (CRD) mechanism vendor to evaluate possible causes of the control rod uncoupling and the intermittent operation of the " full out" backlighting. The licensee concluded that when the control rod was at position 48, the coupling check was causing the control eod to become uncoupled. The licensee satisfactorily completed a coupling check and verified proper response in the neutron monitoring system.
The control rod was then inserted to position 46 to prevent potential uncoupling when the CRD is exercised.
During the 12R outage, the control rod drive mechanism and control rod 30-23 were replaced.
The inner filter of the CRD was found to be very dirty and the filter was " loose and sloppy." The inner filter may have caused the uncoupling rod to be lifted slightly during the coupling check and caused the control rod to be uncoupled.
The licensee reviewed the recent uncoupling event. Control rod 06-23 had previously failed its coupling check during control rod drive exercises in May, June and August 1989.
Control rod 06-23 satisf actorily passed its last coupling check and was inserted to position 46. Nuclear monitoring instrumentation response verified proper control rod blade movement.
The licensee concluded that continued operation with control rod 06-23 at position 46 is acceptable. The bases for this conclusion were: (1) a rod drop accident was not a concern because the rod is now coupled and almost fully withdrawn; (2) although the rod could be uncoupled again, eliminating coupling checks minimized the chance of future uncoupling events; (3) collet finger / housing problem was not a concern because the
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control rod successfully latched at positions 42, 44 and 46; and, (4) this problem did not appear generic because it had been limited to only one control rod during this operating cycle.
The licensee planned to continue operating with this control rod at position 46 during the current cycle. A work request to rebuild control rod 06-23 in 13R and to inspect it to determine the cause of the uncoupling was generated. Additionally, contingency plans were being prepared in the event the control rod blade requires replacement.
The inspector reviewed the licensee's actions in regard to the uncoupled control rod, The actions were assessed to be adequate to address potential safety concerns.
6.0 Loss of MCC 1A22 Event Inspection Report 50-219/89-27 discussed an event on 10/25/89 that resulted in declaration of an unusual event due to a total loss of reactor coolant boundary identified and unidentified leak rate indication. At the time of the event, workers were installing fuse blocks to certain 460V breakers. An improper placement of a fuse bicek in one of the breakers caused a fault in the motor contiol center (MCC) 1A22 which supplied power to the leak rate integrators.
The licensee's critique determined that the root cause of the event was the worker's inattention to detail while performing the work.
The fuse block screws were installed too close to the Bakelite insulation, which put the end of the ground screw too close to the phase lug. The phase lugs became energized when the breaker was reinstalled in its cubicle, and the resulting fault caused a loss of the entire MCC.
Inadequate field supervision and lack of knowledge from the worker's part on how the breaker could become energized when put in its cubicle were considered as contributing causes.
The licensee's corrective action was to revise the work package to give specific detailed instructions on the location of the fuse block. A g
caution statement was added to specify a minimum of one inch clearance l
between the fuse block mounting screws and the phase lug insulation. The revision also required use of templates with the caution statement for layout. A requirement to perform a megger test on the breaker before reinstalling the breaker in its cubicle was also added. Although the licensee's identified root cause and the corrective action do not totally match, the inspector evaluated the corrective actions and found them acceptable. The inspector also visited the work area and inspected some completed fuse block installations. This inspection and review of the revised installation procedure did not identify any unacceptable conditions.
Panel IM-175, which supplies power to the integrators for identified and unidentified leak rate, was lost during the 10/25/89 event due to loss of
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its norval power supply MCC-1A22. As an interim measure IM-175 was connected to MCC-1B22. After 1A22 was returned back to service, the licensee prepared a plan to restore normal power to IM-175.
In addition to the integrators, the IM-175 supplies power to the control logic for the drywell (DW) floor sump (1-8) pumps, the alarm for the DW floor sump level, and the control logic for the fuel pool cooling system valves. Contrary to the licensee's previous understanding, panel IM-175 does not supply power to the level alarm for the DW equipment drain tank (DWEDT).
The lic'ensee's plan indicated that during restoration of normal power supply to IM-175, it would lose power for about 20 minutes. The alternate means for monitoring identified and unidentified leak rate, however, would be available using the level alarm for the DWEDT and the leak rate recorder on panel 3F in the control room.
The licensee's plan was reviewed by the inspector and no unacceptable conditions were identified.
7.0 Hydrogen leakage from the Main Generator During this inspecticn period, the licensee discovered a hydrogen leak in the "B" phase neutral bushing of the main generator. The leak was
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detected when the generator hydrogen makeup increased significantly.
The normal leak rate from the generator is approximately 400 standard cubic feet per minute (scfm). The estimated average leak rate was approximately
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2200 scfm.
In response to the increased hydrogen leakage, the licensee took several immediate actions.
The turbine building was sampled for hydrogen to ensure there were no pockets of hydrogen collecting anywhere in the turbine building.
Samples were taken in the turbine building twice per shift. The licensee positioned several temporary fans to disperse the hydrogen from the source of the leak, Calculations were performed by-Technical Functions which showed that hydrogen levels would not exceed 0.4% if an assumed leak rate of 4000 sefm were dispersed only by natural convection.
Based on this calculation, an action level of 4000 scfm was established. The hydrogen was verified to be leaking into the turbine building and not into the ground transformer cabinet.
Strict controls for welding and other activities which may cause arcing were established for the turbine building.
Several actions were planned to address the hydrogen leak.~ A Gas Tech hydrogen monitor was obtained to provide continuous indication of hydrogen-levels. Technical Functions was refining the calculations on the correlation between hydrogen leak rate and concentration.
Temporary repair of the bushings by tightening the 6 flange bolts was reviewed; however, not considered feasible while operating.
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l-When the plant was shutdown to calibrate the condenser vacuum instrumenta-tion, the licensee tightened the flarige bolts. This action reduced the hydrogen leak rate to less than 400 scfm.
The licensee's actions to monitor hydrogen levels and minimize concentrations in the turbine building provided assurance that hydrogen (
levels would not exceed an explosive limit.
The repair of the leak during y
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the shutdown was appropriate.
The inspector had no further questions.
B.0 lionthlySurveillanceObservation
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On 11/22/89 the inspectee observed the performance of monthly surveillance test 645.4.001, " Fire Esp Operability Test" for fire diesel pump 1-2.
The operator performing the surveillance verified the preremiisites, had the proper approvel to start the procedure, and established communication with the control room.
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Due to the high noise level in the fire pump house the operator had to take the telephone receiver outside the pump house to communicate with the
control room. The licensee's critique of a previous event identified the
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need to evaluate communication facilities at the fire pump house for use
of headphones with radios (see paragraph 11.0, unresolved item 87-08-04).
The licensee's documentation indicated that headphones were pu* chased for use during noisy jobs. However, it does not appear that headphones are being used during surveillances at the fire pump house.
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The surveillance procedure requires that the pressure recording chart be q
replaced after the diesel engine has stopped.- However, the operator
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replaced the chart before the engine stopped, thus performing that step out of sequence. When asked about this, he explained that an attempt to
. replace the chart after the engine stopped could restart the engir,e which-has an auto restart on low system pressure.
He also indicated that he would talk to his supervisor about a procedure change. The inspector';
followup with the licensee indicated that a procedure change _was-not needed. The licensee indicated that a memorandum amplifying the need for -
procedure compliance would be issued.on this as required reading, and the-Involved operator would be consulted on this subject.
The inspector also noted that as soon as the diesel fire pump started, the room became flooded. Water was gushing out from a floor penetration.
through which the relief valve at the_ discharge'of the pump exhausts into
the pond. The licensee indicated that water on the floor did not create any plant safety or equipment operability problem and that_they attempted to close the opening with grouting previously but were not successful.
The inspector did not have any other questions.
9.0 Monthly Mainance Ot,servation On 12/2/89, inspectors observed the troubleshooting and repair of the i
remote valve position indication for the torus vent bypass valve, V-28-47.
An immediate maintenance short form, SF #57271, was genarated to
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investigate and repair the failure. The cause was determined to be a limit switch problem. The limit switch was repositioned and tested satisfactorily. The maintenance was conducted in accordance with the requirements of immediate maintenance as specified in Station Procedure
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105, " Control of Maintenance." Authorization from the Group shift Supervisor was obtained prior to start of work. The maintenance supervisor reviewed the planning, preparation and conduct of the
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maintenance, Written procedures were developed to troubleshoot, repair
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and test the valve position indication.
Current revisions of drawings were used. The use of the immediate maintenance short form for this
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activity was appropriate.
The Group Radiological Control Supervisor briefed the workers prior to entering tha torus room.
The radiological survey and RWP requirements were reviewed with the technicians.
Proper radiological precautions and
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adequate radiological controls support were observed by the inspector.
No ur. acceptable conditions were identified.
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10.0 Observation of Physical Security m
F During daily tours, the inspectors verified that access controls were in
accordance with the Security Plan, security posts were properly manned,
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protected area gates were locked or guarded and that isolation zones were w
free of obstructions. The inspectors examined vital. area access points to verify that they were properly locked or guarded and that access control-
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was in accordance with the security plan.
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11.0 Previously Opened Items (Closed) Unresolved Item 86-24-04. This item was updated in Inspection Report 50-219/89-27 and was left open pending resolution of the acceptability of the licensee's Motor Operated Valve Analysis and Test System (MOVATS) proceaure regarding the need for a technical:reviaw of the MOVATS sign &ture before the valve is declared operational.
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The licensee's MOVATS procedure requires Plant Engineering (PE) to review and approve parameters not meeting specified acceptance criteria before the valve is considertd operable. PE indicated, that in practice, they review all MOVATS signatures before the valve is accepted as operable even
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though this review is not required by procedure.
The inspector questioned the difference between the procedure end the practice. The licensee
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indicated that the M0 VATS procedure would be revised-to require PE review and approval prior to declaring the valve operable in cases of full MOVATS tests. Hewever, when only a current trace is performed, a PE review prior to declaring the valve operable would not be required if all acceptance criteria are met. This revision will be incorporated before the 13R outage. The inspector acted a procedural error in the calculation of
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thrust values as described in Appendix F, " Determining Thrust TMD Displace-i ment Values," in the MOVATS procedure. The licensee indicated the
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required correction would be made in the next revision.
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Generic Letter 89-10 " Safety Related Motor Operated Valve Testing and
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Surveillance", requested tht. licensee to perform a design basis review on the subject and provide the NRC with a program description and schedule for necessary revision. Due to this Generic Letter, the subject of MOVATS is open for additional review and revisions by the licensee.
Based on this and the above commitment, this unresolved item is closed.
-(Closed) Inspector Follow Item 86-39-01.
Inspection Report 50-219/86-39 documented an inspector's concern that certain components should be orotected as vital.
TUIS PARAllAPil faiffAIM 1AFEISBt
!mMAtl0!! AM is NOT l'OR PELIC 0!SCLC5EE, li IS liitsil0Elli<
LEfi BliM.
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This commitment has been incorporated in the Oyster Creek Nuclear Generating Station Integrated Schedule which is required by the approved Long Range Planning program.
Based on the licensee's commitment in the Integrated Schedule and the results of the RER, this item is closed.
(Closed) Unresolved Item 87-08-04 This item involved an event.dur.ing which the fire water system lost its supply.
This item was left open
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pending NRC review of licensee's corrective action.
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Oyster Creek primary fire protection water system consists of two 400 gpm electric pond pumps and two 2000 gpm diesel driven fire pumps (FD) taking i
sucticri f rom a 7.2 million gallon pond. One pond-pump-normally runs to '
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maintain pressure in the distribution header.
The other pump is in standby and starts if the running pump trips, unless the trip is from a fire diesel pump start. The fire diesel pumps start automatically if-system pressure decreases. Control switches are provided in.the control room and also locally at the fire pump house.
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On 4/6/87, one equipment operator (EO) and two fire protection technicians (FT) were sent to the ~ fire pond / fire pump house to conduct inservice -
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testing on fire diesel pump FD 1-1.
This was in preparation-for maintenance on fire diesel pump F0 1-2.
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A series of missed procedural steps, poor communication with the control room, poor coordination between the technicians and action.s outside the procedure resulted.i~n fire diesel pump 1-1 being valved out of the distribution header, both pond pumps. tripped and the fire diesel pump 1-2 manually stopped, thereby rendering the fire water system with no supply.
The licensee held a critique meeting and determined that the E0 and FTs
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failed to comply with several plant procedures, including the surveillance
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procedure of concern, procedure for reporting all safety system status
changes and irregularities to the control room and the procedure regarding.
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coordination and division of responsibility between the E0 and FTs.
The critique recommended certain changes to improve the procedure. Also, changes were recommended to clarify the duties and responsibilities of involved personnel and to identify the appropriate chain of command.
Addition &1 training and use of headphones with radios was also recommended. The recommended procedure changes were implemented.
Licensee indicated that the required training items were also completed, The licensee's documentation indicates that headphones were purchased for use during noisy jobs.
However, the headphones are not routinely used during work in the fire pump house. The licensee,eported this item to the NRC via telephone and also in writing following the technical
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specification requirements.
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The inspector reviewed the revised procedure, observed performance of a
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fire pump surveillance (see paragraph -8.0) and interviewed-the operator.
Based on implementation of these actions, this item is closed.
(Closed) Notice of-Violation 87-20-01. Inspection Report 50-219/87-20 identified that the range of the gauge measuring flow during Emergency Service Water (ESW) system testing was more than four times the reference value for ESW pump 52D. ASME Code Section XI, Article IWP411 and Oyster Creek Station Procedure 125.1, "In-Service Test Program Administration,"
require that the full scale range of instrument gauges shall not be greater than four times the reference value. A Notice of Violation was issued for this noncompliance.
The licensee identified in their response to this violation that when the ESW system flow was rebaselined from 4250 gpm to 3600 gpm, the existing gauge range was not reviewed for compliance with-ASME Section XI. The
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licensee stated that their corrective actions were to replace the gauges with the appropriately sized scales' and to review-all other gauges in the
IST program for compliance with ASME Section XI. The licensee stated that l
no other exceptions were noted.
The inspector verified the gauges on the ESW system met ASME Section XI requirements. Additionally, through personnel interviews and personnel
logs, the inspector verified that a review of all gauges was completed.
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This item is closed.
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(Closed) Unresolved Item 87-33-02. This issue involved work performed on
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the reactor building to torus differential pressure switch DPS 66B using an extended short form, and a valve position verification performed by the same individual assigned to perform the surveillance.
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Regarding the first issue, the inspector noted that an extended short form supplement was prepared to repair DPS 668.
Since DPS 66B is a nuclear safety related, environmentally qualified component, the inspector questioned the use of an extended short form which did not receive the same level of review or work control as a regular short form.
The licensee's documentation indicated that although an extended short
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form was intended for repair of the switch, it was not utilized to repair or replace the switch. When the switch was retested, its reset point was found to be acceptable.
Regardless, the switch was replaced within a few-days via short form 45580.
The licensee did not initiate any corrective action to address the unresolved item since the switch was'not repaired or replaced via the extended short fctm.
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'The licensee's intended use of an extended short form to repair DPS 66B is
contrary to procedure 105, " Control of Maintenance".. This procedure allows use of a standing work request (synonymous to extended short form)
for specific categories of repetitive minor maintenance, such as
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relamping, dcor repairs, adjustment of security equipment, etc., and routine non-RCA f acilities repair and services. The repair of DPS 66B did
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not qualify as a repetitive minor maintenance as described in the procedure. The ut,e of an extended short form in this case would have.been
contrary to the procedure.
In addition, an internal memorandum deted 9/17/87 from the MCF Supervisor-Work Inventory Control clarified the use i
of standing work requests to be limited to certain applications. These applicacions were to include only non-QA safety list components unless expressly authorized. The memorandum further stated that discrepancies discovered during a surveillance would require a unique short form.
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ments and guidance available to licensee personnel for use of.the extended short forms:
Short Form - A000-WMS-1220.13.
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Job Order - A000-WMS 1220.08.
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The procedures contained very little guidance in terms of what constituted e
the boundary of an extended short form.
It also appeared that the licensee depended heavily on the knowledge of the MCF planners and the MCF area supervisors in regards to the appropriate use'of extended short forms.
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To assess current use of the extended work request process, the inspector reviewed two extended work requests in the I & C and electrical areas.
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The individual job orders performed under the extended short forms were mostly in the non-nuclear safety areas and < net the guidelines of Adminis-trative Procedure 105. -Under extended work request 4108, I & C Recorder
Maintenance, a job order 17003 was initiated in May 1989 to repair the l
shutdown cooling and fuel pool temperature recorder.
This job order clso replaced the thermocouple TE52A.
Replacement of the thermocouple was outside the scope of the extended work,equest. The licensee reviewed.
this and concluded that the replacement of the thermocouple under the
extended work request was not proper.
The licensee also reviewed the completed work and concluded that no condition adverse to safety was introduced because of this. The licensee intends to issue a " required reading" memorandum on this issue. The inspector considered this event an example of a poor work control in this area. However, based on licensee action and that only one incident was identified during the entire review, this item will be closed.
The second concern identified in the unresolved item involved an independent valve verification check off being performed by the same individual who was assigned to pe,-form the surveillance.
This was contrary to the
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l procedure which requires that an individual perform the valve verification who did not witness the initial valve line up. The inspector could not t
identify any specific action taken by the licensee to address this procedural noncompliance.
However, upon a review of the. documentation for i
subsequently completed surveillances (604.3.001), no such procedural noncompliance was identified. Also, the inspectors' recent observations of surveillances indicate the licensee is performing an independent-valve
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verification checkof.f.
Based on this observation this unresolved item is i
closed.
(Closed) Unresolved Itsm (87-35-01).
Licensee's use of the. Mass Point
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Method (MPM) for containment integrated leak rate testing.
The Code of Federa'. Regulations, 10 CFR 50, Appendix J, was revised on November 15, 1988. to explicitly permit the use of the MPM when used with a test duration cf 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as an acceptable method for calculating containment leakage.
Prior to issuance of this revision, the specific
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wording of Appendix J did not include the use of the MPM as an alternative i
test.
The licensee is using.the MPM and a test duration of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as currently
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specified ir 10'CFR 50, Appendix J.
Additionally, prior to the inclusion of the MPM in Appendix J, the licensee had requested an exemption that was granted by NRC on April 18, 1988, to permit the use of this test method-for containment leak testing.
The licensee's use of MPM is acceptable, and this item is closed.
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12.0 Inspection Hours Summary Inspection consisted o.f 160 direct inspection hours out of a total of 318 inspector hours on site.
Thirty of these direct inspection hours were performed during backshift periods, and fourteen of these hours were deep backshift hours.
13.0 Exit Meeting and Unresolved Items A summary of the results of the inspection activities performe'd during this report period was made in a meeting with senior licensee management at the end of this inspection. The licensee stated that, of the subjects discussed at the exit interview, no proprietary information was included.
Unresolved items are matters for which more information is required in order to ascertain whether they are acceptable, violations or deviations.
An unresolved item is discussed in paragraph 3.0 of this report.
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ATTACHMENT I Personnel Contacted
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Licensee Personnel
- R. Barrett, Plant Operations Director M. Bradley, MCF,-I & C Maintenance
- G. Busch, Licensing Manager
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- B. De Merchant,-Licensing R. Fenti, QA Mod /0PS Mgr.
P. Fis'chler, Electrical Supervisor
- E. Fitzpatrick, Vice President-& Director R. Gayley, GE Ops Engineer J. Galanto, Mechanical Engineering J. Groemm, MCF, I & C
T. Jenkins, MCF, Management & Construction R. Keaton, Director of Quality Assurance M. Lamberto,' Mechanical Engineering L. Lammers, Plant Materiel Director K. Mulligan, Plant Operations
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R. Peck, MCF, Planning
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D. Ranft, Plant Engineering J. Rogers, Licensing P. Scallon, Plant Operations Manager
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- E. Scheyder, MCF Director B. Shumaker, Plant Operations.
- M. Banerjee
- E. Collins D. Lew Denotes attendance at exit meeting.
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