IR 05000309/1993022
| ML20059C157 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 10/25/1993 |
| From: | Lazarus W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20059C151 | List: |
| References | |
| 50-309-93-22, NUDOCS 9311010035 | |
| Download: ML20059C157 (11) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report Number:
50--309/93-22 Docket Number:
50-309
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Licensee Number:
Licensee:
Maine Yankee Atomic Power Company 83 Edison Drive Wiscasset, Maine
- 5 Facility:
Maine Yankee Atomic Power Station Date:
August 30 through October 11,1993
Inspectors:
Jimi T. Yerokun, Senior Resident Inspector William T. Olsen, Resident Inspector Dan T. Moy, Reactor Engineer
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Approved by:
An WXLazarus, Chief, Reactor' Project Section 3B Date Scope:
Resident inspection of plant activities including operations, maintenance and surveillance, engineering and overall plant support.
05erview:
See executive summary.
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9311010035 931025 DR ADOCK 050003 9
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EXECUTIVE SUMMARY Operations Maine Yankee completed the Cycle 13/14 Refueling Outage during the inspection period.
Outage activities were well conducted. Core reload was completed safely and without incident. The reactor was made critical on October 9,1993.
Maintenance and Surveillance Maine Yankee personnel performed maintenance and surveillance activities in accordance with station directives and procedures. Except for an isolated instance which resulted in damage to a safety related component, maintenance activities were well controlled and performed.
Engineering and Technical Suppon Technical evaluations of identified problems were good. Engineering personnel were technically competent. However, total leakage of PCC Isolation Valves and HPSI pump operation were left unresolved pending review of licensee evaluations of these issues.
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Plant Support Radiological contrals were well implemented especially with all the ongoing outage activities.
A non-cited violation involving failure to follow procedure with regards to control of check source was identified.
As a result of inadequate planning and scheduling, plant personnel safety was jeopardized during a Fire Protection carbon dioxide testing in the turbine building.
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Safety Assessment /Ouality Verification Outage risk management was effectively implemented during the outage. The Plant Operations Review Committee meetings were conducted in a professional manner with good safety perspective. Detailed technical reviews of modifications and procedures were performed.
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TABLE OF CONTENTS
EXECUTIVE SUMMARY ii
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TABLE OF CONTENTS
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1.
OPERATIONS
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1.1 Fuel Handling Activities...............
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1.2 Control Room Startup Activities
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1.3 Cycle 14 Startup Testing..................
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MAINTENANCE and SE RVEILLANCE...............
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2.1 Overthrusting of Valve HSI-M-42......................... 2 2.2 Demineralizer Water Storage Tank Leak....................
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3.
ENGINEERING and TECHNICAL SUPPORT
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3.1 RMS-9 Circuit Breaker
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3.2 PCC-M-150 Design Basis Test
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3.3 HPSI Pumps Runout Flowrate
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3.4 Containment Boundary Repositioning (" Rebounding")...........
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3.5 Eddy Current Testing..
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4.
PLANT SUPPORT
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4.1 Radiological Controls..
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4.2 Security.......
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4.3 Fire Protection
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SAFETY ASSESSMENT / QUALITY VERIFICATION................. 7 6.0 ADMINISTRATION....
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6.1 Persons Contacted
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6.2 Other inspections
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j 6.3 NRC Senior Management Visit
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6.5 Exit Meeting....
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DETAILS i
1.
OPERATIONS The plant was in a refueling outage at the beginning of this inspection period. Outage activities were well conducted. Core reload was completed safely and without incident. The reactor was made critical on October 9,1993. On a daily basis, inspectors verified adequate staffing, appropriate access control, adherence to procedures and Limiting Conditions for-Operation, operability of required protective systems, status of control room annunciators,
status of radiation monitors, and emergency power source operability. The inspectors also
verified operability of selected Engineered Safety Features (ESF) trains and ascertained that overall condition of plant equipment, radiological controls, security and safety was good.
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Plant material condition and housekeeping were well maintained.
Prior to startup, NRC inspectors noted leakage from the Loop 1 Hot Leg isolation Valve, RC-M-11. The licensee stated that this leakage was acceptable and that the valve was designed with a permanent piping in place which routes leakage from the valve packing into the pressurizer quench tanks. This would allow for continuous monitoring and trending of valve packing leakage during plant operation. The leakage was approximately a quart a day from the valve packing gland into the pressurizer quench tank and was well within the Technical Specification (3.14) limit of 10 gpm for identified RCS leakage. The licensee will continue to monitor this leakage over the cycle and plans to replace the packing next refueling outage.
1.1 Fuel llandling Activities On September 13, the inspector observed Maine Yankee operations personnel conducting refueling activities in the containment building. The operators performed the fuel moves as
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directed by Station Procedure No.10-1, Core Reloading, Revision 17. The fuel ' moves were safe and deliberate and no procedure deviations were noted by the inspector. A minor r
discrepancy was identified by the inspector in that Station Procedure No.13-3, Transfer Machine and Upender Operation, Revision 11, directed operators to perform tasks and required signatures after each step but did not provide for repetition after a fuel assembly move was completed. The procedure was revised to correct this discrepancy.
1.2 Control Room Stnrtup Activities
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On several occasions during the plant startup the inspectors observed control room personnel during performance of normal startup procedures and required surveillance tests. Maine l
Yankee operations department developed and implemented a new set of maneuvering procedures before the start of the refueling outage to control startups and shutdowns. This was the first occasion to use these procedures during a startup. Operators found the
procedures helpful in that they provided a more logical and flexible way to operate the plant.
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-1.3 Cycle 14 Startup Testing The inspector reviewed the low power physics test Procedure No.11-2, Revision 16, and determined that the procedure contained good prerequisites, instrument calibration -
requirements, and step by step instructions to perform the test. Test data of October 9,1993, were reviewed by the inspector and test parameters were verified to be within the acceptance criteria. The inspector noted the' acceptance criteria for the isothermal
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temperature coefficient (ITC) at hot zero power (HZP) of 0.50 x 10d ap/*F was higher than the 0.20 x 104 ap/ F recommended by ANSI /ANI-19.6.1-1985, " Reload Startup Physics Tests for Pressurized Water Reactors." The inspector verified that the licensee's core performance analysis report, Section 4.6 and 5.1.3 for cycle 14, were conservative with
the use of 0.50 x 10" ap/ F as bounded in the Maine Yankee FSAR safety analysis. All acceptance criteria were in accordance with Maine Yankee cycle 14 core performance analysis report, dated April 1993. All data were found technically adequate.
2.
MAINTENANCE and SURVEll1ANCE The inspectors observed and reviewco inaintenance and problem investigation activities to ascertain that they were safe and had proper QA/QC involvement. The inspectors also verified that these activities were in compliance with regulations, administrative and
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maintenance procedures, work order (WO) requirements, codes and standards, and that they involved proper use of safety tag, equipment alignment, jumper, and radiological controls for worker protection. The inspectors verified that retest requirements were properly addressed when required. The inspectors observed portions of activities including the following:
WO 93-2843, RPSCIP selector switch replacement
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WO 93-0944, Investigate and repair CH-A-32
WO 93-3129, Investigate and Repair valve / actuator for HSI-M-42
WO 93-02705, Clean / inspect D/G IB air start line and components Procedure No. 3-1.15.2, ECCS Operational Test Recirculation Actuation System
Maine Yankee personnel performed the observed maintenance and surveillance activities in accordance with station directives and procedures.
2.1 Overthrusting of Valve HSI-M-42 l
On September 21, Maine Yankee r'aintenance personnel tested high pressure safety injection (HPSI) system valve, HSI-M-42, per Station Procedure No. 5-18-3, Limitorque Valve Operator Operationr! Test. This test was performed following actuator overhaul which was determined to be required during static?nd dynamic testing of the valve. At the end of the valve stroke to the close position, the maintenance department contractor electricians observed the motor operated valve (MOV) actuator torque switch actuate and actuator motor
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current decrease to zero. At the same instant, they heard a loud noise from the valve actuator. The valve actuator appeared to be separated on one side from the valve. Also the
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valve stem appeared to be bent. Investigation by mairtenance department revealed that the contractor electricians had failed to properly perform a step of the actuator overhaul-procedure, Maine Yankee Procedure Noi5-18-5, Limitorque Operator Overhaul (SMB-0
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through SMB-4) thereby allowing the MOV torque switch to be set improperly. These MOV.
testing and maintenance activities were witnessed by a regional inspector who was on site to
conduct an inspection of the MOV testing activities relative to NRC Generic Letter 89-10.
The results of his inspection and details of the HPSI valve damage and subsequent regulatory actions are documented in NRC Inspection Report 50-309/93-13. The valve was repaired and successfully retested prior to plant startup.
J 2.2 Demineralizer Water Storage Tank Leak A leak in the demineralizer water, storage tank (DWST) which occurred prior to this
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inspection period was identified and repaired. Through the surveys of the tank internals performed, other leaks and potential leaks were identified and repaired. Closcout plan i
(COP)92-010 was developed to address the safety significance, plans and comments related
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to the crack in the tank. While it took a few " fill and drain" evolutions to identify and repair the cracks, the licensee's efforts were observed to be good. At no time was the leak-significant enough to cause operability problems for the emergency feedwater (EFW)
systems. During the periods when the tank was drained for repairs, EFW systems required i
to be operable were aligned to the backup primary water system.
3.
ENGINEERING and TECIINICAL SUPPORT
3.1 RMS-9 Circuit Breaker Overcurrent Trip Device Failure (Update)
Maine Yankee previously experienced inadvertent trips of RMS-9 overcurrent trip devices
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installed on circuit breakers supplying safety-related equipment. This was documented in j
NRC Inspection Reports 50-309/93-18 and 50-309/93-21.
Maine Yankee replaced the defective RMS-9 overcurrent trip devices with electro-mechanical (EC) overcurrent trip devices prior to plant startup until the vendor provides resolution to the inadvertent tripping problem. Maine Yankee installed EC electro-mechanical overcurrent trip devices in the feeder circuit breakers for safety related motor control centers (MCC) 7A, 7B, 8A and 8B. In addition, EC trip devices were installed in motor circuit breakers for service
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water pumps P-29f., B, C, and D which are also safety related equipment and part of the station ultimate heat sink. The cicctric motor driven fire pump (P-4) circuit breaker and the station blacl:out diesel generator (DG-2) output circuit breaker were also modified with EC devices to prevent inadvertent trips during operation. With assistance from vendor personnel, Maine Yankee conducted additional tests and gathered more data in order to identify the cause of the inadvertent RMS-9 trips, but have act been successful in identifying the root cause of the problem.
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3.2 PCC-M-150 Design Ilasis Test During the performance of PCC-M-150 design basis test per Procedure No. 4-1-55, the inspector observed that when alve PCC-M-150 (letdown _ heat exchanger isolation valve) was
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fully closed, there was some leakage (about 80 gpm) through the valve. While the test demonstrated the capability of the valve to close under full How conditions, the inspector
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identified a concern regarding the valve leakage in light of the safety function of trie valve.
Upon a recirculation actuation signal (RAS), this valve closes to isolate primary component cooling (PCC) flow to non-safety loads to maximize cooling capacity for the residual heat removal (RHR) system.
The licensee then initiated a technical evaluation (TE No. 343-93) to address this concern and determined that an overall leakage of 250 gpm through automatic isolation valves which operate to form the PCC system boundary during post-LOCA recirculation phase was
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acceptable. This was based on the fact that the calculated required minizaum flows to safety
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related components (RHR Heat Exchanger,400 gpm; Diesel Generator,500 gpm; and others,100 gpm) will still be available. The measured leakage through PCC-M-150 and all the other PCC isolation valves was 100 gpm.
The inspector found this technical evaluation acceptable but remained concerned over the lack of any program or controls to check and ensure that the leakage through PCC-M-150,
and similarly functioning valves, will remain within the limit that ensures adequate PCC flow
to safety related loads following a RAS. The licensee stated that the need for programmatic monitoring of PCC valve leakage was under review. This item is unresolved pending
completion of the licensee's review. (Unresolved Item No. 50-309/93-22-01)
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3.3 HPSI Pumps Runout Flowrate During a dynamic test of motor operated valve (MOV) HSI-M-41, wl.at appeared to be_.
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runout conditions were observed at high pressure safety injection (HPSI) Pump P-14A. The pump was noisy and with a Dow measured at 796 gpm (not including the recirculation Dow)
the pressure was less than 390 psig. Upon further testing, the pump noise was still present when total flow was 800 gpm. While pump vibration and temperature remained normal, the
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licensee was unable to determine that initial pump cavitation was not occurring.
The inspector expressed concerns over the licensee's inability to determine the upper design
- flow limit to prevent potential degradation to the pump. A review of previous tests, including the preoperational tests, still did not provide any clear indication of what would be an acceptable maximum Dow for the HPSI pumps. This was not an immediate safety concern, because HPSI would not be required operate under these surveillance conditions.
However, the inspector was concerned that the licensee has not determined the maximum
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flowrate of the HPSI pumps to preclude operation above this flow during surveillance testing.
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f The licensee stated that the operation of the HPSI pump was under review. This item is
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unresolved pending the completion of the licensee's review.
(Unresolved Item No. 50-
-309/93-22-02)
3.4 Contaimnent Boundary Repositioning (" Rebounding")
Containment Purge Exhaust Header Inboard Isolation Bypass Valve Flow Instrument During a plant tour on September 29,1993, NRC inspectors noted that the differential pressure cell across the containment purge exhaust header inboard isolation bypass valve,
VP-A-4, along with its high and low pressure side tubing appeared not to be seismically qualiRed. Although the manual isolation valves for differential pressure cell were safety-related, normally locked-closed, containment isolation valves, the effect of the non-seismic'
(non-safety related) component and its tubing on the integrity of the containment was.
discussed with the licensee. During containment ventilation per Procedure No.1-12-1 A, On-l Line Containment Ventilation, the non-seismic instrument was aligned to the containment by opening the three manual isolation valves to it. Procedure No. 1-12-15-1, Maintaining Containment integrity, was also in effect whenever these manual isolation valves were opened.
The inspector revimved the licensee's technical evaluation and administrative procedures that ensure that repositioning of manual isolation valves did not present an un-analyzed condition j
(outside design basis). Exception 1 of Technical Specification 3.11.B stated that manual containment isolation valves may be repositioned under administrative controls provided prior compensatory measures were taken to isolate the penetration. Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the licensee i
shall verify that the compensatory measures meet the same design criteria as the original containment isolation valve. Item 2 of the NRC's safety evaluation report transmitted by letter dated December 26,1984, to Maine Yankee Atomic Power Company found this i
exception to Technical Sneci6 cation 3,1.B acecotable since containment integrity was
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effectively maintained.
The inspector concluded that the normal configuratic1 of the containment purge exhaust
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astrument was acceptable since header inboard isolation bypass valve differential pr au normally locked close manual isolation valves sep :te tk Jon-seismically quali6ed l
instrument from the containment. However, durit m-!i.e containment ventilation, the licensee voluntarily entered a Technical Specificatit.. action statement without evaluating the l
net safety benefit. The licensee stated that the safety benent of this practice was under J
review. This item is unresolved pending completion of licensee's review. (Unresolved Item No. 50-309/92-22-03)
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During the refueling outage, Eddy Current Testing (ECT) of the steam generators was performed. Preliminary review by the NRC resident inspectors indicates that the results were good. A total of 3,436 (60%) tubes were sampled in steam generator (S/G) I and sixteen of these were plugged. In S/G 2, four of the 3,642 (64%) tubes sampled were plugged. Thirty three of the 5,645 (100%) tubes sampled in S/G 3 were plugged.
4.
PLANT SUPPORT i
I 4.1 Rndiological Controls
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Inspectors reviewed radiological controls including organization and management, external radiation exposure and contamination control. The inspectors also monitored standard industry radiological work practices, and conformance to radiological control procedures and 10 CFR 20 requirements.
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On September 5,1993, the licensee ideMified that a 10,200 dpm thorium 230 (Th-230) check
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sou ce was missing from the meter issue room storage cabinet. Two days later, the source was located in the meter issue room but not in the storage cabinet. Licensee's investigation -
revealed that a contractor employee had used the source without properly using the
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established log book for signing the source out of and back into the storage cabinet.
Radiological Incident Report (RlR) No. 93-7 was issued by the licensee documenting this
occurrence. The licensee corrective actions included emphasizing to personnel the need to I
adhere to the source users log requirements.
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This incident was a violation of Procedure No. 9-10-100, Radioactive Material Control
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Program. Specifically, step 7.7.5 of the procedure requires that source users log the source and intended use on a source sign out/in sheet. This violation of plant procedure is not being cited because the criteria of 10 CFR 2, App. C, VII.B for exercise of discretion were met.
The licensee identified the violation; the safety significance was nominal; and, immediate corrective actions were taken.
4.2 Security
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j requirements, the physical security plan, and approved procedures. The checks included security staffing, protected and vital arut barriers, vehicle searches and personnel j
identification, access control, badging, and compensatory measures when required. No
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discrepancies were identified.
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4.3 Fire Protection Carbon Dioxide. System Testine Incident During performance of Procedure No. 3-19-1, Carbon Dioxide System Testing, on the main turbine generator cardox zones on September 28,1993, plant personnel working in the main condenser had to be rapidly evacuated. This was caused by cardox vapor entering the condenser through openings in the turbine exhaust casing. This casing was being used as a
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fresh air supply for personnel working in the condenser. A safety watch, posted for confined space entry, observed the vapor entering the space, and directed personnel to evacuate.
Some individuals inhaled some of the cardox vapor in the process. Personnel were examined onsite and released back to work. After the cardox vapor had dissipated, tests of the space
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showed that oxygen levels were acceptable and the workers were able to resume the work.
Prior to performing the cardox test, the turbine deck was cleared of workers, a watch was posted on each end of the deck and a FEMCO (public address) announcement was made.
However, test personnel were unaware of ongoing work under the test zone. The workers in-the condenser were unable to hear the announcement. No other similar tests were left to be performed during the outage.
Initial investigation by the licensee revealed that the work planning and scheduling for the cardox test was less than adequate. As a result of changes in the schedule for the test, there was a lack of proper coordination to assure personnel safety in all areas of the turbine building. Planned corrective actions include enhancing the procedure and getting the industrial safety group more involved in future tests. The inspectors concluded the planned
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actions were adequate.
5.
SAFETY ASSESSMENT / QUALITY VERIFICATION Plant personnel maintained proper safety perspective in addressing emerging issues. Unusual l
Occurrence Reports (UOR) were properly initiated to address problems. Outage risk management was effectively implemented. There was continuous outage risk appraisal
during the daily morning meetings. Good use of the Critical Safety Functions were made.
A good questioning attitude was exhibited by management during planning of emergent
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work.
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On several occasions during the report period, the inspectors attended Plant Operations Review Committee (PORC) meetings to ascertain that the committee performed the required reviews and oversight to ensure nuclear safety as required by station technical specifications,
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section 5.5. The committee had the proper quorum and met within the frequency specified in the station Final Safety Analysis Report (FSAR). The meetings were conducted
professionally and good safety perspective was exercised. On occasion, procedure revision t
approvals were deferred until questions raised by the committee were resolved. Also several closecut plans were presented to the committee for review after completion of the corrective
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actions and further questioning by the committee members resulted in further evaluation of j
additional identified problems before closure of the item. During reviews of plant modifications and procedure revisions, the PORC demonstrated a questioning attitude by initiating in-depth discussions of the safety issues involved with many items on the agenda.
The inspector determined that the PORC was effective.
6.0 ADMINISTRATION i
During the inspection period the inspectors conducted backshift inspection on 8/31, 9/1, 9/2, 9/8,9/28, and 9/29, and deep backshift inspection on 9/12,9/21,9/29,10/5,10/6, and 10/10.
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6.1 Persons Contacted
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During this inspection period, inspectors conducted interviews and discussions with various licensee personnel, including plant operators, maintenance technicians and the licanste management.
6.2 Other inspections Other inspections conducted during this inspection period include: MOV Dynamic Testing
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Inspection (50-309/93-13); NRC Mobil Laboratory Non-destructive Examination (50-309/93-
15); and Engineering and Technical Support inspection (50-309/93-20).
6.3 NRC Senior Management Visit
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t On September 29,1993, Thomas T. Martin, NRC Region I Administrator visited the site.
He met with licensee management, and toured the plant.
On October b,1993, Dr. Thomas E. Murley, Director of Nuclear Reactor Regulation,.
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visited the site. After discussion with the resident inspector, Dr. Murley accompanied them on a tour of the facility. Dr. Murley later met with Maine Yankee management.
6.4 Interface with the State of Maine Periodically, the resident inspectors and the onsite representative of the State of Maine discussed findings and activities of their corresponding organizations.
6.5 Exit Meeting Inspectors periodically held meetings with senior facility management to discuss the inspection scope and findings. At the end of the inspection period, the inspectors presented a summary of findings at an exit meeting on October 13, 1993.
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