IR 05000309/1988024
| ML20235S248 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 02/15/1989 |
| From: | Lester Tripp NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20235S243 | List: |
| References | |
| 50-309-88-24, NUDOCS 8903070039 | |
| Download: ML20235S248 (11) | |
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U.S. NUCLEAR REGULATORY COMMISSION Region I Report No.:
50-309/88-24 License No.:
DPR-36 Licensee:
Maine Yankee Atomic power 83 Edison Drive Augusta, Maine 04336 Inspection At: Wiscasset, Maine Conducted:
December 20, 1988, through January 30, 1989 Inspectors:
Cornelius F. Holden, Senior Resident Inspector R hard
. Fre djnberger, Resident Inspector Approved By:
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(lowell E. TripM Chief Date Reactor Projects Section No. 3A Summary:
Inspection on December 20, 1988 to January 30,1989 (Report Number 50-309/88-24)
Areas Inspected:
Routine resident inspections of plant operations including:
followup on previous inspection findings, licensee event followup, operational safety verification, maintenance, surveillance, physical security, radiation protection and fire protection.
Results:
No violations were identified. A review of the inadvertent actuation of safety injection is contained in section 5.a.
Several concerns were iden-tified with regard to the implementation of the fire protection program (Detail
'10)
An Unresolved Item concerning a deficient fire door was closed (Detail
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4).. A plant trip on January 10, was the result of a loss of power to the main turbine electo-hydraulic control system (Detail 5.d.).
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DETAILS 1.
Persons Contacted Within this report period, interviews and discussions were conducted with various licensee personnel, including plant operators, maintenance tech-nicians and the licensee's management staff.
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2.
Summary of Facility Activities At the beginning of this report period the plant was at 96 percent power and increasing slowly following the completion of cycle 10/11 refueling outage. The plant was taken off line on December 22, in order to replace reactor coolant pump seals.
The plant was taken critical on December 31 and returned.to 100 percent power on January 2,1989.
The plant tripped from 100 percent power on January 10 due to a loss of electro-hydraulic control power. The plant returned to 100 percent power on January 13 and remained at full power for the rest of the inspection period.
3.
Review of Licensee Event Reports (IP 90712)
The inspector reviewed Licensee Event Report (LER)88-010 Plant Trip on High Heater Drain Tank Level to determine 'that deportability requirements were fulfilled, immediate corrective action was taken, and corrective action to prevent reoccurrence had been accomplished in accordance with Technical Specifications.
The inspector found the report to be complete and accurate. Appropriate corrective actions were implemented.
4.
Followup on Previous Inspection Findings NRC Inspection Report 50-309/88-21 Detail 3.d.
identified an unresolved l
item associated with an inoperable fire door. Further inspection was con-ducted during this report period.
Licensee records indicate that a con-tinuous security post was established in the area of fire door 3402 for the duration of the period when the door should have been considered inoperable.
Although the purpose of the security post was not a fire watch, it is reasonable to assume that the security officer would have responded to alert the control room if a fire had occurred. Specifics of the review are included in Detail 10 of this report.
Unresolved item 50-309/88-21-01 is closed.
5.
Operational Safety Verification (IP 71707)
On a daily basis, during routine facility tours the following were checked:
manning, access control, adherence to procedures and LC0's, instrumentation, recorder traces, protective systems, control room annun-ciators, radiation monitors, emergency power source operability, control room logs, shift supervisor logs, and operating orders.
On a weekly o
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basis, selected Engineered Safety Features (ESF) trains were verified to be operable. The condition of the plant equipment, radiological controls,
. security and safety were assessed. On a b1 weekly frequency the inspector reviewed a safety-related tagout, chemistry sample results, shift turn-evers, portions of the containment isolation valve lineup and the posting of notices to workers.
Plant housekeeping and cleanliness were also evaluated. Backshift inspections were conducted on December 31,.1988 and January 2, 7, 10, 21 and 23, 1989.
The inspector observed selected phases of the plant's operations to deter-mine compliance with the NRC's regulations. The inspector determined that the areas inspected and the licensee's actions did not constitute a health and safety hazard to the public or plant personnel.
The following are noteworthy areas the inspector reviewed:
a.
P_lant. Shutdown and SIAS Actuations r llewna the plant startup on December 15,1988, the licensee had o
been trending Reactor Coolant Pump (RCP) seal performance. The third stage of the number 2 RCP had failed shortly after startup. Seal water return flow was high but acceptable from number 3 RCP.
This was attributed to the material used for an 0-ring in number 3 seal.
This 0-ring was unique to number 3 RCP seal since it accommodated for the slightly smalter shaft diameter of number 3 RCP. The ifcensee observed a trend of decreasing third stage seal differential pressure on numbers 1 and 3 RCPs and attempted to _ reverse these trends by altering seal water parameters with no success. On December 21, the licensee decided to shut down and cool down the plant in order to replace all three RCP seals with the old design seals.
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At 3:05'a.m. and 3:10 a.m.,
on December 22, 1988, a Safety Injection Actuation Signal (SIAS) inadvertently occurred during the plant cool-down. Reactor Coolant System (RCS) pressure actuates a SIAS at 1585 psig.
During a plant cooldown, as pressure is reduced, a SIAS Block.
is enabled at 1685 psig on three of the four pressure channels. Oper-ators are then able to block the SIAS by placing the SIAS switches in the block position.
During this plant cooldown the operators failed to block the SIAS signal and the first actuation occurred.
The operators realized their mistake and reset the Safety Injection relays and returned these systems to their normal lineup. This action required securing emergency core cooling system pumps and returning valve lineups to normal for c'harging and letdown.
These actions were accomplished from the main control board.
During the time Safety Injection had been running, pressure increased in the RCS to the point where SIAS automatically unblocked. =When letdown was returned to normal as a part of the recovery from SIAS, RCS pressure once again decreased and initiated SIAS for the second time. The overall impact of the SIAS was minimal since the plant was shutdown and in the process of cool-ing down.
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1 The plant conducted an extensive review of this event utilizing a number of different techniques including an Operations - Department Review and an Event Review Board.
These reviews were thorough,
. resulting in the identification of a number' of contributing -factors and ' recom.nended extensive corrective actions.
The inspectors will continue to follow the licensee's corrective actions for this event.
b.
' Emergency Feedwater Pump Discharge Check Valve Functional Test The inspector reviewed the functional testing associated with the a
replacement of the discharge check valve for Emergency Feedwater Pump P-2bC (EFW-314). Due to a defect that apparently had existed in-the body of the valve since initial installation, a small leak was iden-tified in EFW-314 -during the past operating cycle.
A temporary repair was made and the valve was replaced during the recent refuel-ing outage.
The' inspector witnessed the In-Service Inspection (ISI) surveillance" of the-valve which was conducted near the beginning of the refueling outage. Since the valve was replaced. after it was full flow ~ tested in accordance with procedure 3.1.20, " Safeguards Valve Testing," the inspector reviewed the functional testing of the new EFW-314.
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review included Discrepancy Report / Repair Order 0261-88, procedure 3.17.8.1 "ISI/IST Pump Tests for Work Performed Under Discrepancy F
Reports" and procedure -3.17.8.2 "ISI/IST Valve Tests for Discrepancy-Reports or Repair Orders". The new. EFW-314 was functionally tested.
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at the end of the refueling outage as the steam generators were filled. The' functional test consisted of a full flow test similar to that required by 3.1.20 " Safeguards Valve Testing" and. observation of the operation of the Emergency Feedwater Pump (P-25C). The inspector
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considered the functional -test performed to be adequate to verify that the replaced component would perform its safety function.
The inspector had no further questions.
c.
Moderator Temperature Coefficient As a part of low power physics testing at the start of cycle 11, the L
licensee measured the Moderator Temperature Coefficient (MTC).
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predicted at the start of the outage, MTC was expected to be more positive for hot, zero power all rods out, xenon free and beginning of life conditions.
On December 15, 1988, the licensee measured an -
MTC of.57 X 10 -4 delta rho per degree Fahrenheit (F).
The Maine Yankee Technical Specifications (T.S.) 3.10 figure 3.10-10 dictate acceptable values for MTC. At the zero power condition, MTC can not be more positive than.5 delta rho / degree F.
In order to assure the plant was within this limit, a procedure change request was initiated l
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that1 required group 5 control element. assemblies to be inserted to
100 steps for zero power with a ramp up to all rods ~out at 16 percent power.. : After 16 ' percent. power, MTC is below.5 delta rho / degree-F.
After approximately' 700 megawatt days-per metric ton -(mwd /Mt)- power-operation, MTC was calculated to be 'sufficiently negative that the restrictions on rod position were no longer necessary.
These restrictions were removed at approximately 840 mwd /Mt on January 27, 1989.
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'The - inspectors observed operator adherence to these administrative
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controls throughout this report period.and - discussed these limits with the operators. All operators were knowledgeable of the limits',
when they were required and adhered. to the limits imposed.
The inspector had no further questions.
d.
Reactor Trip At 8:19.p.m. on January 10, 1989,- the plant tripped from 100 percent power. The "first out"' annunciator indicated that control power had -
been lost to the Electro Hydraulic Control- (EHC) unit to the main turbine.
All safety systems responded as required.
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Breaker-(RTB) #4 was not indicated as. tripped on the computer. data system,;however its indicated position was verified open by the con-trol room operators. This indication problem was later. traced to a '
faulty p4 connector-at the computer. RTB #4 opening time was tested'
satisfactorily prior to reactor startup.
The Instrument and Control (I&C) section conducted troubleshooting of the EHC system.
The EHC system has both normal and-backup control'
power supplies.
No deficiencies - were. identified. The licensee con-tracted a vendor representative for onsite troubleshooting assis-tance. A simulator was utilized to functionally test the EHC system.
The licensee was unable to identify a root. cause for the loss of EHC control power. Even though no problems were identified with the EHC system, the licensee elected to replace. two trip relay cards and a speed channel comparitor card.
Additionally, a trend recorder was installed to monitor control power.
A special Plant Operations Review Committee (PORC) meeting was held on January 11 in order to determine if any conditions existed which would prevent a reactor startup while troubleshooting continued on' the EHC system.
PORC g
concluded no issues would prevent a startup.and the reactor was taken critical on January 11. At the conclusion of the troubleshooting the licensee again held a special PORC meeting and concluded the plant was ready to restart.
The plant was phased on to the power grid on January 12 and returned to full power on January 13, 1988.
The inspectors reviewed the event followup, witnessed EHC trouwe-
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shooting and testing and witnessed the reactor and plant startup.
The inspectors had no adverse findings.
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c.
Containment Control Air On January 25, 1989, the licensee made preparations to perform main-tenance on Containment Control Air Compressor (C-5B) unloader valve.
The containment control air system consists of two air compressors, air dryers, air receivers and associated piping and prov'ies air to a variety of air operated valves in containment. One of the two air compressors (C-5A and C-5B) is normally operated in run and the un-loader valve cyc1'es to maintain pressure between 90 and 100 psig.
The other air compressor is in standby and will start at 85 psig.
Operations and Instrument and Control (I&C) personnel entered con-tainment at approximately 10 a.m.
to work on C-5B unloader valve.
C-5A was in run and during the time the compressor was unloaded the motor for the compressor exhibited loud noises indicating imminent failure.
I&C was unable to adjust the unloader for C-5B which required replacement of this unloader.
Operations made preparations to crosstie instrument air with the con-tairment control air by means of a normally closed manual containment isolation valve, IA-135.
An operator was stationed at IA-135 with communication capability with the control room. Operation of IA-135 would have required entering Remedial Action Statement of Technical Specification (T.S.) 3.11.B which states in part that manual valves must be closed in order to be considered as a containment boundary, however, operation of the plant is allowed for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> until the valve is returned to the closed position or the penetration is iso-lated.
If containment control air had been lost, the plant would have lost letdown flow and seal water return flow from the reactor coolant pumps necessitating a manual plant trip.
The precautions taken would have insured that the systems that rely on containment control air would have remained operable and an orderly plant shut-down would have been conducted if the plant could not meet the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Remedial Action of T.S. 3.11.B.
The C-5B unloader valve was replaced at 1:20 p.m. and tested satis-factorily.
C-5B was placed in service and C-5A was removed from service and was worked around the clock and returned to service on January 26, 1989. At no time were the plant instrument air system
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and the containment control air systems crossconnected. The inspec-tor was present in the control room while repairs to C-5B were per-formed.
The operators had taken prudent precautions to assure that
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in the event of failure of the containment control air system the
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plant was adequately prepared to perform an orderly shutdown.
The inspector had no further questions.
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l 6.
Plant Maintenance (IP 62703)
The inspector observed and reviewed maintenance and problem investigation
activities to verify compliance with regulations, administrative and main-tenance procedures, codes and standards, proper QA/QC involvement, safety l
tag use, equipment alignment, jumper use, personnel qualifications, radio-logical controls for worker protection, retest requirements, and reporta-bility per Technical Specifications.
Portions of the following maintenance evolutions were reviewed:
Discrepancy Date Report Number De_scription 1/18 0089-89 PM-M-26-M-E D/G-1B Lube Oil Pump Inspection 1/18 0095-89 PM-M-26-2^-B D/G-1B Replace Lube Oil Pumps, oil and air box inspections 1/18 6838-88 PM-M-26-5A-I D/G-1B Five year PMs, retorque crab bolts, cooling system pressure cap and crankcase pressure detector replacement 1/18 3112-88 FN-20B Damper linkage adjustment No discrepancies were identified.
7.
Surveillance Testing (IP 61726)
The inspector observed parts of tests to assess performance in accordance with approved procedures and LCO's, test results, removal and restoration of equipment, and deficiency review and resolution.
Portions of the following surveillance were reviewed:
Date Procedure Number Title 12/28 3-6.2.2.14 Safety Injection Actuation Signal Initiation Channels 1/6 3.1.20 Safeguards Valve Testing 1/18 3.1.4 Emergency Diesel Generator - Monthly Sur-veillance Testing No discrepancies were identified.
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Observations of physical Security (IP 71707)
' Checks were made to determine whether ' security conditions met regulatory.
requirements, the physical security plan, and approved procedures. Those checks included security staffing, protected. and Lvital area - barriers,
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vehicle searches and personnel identification, access control, badging,.
and compensatory measures when required. The licensee is in the process
of upgrading the security program. The inspectors have witnessed portions-of those upgrades and they appear effective.. These changes include up-grades to the processing of individuals and packages for access to the l
plant, lighting improvements and changes to onsite storage. The inspector will. continue to follow the licensee's actions in this area.
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Radiological-Controls (IP 71707)
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Radiological controls were observed on a routine basis during the report-l ing period. Areas reviewed included Organization and Management, external l
radiation exposure control and contamination control.
Standard industry
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radiological work practices, conformance to radiological control proced-l ures and 10 CFR Part 20 requirements were. observed.
Independent surveys
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.of ~ radiological boundaries and random surveys of nonradiological points throughout the facility were taken by the inspector..No discrepancies j
were notod.
i 10.
Fire Protection i
NRC Inspection Report 50-309/88-21 detail 3.d.
identified an unresolved item associated with an inoperable fire door (3402). The following para-graph from that report is included as background information:
On December 19, - during a plant tour the inspector identified-that
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fire door 3402 had apparently been damaged, resulting 'in the inoper-
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ability of the latching mechanism and excessive clearance around the
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lower portion of the door. No fire watch had been posted.
Fire door
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3402 is classified in procedure 19.29 " Fire Door Access and Repair Guidelines," Revision. 6, Attachment A, as being required to be oper-l able to meet Technical Specification 3.23 E and Appendix R require-
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ments. The door is a three hour rated double door connecting eleva-I
. tion 36' of the Primary Auxiliary Building to the Fuel Handling Building. The upper portion of the door. is penetrated by a monorail j
which has a removable device installed to seal the area where the
I monorail passes through the door.
This device was found to be.
approximately six inches away from its intended position. The licen -
see was informed of the condition of the door and a continuous fire watch was posted in accordance with the requirements of Technical
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Specification 3.23 E until repairs could be made to the door.
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The licensee discussed the condition of the door in the Morning Managers Meeting on December 21, and initiated a mystery (MYS-88-12-12) to address how the door became damaged. A mystery is a management tracking tool used to identify issues about which more information is required and to ensure resolution.
An investigation was conducted to identify how the door became damaged and the monorail seal mispositioned.
The investigation concluded that the door was apparently damaged during the morning of December 19 as a result of the movement of valve rebuilding equipment, a large steel table and other hand tools out of the upper level of the Pri-i mary Auxiliary Building.
The monorail seal was apparently mispositioned
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on December 1.
A craftsman received an electric shock when he was remov-ing the seal.
The seal apparently contacted the power supply mounted on the monorail during removal and therefore was not repositioned properly to prevent accidental contact with the power supply again.
Contractor per-sonnel were identified as responsible for both the damage to the door and
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the mispositioning of the monorail seal. Although there was an accident report initiated to identify the electric shock incident, no verbal or written report was made to identify the damage to the fire door.
Corrective actions included the following:
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a memo from the Fire Protection Coordinator to all plant personnel was issued January 3, which emphasized the importance of fire doors,
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the Security Department added a check of fire doors to the officers duties associated with their rounds which are completed every two hours, security alarms associated with fire door 3402 and fire door 3401
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(between lower level PAB and fuel building) were temporarily recon-nected to allow additional monitoring,
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the fire seal around the monorail and other similar seals throughout the plant were labelled to identify them as fire barriers and to warn of an electrical shock hazard.
Assuming that the monorail seal was mispositioned on December 1, when the electric shock accident was reported, the fire barrier would be considered inoperable from that date until December 19, 1988, when identified by the inspector which resulted in the licensee posting a continuous fire watch.
The licensee's investigation identified that a continuously manned secur-ity post was established in the area of fire door 3402 during the period that the fire barrier was considered inoperable.
Although the security post was not established as a fire watch, the officers had communication capability to alert the fire brigade if there had been a fire in the area.
l It is reasonable to assume that they would have done so, therefore the inspector considered that the intent of the remedial action for a contin-uous fire watch to be established when the fire barrier was degraded was met.
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j To better understand how the requirements of 10 CFR 50, Appendix R, ' para-graph' N.
"Fi re Doors" are implemented at the facility, the inspector reviewed the following procedures:
Operations Department Procedure 1-200-10.1 " Primary and Secondary Auxiliary Operations Logs and Forms". Revision 1, Fire Protection Procedure 19.12 " Fire Doors" Revision 5, and
Fire Protection Procedure 19.29' " Fire Door Access and Repair Guide-lines" Revision 6.
The auxiliary operators log procedure contains a form which requires the operator to tour various areas of the plant checking, among other items, fire hazards and the condition of fire protection equipment.
The form includes asterisks next to the areas which have fire doors that are required to be check closed daily in accordance with 10 CFR 50 Appendix
"R".
There were no asterisks in areas which would have alerted the oper-ators to check fire doors 3402 or 3401. Fire door 3401 is located on the 21' elevation of the Primary Auxiliary Building to Fuel Building Wall (Directly below fire door 3402).
'The inspector also noted that the operations procedure did not provide clear guidance on which doors were to be checked. An asterisk next to the areas -listed indicates that~ a door in that area requires a check. The procedure does not specify which door in the area 'is the one required to be checked.
However, operations personnel are fire brigade trained and should ~ be aware of the condition of doors which are labelled. as fire doors.
The Fire Protection Procedures reviewed also contained several discrepan-cies. The fire door classification procedure had fire doors 3401 and 3402 classified _ as " alarmed" doors, yet the alarm to those doors had been dis-abled several months earlier in order to address concerns of door opera-bility. The " Fire Doors" procedure, which provides monthly inspections of the fire doors throughout the plant, failed to list the fire door located in the fire pump house separating the' electric driven fire pump from the-motor driven fire pump as requiring an inspection. The monthly fire door inspections are used to meet the semi-annual Appendix R fire door inspec-tion requirement.
These procedural discrepancies were identified to licensee management.
Licensee review identified two more fire doors which were not listed as requiring monthly inspections although they were required to meet Appendix R.
The procedures were revised to correct the discrepancies identified.
Further review indicated that all three doors were inadvertently removed from the procedure during a revision in 1984. Although these doors were not inspected on a semi-annual basis as required by 10 CFR 50, Appendix R, paragraph N, all were checked at least shiftly for other reasons. One is also an alarmed security door, and another is a locked high radiation area door.
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Licensee management decided that increased attention to Fire Protection requirements was warranted based on the discrepancies identified. There-fore, the facilities section was reorganized to allow all of the duties that the Fire Protecdon Coordinator performed 'which were not related to fire protection to be removed from his responsibility. Also, an audit of the Fire Protection Surveillance procedures was initiated.
Although the discrepancies identified in the Fire Protection Procedures could be classified as violations of 10 CFR 50 Appendix R requirements, no Notice of Violation will be issued in accordance with the revised enforcement policy, due to the minimal safety significance of the issues and the extensive corrective actions taken by the licensee.
The inspector did identify the following items of concern:
a.
Procedure reviev/s since 1984 did not identify the discrepancies in the Fire Protection Procedures,
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b.
Contractor personnel did not have adequate oversight or awareness o'f fire protection requirements resulting in the mispositioning of the
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monorail seal and the unreported damage to fire door 3402, c.
Apparent informality of the assessment of fire protection deficien-cies for deportability in accordance with 10 CFR 50.72 and 50.73 The Fire Protection Program will receive added attention in future routine inspections.
11.
Exit Interview (IP 30703)
Meetings were periodically held.with senior facility management to discuss the inspection scope and findings. A summary of findings for the report -
period was also discussed at the conclusion of the inspection. The licen-see did not identify 2.790 material.
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