IR 05000213/1989024
| ML20006A950 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 01/16/1990 |
| From: | Haverkamp D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20006A949 | List: |
| References | |
| 50-213-89-24, NUDOCS 9001310008 | |
| Download: ML20006A950 (11) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION 1 Report No.
50-213/89-24
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License No.
DPR-61 i \\
i Licensee:
Connecticut Yankee Atomic Power Company
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P. O. Box 270 Hartford, Connecticut 06141-0270 Facility:
Haddam Neck Plant Location:
Haddam Neck, Connecticut Dates:
November 22, 1989 - January 8, 1990 Reporting Inspector:
John T. Shedlosky. Senior Resident Inspector Inspectors:
Andra A. Asars, Resident Inspector John T. Shediosky, Senior Resident Inspector Approved by:
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I//dfo Donala R. Haverkamp,_ Chief
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Date Reactor Projects Section 4A
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Division of Reactor Projects Inspection Summary:
Inspection on November 22, 1989 - January 8, 1990 (Inspection Report No. 50-213/89-24)
Areas Inspected:
Routine safety inspection by the resident inspectors. Areas reviewed included the reconstitution of fuel assemblies, work conducted on the reactor vessel thermal shield removal, the discovery of a mispositioned valve within the fuel oil supply to an emergency diesel generator and licensee follow-up to two incidents which involved alleged mishandling of fire arms by security force personnel.
Results: This was the third -routine resident inspection during the 1989 Refueling Outage. The licensee had taken direct and prompt actions following the discovery of a mispositioned valve within the fuel oil supply to the "B" emergency diesel generator; the-cause of this event remains undetermined (section 2.2.1).
Significant progress was made in the process of reconstitu-tion of nuclear fuel assemblies.
Fifty-seven assemblies have been completed at the end of the inspection period.
The licensee made inspections of additional randomly selected fuel rods in an attempt to justify,' through a statistical sample, not inspecting every rod which was exposed to debris in the reactor coolant and is scheduled for reuse.
This sampling process is under review by.
the NRC Office of Nuclear Reactor Regulation (section 4.1.2).
Most of the
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9001310008 900117 PDR ADOCK 05000213 o
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preparation work for removal of the reactor vessel thermal shield from the core support barrel was completed at the end of the inspection period.
Final preparations are being made prior to starting the cutting process, scheduled for the week of January 15 (section 4.1.1).
Prompt actions were taken with
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individuals on the station security guard force following reports of mishantiled
fire arms. The incidents appear to be isolated to two persons (section 5.1).
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k TABLE OF CONTENTS
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Summary of Facility Activities (71707)*..............
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Plant Operations (71707, 71710, and 93702)
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2.1 Operational Safety _ Verification c.
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s 2.2 Followup of Events Occurring During the' Inspection Period...
2.P. 1 Diesel Generator Fuel Cut-off Valve Hispositioned....
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Radiological Controls (71707)...................
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Maintenance and Surveillance (61726, 62703, and 71707)_
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4.1 Maintenance Observation
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4.1.1 Preparation for Reactor Vessel Thermal Shield Removal.
4.1.2 Inspection and Reconstitution of Reactor Fuel.....
4.2 Surveillance Observation..
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Security (71707)
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5.1 Mishandling of Firearms by Station Security Guards......
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Safety Assessment and Quality Verification (40500, 71707, 90712, and 92700)...-.....................
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6.1 Plant Operations Review Committee
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6.2 Review of Written Reports
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Exit Interview (92703)
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- The NRC Inspection Manual inspection procedure or temporary instruction that was used as inspection guidance is listed for each applicable report section.
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DETAILS 1.
Summary of Facility Activities The Fif teenth Refueling Outage continued during this inspection period.
Major work activities included fuel inspection and reconstitution and preparations for removal of the reactor core support barrel thermal shield. All reactor fuel remained'in the spent fuel pool.
2.
Plant Operations 2.1 Op'erational Safety Verification The inspector observed plant operation and verified that the plant -
was operated-safely and in accordance with licensee procedures-and regulatory requirements. Regular tours were conducted of the follow-ing plant areas:
control room security access point
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primary auxiliary building protected area fence
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radiological control point intake structure
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electrical switchgear rooms diesel generator rooms
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auxiliary feedwater pump room turbine building
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Control room instruments and plant computer indications were observed
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for correlation between channels and for conformance'with technical specification (TS) requirements. Operability of engineered safety i
features, other safety related systems and onsite and offsite power sources were verified. The-inspector observed-various alarm condi-tions and confirmed that operator response was-in accordance with j
plant operating procedures.-
Routine operations surveillance testing i
was also observed. Compliance with TS and implementation of appro-
'priate action statements for equipment out of service was inspected.
Plant radiation monitoring system indications and plant stack traces i
were reviewed for unexpected changes.
Logs and records were reviewed
to determine if entries were accurate and identified equipment status or deficiencies. These records included operating. logs, turnover sheets, system safety tags, and the jumper and lifted lead book.
Plant housekeeping controls were monitored, including control and
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storage of flammable material and other potential-safety-hazards.
The inspector also examined the condition of various fire protection, meteorological,:and seismic monitoring systems.
Control room and
shift manning _were compared to regulatory requirements and portions-i of shift turnovers were observed.
Control room access was properly -
controlled and a professional atmosphere maintained.
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In addition to 114 hours0.00132 days <br />0.0317 hours <br />1.884921e-4 weeks <br />4.3377e-5 months <br /> of inspection during normal utility working hours, the review of plant operations was routinely conducted during portions of deep backshifts (weekend and midnight shifts).
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coverage was provided for 10.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> during deep btekshifts. Opera-tors were alert and displayed no signs of inattention to duty or
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fatigue.
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2.2 Followup of an Event Occurring During Inspection Period i
During the inspection' period, the inspectors provided 'onsite coverage and follow-up of an unplanned event.
Plant conditions, alignment of.
I safety systems, and licensee actions were reviewed.
The inspectors I
confirmed that the required notifications were made to NRC.
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event follow-up, the inspector reviewed the corresponding plant information report (PIR) package, including the event details, root cause analysis, and corrective actions taken to prevent recurrence.
The following event was reviewed:
2.2.1 Diesel Generator Fuel Cut-off Valve Misposittened During a routine post maintenance test of emergency diesel generator (EDG) 2B on December 21, it was. identified that the
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manual emergency fuel cut-off valve to the engine driven fuel-oil pump was shut. -Although the engine ran at full load with; its electri. motor driven fuel oil pump, operators shut the
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engine down, opened the fuel' cut-off valve and successfully restarted the engine. The identical valve for EDG-2A was verified to be in-the correct position.
The EDG is supplied with fuel by two redundant pumps; one mechanically driven by the. engine and one electrically driven by i
a d.c. motor. An emergency fuel oil cut-off valve is located at j
the suction of the engine driven pump.
Its purpose is to stop a
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"run-away" diesel engine by blocking fuel flow through the only i
operating pump in the event that.the machine were-to run after
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the fuel injector racks were tripped.
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.1 The previous surveillance had been successfully' performed on November 28. This test would have identified if the cut-off-i valve had been'mispositioned because during the11 minute idle i
period (to allow cooldown af ter running at full load) only the engine driven pump supplies fuel oil.
The electrical motor driven pump is stopped by the diesel shutdown circuit.
Addi-y tionally, the discharge pressure of each pump is indicated on the local control panel and was observed to be normal.
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i The EDG was taken out of service on January 6, for replacement, of the engine driven pump and fuel oil filters as a precaution because of being operated with its fuel suction supply shut on L
December 21. On examination of the pump, no internal damage was found. A monthly surveillance test was run on January 8, and
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the EDG was retuned to standby service.
The inspectors conducted an immediate review of the incident shortly after the licensee's discovery.
It has not been deter-mined how the fuel cut-off valve came to be mispositioned.
The valve operator is a push pull plunger and is in a protected location below the EDG controls panel.
The valve is clearly identified as an emergency fuel shut off with a red warning
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1abel; its handle is also painted red.
It is normally pushed in to open, pulled out through a detent to shut off fuel.' The-location, identification and operation makes eliminates acci-dental shut-off of the valve unlikely.
Although the finding was made during a post maintenance surveil-lance test, the maintenance involved components in the air start compressor system which are not.in the vicinity of the valve.
The licensee has added operator verification of cut-off valve position to the shif tly operator duties and _ logs.
The inspectors reviewed the' licensee's response to this event,
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the corrective actions, and subsequent EDG maintenance and testing.
No discrepancies were identified.
3.
Radiological Controls During routine tours of the accessible plant areas, the inspectors observed the implementation of selected portions of the licensee's radio-logical controls program. The utilization and compliance with radiation work permits (RWPs) were reviewed to ensure detailed descriptions of radiological conditions were provided and that personnel adhered to RWP requirements. The inspectors observed controls of access to various radiologically controlled areas and use of personnel monitors and frisking methods upon exit from these areas.
Posting and control of-radiation areas, contaminated areas and hot spots, and labelling and control of containers holding radioactive materials were' verified to be in accordance with licensee procedures. During this inspection period, radiological controls for outage activities were observed. Health physics technician control and monitoring of these activities were determined to be ad(quate, j
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Maintenance and Surveillance 4.1 Maintenance Observation The inspector ubserved various maintenance and problem investigation
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activities for compliance with procedures, plant technical specifica-'
i tions, and applicable codes and standards. The inspector also verified the appropriate quality services department (QSD) involve-ment, safety tags, equipment alignment and use of jumpers, radiologi-cal and fire prevention controls, personnel qualifications, post-maintenance testing, and reportability.
Portions of activities that-were reviewed included:
reactor vessel thermal shield removal preparation, and
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fuel inspection and reconstitution.
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4.1.1 Preparation for Reactor Vessel Thermal Shield Removal Most of the preparation work for the removal of the reactor vessel thermal shield from the core support barrel was com-pleted during this inspection period. These activities were performed in accordance with plant design change
record (PDCR) No, 987, Connecticut Yankee Thermal-Shield i
Removal Program, Revision 0, with early construction approval dated December 11, 1989; and, vendor procedures VP-514, Thermal Shield Support Block Bolt' Removal and Installation at Connecticut Yankee, CYW; MP 2.7.1, CYW-16, dated October 27, 1989 and VP-519, Westinghouse: Thermal-Shield Removal, Phase I, MP 2.7.1, CYW 17, dated December 11, 1989.
These plans and procedures directed the removal of bolted fasteners and pins from the thermal shield support struc-l ture and the replacement of three lower support block bolts
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with temporary bolts.
The limiter keys and key ways ^were replaced with temporary bolted attachment devices intended
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to stabilize the thermal shield during the removal process.'
All of this " Phase I" work is reversible.
It was conducted
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prior to completion of the safety analysis which supports
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the modification.
Discussions were also being made with
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members of the NRC Office of Nuclear Reactor Regulation
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(NRC NRR) staff. A status meeting was' held in the NRC's
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Rockville, MD office on December 21.
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i The licensee intends to submit a request for operating license amendment per 10 CFR Part 50.90 to address the elimination of the storage location for the reactor vessel material surveillance specimen capsules. The licensee intends to install external vessel neutron dosemetry and place the remaining specimen capsules in a host reactor.
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The remaining changes will be addressed through licensee-I reviews made under 10 CFR Part.50.59.
i The work was included in frequent reviews made by the inspectors.
In progress observations were made of both work control and radiation protection. A significant-portion of the activities were performed by divers.
Some
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automated machining equipment,-electron discharge machin-ing, was in use.
However even this equipment was.posi-
tioned with the assistance of a diver. There were no
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unacceptable conditions identified.
J At the end'of-the inspection period, was the status of the I
core barrel assembly was as follows:
q The vessel surveillance capsule assemblies were removed and
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stored.
.i Segments of the-surveillance capsule assembly holders-U
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between the thermal shield and the first support bracket
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were removed.
-1 The thermal shield lower support blocks were prepared
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by the removal of all existing permanent bolts and all upper dowel pins.
Temporary top bolts were installed-
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in selected blocks.
Upper limiter keys'and key ways were removed including
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bolting and dowel pins.
Temporary thermal shield hangers and stabilizer base
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plates were installed in place'of the limiter keys
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assemblies.
The core support barrel was placed back into the
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reactor vessel.
Activities in progress at the end of the inspection period included: locating base plates for seismic, supports around-the reactor cavity in the containment, elevation 48' 6";
the final check out and delivery to the site of the machin-ing equipment which will' be used to make the primary' cuts of the thermal shield; and, analysis and planning to s
support the removal process.
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4.1.2 Inspection and Reconstitution of Reactor Fuel The reconstitution of fifty-seven fuel assemblies.had been completed at the end to the inspection period.- This work was conducted in accordance with vendor procedures VP-473, Babcock & Wilcox Procedure No. F0-029 CY Eddy Current Inspection of Fuel Rods, Revision 0, dated September 8.-
1989; VP-474, Babcock & Wilcox Procedure No. FO-028 CY. Fuel Rod Handling', Revision.0, dated September 8, 1989;-and,
VP-475, Babcock & Wilcox Procedure No. F0-026 CY Inspection-and Fuel Assembly Reconstitution Equipment Installation and Setup. The overall status of the fuel reconsititution, program is described in Northeast-Utilities letter to the-NRC dated December 13,1989, " Cycle 16 Fuel. Recovery Program."
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This work was frequently observed by the. inspectors. There.
were no unacceptable conditions identified.
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The status of these activities was also included in the-above.c.entioned meeting with NRC NRR on December 21.
During that meeting the licensee was requested to include.
j inspections of additional randomly selected fuel' rods from
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the fuel assemblies scheduled for reuse. This request was based on analysis results that a' reactor-coolant system-depressurization accident could not result-in more than
!j 1240 failed fuel rods and still be within 10 CFR Part 100-l requirements.
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The licensee devised a sampling program which is currently under NRC NRR review.
In that program, additional fuel
rods which were not known to be failed or at debris sites were examined by eddy current testing (ECT) to reach a 95%
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confidence level that no more than 1100 rods could fail due to unknown clad flaws from encounters with debris. -The sample size and selection method is included in the NRC NRR review.
4.2 Surveillance Observation i
The inspector witnessed a surveillance test to determine whether properly approved procedures were in use; technical-specification
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frequency and action statement requirements were satisfied; necessary-
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equipment tagging was performed; test instrumentation was'in cali-
bration and properly used; testing was performed-by qualified
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personnel; and test results satisfied acceptance criteria or were l
properly dispositioned.
Portions of the following activity were reviewed:
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SUR 5.1-178, Emergency Diesel Generator EG-2B Manual Starting and-
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Loading Test No unacceptable conditions were identified.
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Security During routine inspection tours, the inspectors observed implementation of
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portions of the security plan. Areas observed included access point
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search equipment operation, condition of. physical barriers, site access control, security force staffing, and response to system alarms' and degraded conditions.
These areas of program implementation were deter-mined to be adequate.
5.1 Mishandling of Firearms by Station Security' Guards On January 3, station security supervision received formal complaints from two security guards that, on separate. occasions, two security guards were observed to have mishandled their weapons while in the protected area.
Both incidents occurred on the backshift, the first c: curred on or about December 20, and the second on December 29.
The guards were observed to have drawn or spun their weapons after unloading them.
Upon receipt of this information, the-site access for all guards involved was suspended pending an investigation by.
their employer, Burns International, and the licensee.
When interviewed, the first accused individual quit without confirm-ing the allegation, the second stated that he had unloaded his_ weapon only to verify its cleanliness.
This guard was terminated because, unnecessary handling of a weapon is strictly' prohibited and heavily stressed in training sessions.
These incidents are believed to-be isolated.
Security personnel were
interviewed to determine if there were any additional examples of mishandling of weapons; none were identified. Additionally, all security guards received supplementary instructions on proper han-dling of firearms.
The inspectors found the licensee's response to this incident to be appropriate.
6.
Safety Assessment and Quality Verification 6.1 Plant Operations Review Committee The inspector attended several Plant Operations Review Committee (PORC) meetings.
Technical specification 6.5 requirements.for required member attendance were verified. The meeting agendas included procedural changes, proposed changes to the Technical Specifications, Plant Design Change Records, and minutes from previ-ous meetings.
PORC meetings were characterized by frank discussions and questioning of the proposed changes.
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tion was given to assure clarity and consistency among procedures.-
Items ~for which adequate review time was not available were postponed to allow committee members time for further review and comment.
Dissenting opinions were encouraged and resolved to the satisfaction-
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of the committee prior to approval.
The inspectors observed that PORC adequately monitors and evaluates plant performance and conducts-
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a thorough self-assessment of' plant activities and programs.
6.2 Review of Written Reports Periodic and special reports and licensee event reports (LERs) were re' viewed for clarity, validity, accuracy of the root cause and safety
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significance description, and adequacy of corrective action. The inspector determined whether.further information_was required.
The inspector also verified that the reporting requirements of 10 CFR-50.73, station administrative and operating procedures, and Technical Specification 6.9 had been met.- The following reports were reviewed:
LER 89-20 Significant Fuel Damage Identified During-Testing of Fuel Assemblies LER 89-21 Reactor Coolant Pump Seal Water Piping Identified as
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not Seismic Special Report concerning Inoperable Diesel Fire Pump, dated
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December 29, 1989 Haddam Neck Plant Monthly Operating Report 89-11, covering the period
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November 1, 1989 to November 30, 1989 Haddam Neck Plant New Switchgear Building Construction Report, dated-December 15, 1989
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No unacceptable conditions were identified.
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Exit Interview During this inspection, periodic meetings were held with station manage-ment to discuss inspection observations and findings.
At the close of the inspection period, an exit meeting was held to summarize the conclusions of the inspection. No written material was given to the licensee and no proprietary information related to this inspection was identified.
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