IR 05000309/1989017
| ML19325E479 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 10/26/1989 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19325E478 | List: |
| References | |
| 50-309-89-17, NUDOCS 8911070211 | |
| Download: ML19325E479 (28) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
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Report No:
50-309/89-17 l..
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License No:
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Licensee:
Maine Yankee Atomic Power 83 Edison Drive Augusta, Maine 04336
Inspection At: Wiscasset, Maine
l Conducted:
September 1-30, 1989
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. Inspectors:
Ccinelius F. Holden, Senior Resident Inspector f
Richard J. Freudenberger, Resident Inspector
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Eric J. Leeds, Licensing Project Manager Approved by:
& O, k e+M. L solt t let
Ebe C. McCabe, Chief,' Reactor Projects Section 3B Date
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Summary: Inspection on September 1-30. 1989 (Report Number 50-309/89 13 Areas Inspected: Routine resident inspection of plant operations including i
previous inspection findings, special reports, licensee events, operational
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safety, maintenance, surveillance, physical security, radiation protection, and
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fire protection. The inspection involved 170 inspector hours including 23
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backshift and 4 deep backshift hours.
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Results:
Observations of the repair to the Number 2 Main Feedwater Regulating l
Valve indicated a well coordinated and effectively implemented response (detail
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6.a).
Investigation into the Loose Parts Monitor indications was conducted in
a timely and effective manner. Areas for improvement were noted in the imple-
mentation of the Radiation Control Program (details 6.b and 9). An anomaly I
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identified during the routine Reactor Protective System surveillance received good follow-up (detail 7).
Response to a bomb threat to the licensee's cor-l porate offices'resulted in a search of the plant and appropriate notification
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to the NRC (detail 8.).
A review of plant modifications concluded that the
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licensee adequately implements 10 CFR 50.59 (detail 10).
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8911070211 891027
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.i TABLE OF CONTENT
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1.
Persons Contacted....................................................
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S umma ry o f Fa c i l i ty Ac t i vi t i e s.......................................
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3.
Review of Licensee Event Reports (IP 92700)..........................
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4.
Follow-up on Previous Inspection Findings (IP 71707).................
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Operational Safety Verification (IP 71707,40500)....................
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Plant Maintenance (IP 62703).........................................
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Surveillance (IP 61726)..............................................
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8.
Observations of Physical Security (71707)............................
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9.
Radiol og i cal Control s (IP 71707).....................................
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Periodic 10 CFR 50.59 Safety Evaluation Review (IP 37700)............
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State Liaison (IP 94600).............................................
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Exit Interview (IP 30703)............................................
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DETAILS 1.
Persons Contacted
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Interviews and discussions were conducted with various licensee personnel, I
including plant operators, maintenance technicians and management staff.
2.
Summary of Facility Activities a.
The plart was at full power at the beginning of the report period. A power reduction was initiated on September 1 for routine Turline i
Valve Testing.
During the power reduction, difficulty with the
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i Number 2 Steam Generator Main Feedwater Regulating Valve controller resuhed in a power reduction to approximately twelve (12) percent power for repair (see detail 6.a).
The valve controls were repaired and the plant was returned to full power on September 3.
On Septem-ber 28, a power reduction was accomplished for mussel control opera-
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tions.
On the following day, the plant was returned to full power,
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where it remained for the rest of the report period.
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b.
The NRC Region I State liaison Officer, the Division of Reactor Pro-jects Branch 3 Chief, and the Section 3B Section Chief met with Rep-t resentatives of the State of Maine, Office of Human Services to dis-
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cuss interface issues en September 25, 1989.
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On September 27, 1989, the NRC held a meeting with Maine Yankee in
the Region I office in King of Prussia, Pennsylvania, to discuss the Safety System Functional Inspection (SSFI) (Report 50-309/89-80).
Attachment A to this report includes the list of attendees. Attach-ment B contains an outline of the items discussed. The meeting pur-pose was to discuss the issues and the progress Maine Yankee has made
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toward resolving them.
3.
Review of Licensee Event Reports j
The inspector reviewed the following Licensee Event Reports (LERs) to de-termine that reportability requirements were fulfilled and immediate and long term corrective action was taken.
The following LERs were reviewed:
88 10 Plant trip on High heater Drain Tank level 88-11 Inadvertent Safety Injections During Plant Cooldown 89-01 Plant Trip on loss of EHC Control Power 89-02 Environmental Qualification Discrepancies Identified in
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Containment Cable Connector 89-03 Plant Trip Due to Inadvertent Actuation of Generator Protective Relaying
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No inadequacies were identified.
The inspector found the reports to be complete and accurate.
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4.
Follow-up on Previous Inspection Findings Closed - Unresolved Item (50-309/86-02-03) - The unresolved issue was the incorporation of actions specified in IE Bulletin 80-11 into engineering specifications and administrative procedures. The inspector reviewed engineering department procedures 17-21-2, " Engineering Design Change Re-
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quest - Maine Yankee," 17-21-3, "Engineerfng Design Change Request -
(YNSD),"17-226, " Technical Evaluations," and 17-22-1, " Document Revision Procedure." Also, a recently completed modification which changed the i
configuration of a safety-related block wall was verified to be accurately updated in the Masonry Wall Survey Binder.
No discrepancies were identi-fied. This item is closed.
5.
Operational Safety Verification
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Daily, during routine facility tours, the following were checked: manning, access control, adherence to procedures and Limiting Conditions for Opera-L tions (LCOs), instrumentation, recorder traces, protective systems, con-
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trol room annunciators, radiation monitors, emergency power source oper-t ability, operability of the Safety Parameter Display System (SPDS), con-
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trol room logs, shif t supervisor logs, and operating orders. Weekly, selected Engineered Safety Feature (ESF) trains were verified to be oper-
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able. The condition of the plant equipment, radiological controls, secur-ity and safety were assessed. Biweekly, the inspector reviewed a safety-I related tagout, chemistry sample results, shift turnovers, portions of the containment isolatica valve lineup, and the posting of notices to workers.
Plant housekeeping and cleanliness were also evuluated.
The inspectu also observed selected operations to assess safety and com-pliance with the NRC's regulations. The following is noteworthy, i
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Charging Pump (P-14B) Step-up Gear Failure On September 8, the running Charging pump (P-14B), which also func-tions as one of the High Pressure Safety Injection (HPSI) Pumps, ex-perienced a gradual increase in step-up gear bearing temperatures.
No other abnormalities were noted with the pump's operation. When the hottest bearing reached the critical high alarm value (180 F),
the pump was secured. Due to the fact that a single parameter was
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out of range, with no other indication of pump degradation, the lic-ensee considered the pump to be capable of performing its safety
function.
The Plant Engineering Department (PED) was contacted to
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perform an evaluation. The pump was restarted a short time later,
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when PED personnel were available.
Vibration data taken on the step-up gear housing were in excess of Inservice Testing (IST) oper-
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ability requirements. The pump was declared inoperable and the re-l
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l medial act'on of the Technical Specifications was entered.
The spare l
charging pump was promptly aligned to restore the required redundancy i
to of the HPSI system.
The inspector observed portions of the evolu-
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l tion in the control room, and noted that the operators' actions to
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identify, trend, and evaluate the equipment failure and its impact on the operability of the HPSI system were prudent. Disassembly and inspection of the step-up gear revealed that the bearings had worn excessively, however there was no damage to the gears themselves.
The pump is scheduled to be returned to service by mid-October. The licensee has expanded previously scheduled plans to overhaul the spare HPSI pump (P-145) to inr.lude an overhaul of the spare pump step-up gear as soon as the "B" pump is returned to service. The t
need to rebuild the "A" pump step-up gear will be evaluated after the rebuild of the "S" pump.
The inspector considered the licersee's plans for inspection of similar equipment for similar failures to be appropriate.
No operational rafety inadequacies were identified, and operational per-formance was assessed as good.
6.
Plant Maintenance
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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 tag use, equipment alignment, jumper use, personnel qualifications, radio-logical controls for worker protection, retest requirements, and report-ability per Technical Specifications.
Portions of the following maintenance evolutions were reviewed with no unacceptable conditions identified. Additional detailed information and inspector observations are included in the following paragraphs.
Discrepancy Date Report Number Description 9/1 2740-89 Main Feed Regulating Valve (FW-F-207), inves-tigate and repair oscillations.
9/7 3416-89 Emergency Diesel Generator Preventive Main-3477-89 tenance - air start oiler check, generator 3478-89 brushes check, lube oil system check, in-tegral fuel oil tank water and sediment check and air intake cleaner replacement.
9/18-22 3326-89 Loose Part Monitor System (LPMS) indication investigation.
9/14 3157-89 PCC-M-43 and SCC-M-165 reach rod modifi-3158-89 cations, i
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Main Feed Regulating Valve Failure p
On September 1, the licensee commenced a planned power reduction to
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seventy-five percent power for turbine valve surveillance and mussel I
control operations. As described in previous resident inspection reports, the Main Feedwater Regulating Valve (MFRV) to Steam Genera-tor Number 2 was operating erratically.
Licensee efforts to identify
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tha cause 3 the oscillations had been unsuccessful. During the
power re M 'on, the response of the number 2 MFRV became extremely limited o4 ? controller in either automatic or manual.
The operator.watiled feedwater flow with the MFRV isolation valve, a
motor-operated gate valve. This allowed the MFRV to go full open.
The operator then used the MFRV isolation valve in conjunction with the MFRV bypass valve to control feedwater flow to the Number 2 Steam Generator.
The control room operators consulted with the Instrument and Controls (I&C) Section to ensure that the MFRV was operable to fulfill its feed train trip function as rtquired by the Technical Specifications.
The MFRV is required to close on a feed train trip signal and re-quires air to open aga:nst spring pressure. On a feed train trip signal, two solenoid operated valves in the instrument air supply line to the valve diaphragm vent diaphragm pressure to the atmos-phare, allowing the spring to force the valve closed. This function
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was unaffected by the valve positioner failure, therefore Technical Specification 3.22, "Feedwater Trip System," did not apply.
Licensee management evaluated potential courses of action to allow a controlled powcr decrease. Options available included use of the manual handwheel of the MFRV, which would defeat the.
i train trip function of the valve, and as suggested by an I&C technician trouble-shooting the valve controls, the installation of a temporary modifi-cation to supply instrument air directly to the va?ve diaphragm. The latter approach was choJen. A temporary air pressure regulator was
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installed upstream of the Feedwater Trip System Solenoid-Operated Valves to provide local manual control of the air pressure to the valve diaphragm to control the MFRV position.
This arcangement-allowcd for improved contrul of feedwater f?ow while maintaining the Feedwater Trip System opera! +.
With the temporary air pressure regulator installed, a controlled power reduction was commenced. At power levels below approximately
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fifteen percent, the MFRV bypass valve provides sufficient flow to
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allow the MFRV to be isolated The valve positiorer was isolated,
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replaced, functionally tested, and returned to service later the same day.
The inspector obst4.ed several aipects of tha licensee'
.:.ities associated with this effort including: nanagement's cc c
..<. ion of the potential actions available to allow repair of ti fve, the t
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I briefings provided to the operators who locally operated the valve
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and who were in the control room, the conduct of operators both locally at the MFRV and in the control room, and the repair and func-tional test of the MFRV.
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The inspector considered the use of the temporary modification to control the MFRV instead of taking manual handwheel control to be a
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positive action to enhance safe operttions by reducing the amount of time the licensee would be operating the plant under the Technical Specification action statement. The inspector also observed that redundant communication equipment was available between the control r
room and the local manual cuntroller, that briefings provided to the
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operators and I&C technicians were thorough, and that the repair work and functional testing was conducted in a cautious and timely fashion.
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Liccnsee examination of the failed positioner determined that the failure cause was a pinhole leak in a diaphragm in the positioner.
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The licensee had developed a closecut plan matrix to track action items in response to the #2 MFRV oscillations. Action itein to be addressed ta allow completion of the closeout plan included investi-gation of the #1 and #3 MFRV positioners at the next shutdown, an assessment of other critical valves in the facility which have similar positioners for testing and possible replacement during the next refueling shutdown, and a survey of shared industry information to determine if this type of failure is common in the industry.
The inspector assessed the licensee's MFRV positioner repair actions
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Indications on the Reactor Coolant System Loose Parts Monitor To monitor for potential degradation of the thermal shield support system, which has previously generated loose parts, Maine Yankee
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utilizes an acoustical Loose Parts Monitoring System (LPMS).
Five LPMS accelerometers are located on the reactor vessel and one is on each of the steam generators.
On Saotember 12, routine monitoring of the LpMS indicated impacting noises on all reactor vessel and the rui.
r 1 steam generator ac-celeremeters.
The acoustic signal, a.though confirmed, was not of sufficient magnitude to reach the alarm setpoint. Tapes were made
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and sent to a consultant for evaluation.
Initial evaluation indi-ccted that the thermal shield support system was not involved, that the indication was apparently closest to the number I steam generator detector, that the mass was estimated to be between one and ten pounds, and that the potential loose part was likely to be in th-reactor coolant system.
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On September 18, three (3) impacts with sufficient magnitude to cause the LPMS to alarin were recorded. A visual inspection of the Steam Generator Number 1 (SG-1) area was then conducted, temporary ac-celerometers were installed on steam generator 1 and an accelerometer was moved from SG-1 to the loop 1 hot leg isolation valve on Septem-ber 19, 20 and 21 respectively.
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With aid from consultants, the information collected from the tem-porary accelerometers was evaluated. The following conclusions were made:
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The impacts 6:ere originating from the area of the loop I hot leg isolation valve.
2)
The impact energy was sufficiently low that little or no damage was occurring.
3)
Assuming the postulated part was fully loose, its mass was ap-proximate'y four (4) pounds.
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Due to a potantial problem identified at other facilities with ex-cessive stress on the loop isolation valve stems, the motor-operators on the loop isolation valves at Maine Yankee had been modified during the 1988 refueling outage.
In the past, the motor-operators opened the valves based on torque.
The modification changed the motor-
operator to cause the valve to open on limit (valve position) to re-j duce the stress on the stem. This change.. uld also cause the valve to not fully backseat. That might result in impacts being generated by movement of the valve disk.
The licensee evaluated the impact of " disk flutter" on the valve in-
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L ternals. Based on the low energy associated with the impacts, it was concluded that the valve internals were not experiencing excessive stress and that fatigue failure was not a concern. Therefore, fail-ure of the loop isolation valve was not deemed crediole.
The preferred method of terminating a Stetm Generator Tube Rapture L'
(SGTR) is Reactor Coolant System depressucization, and the loop I
isolation valves are not classed as safety-related.
However, the
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valves are available as a secondary means of isolating a faulted
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steam generator in a SGTR event.
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Licensee plans for future action on the LPMS indications included a l
monitoring program with provisions for further evaluating action levels and reactor shutdown criteria.
To correct the impacting, the
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licensee planned an external examination of the valve during the next i
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The purpose of the examination was to verify l
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s that the valve is properly backseated and there is nothing impacting the valve. An overhaul of the valyc was to be scheduled for the Cycle 11/12 refueling outage in the spring of 1990.
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The inspector considered the licensee's response to the identifica-l tion of the part loose part to be timely, well coordinated, and thorough. Management attention r.rovided U =1y review of the data by consultants and the installation of additional equipment to isolate and evaluate the consequences of the impact source.
The identifica-i tion and follow-up of a potential problem during the course of rou-tine monitoring of the LPMS, prior to the indications reaching suf-L ficient magnitude to cause alarms, was an instance of good perform-ance.
As part of the review of the licensee's response to the potential loose part. the inspector observed the containment entry to install
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magnetically mounted accelerometers on the number 1 steam generator.
The containment entry included entries into the loop 1 area during
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t power operation.
Radiation levels in this area are routinely as high
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as 20-30 R/hr.
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The inspector reviewed Radiation Controls Procedure 9.1.32, "Contain-ment Loop Entry at Power," and the Radiation Work Permit (RWP) which authorized the loop entry. No discrepancies were identified.
Brief-ings provided to the crew prior to the containmant entry were thorough. However, while observing the loop entries and work per-formed in the annulus, the inspector made the following observations.
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The installation of one of the magnetic accelerometers required
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l the movement of a small section of insulation on a steam gene-l rator manway.
No respiratory protection was used.
No air samples were taken during this evolution.
The Radiation Con-l trols Supervisor was aware that the insulation would be moved l
and made the decision that air sampling or a respirator was not l
required.
2)
The Health Pb" sics (HP) technician assigned to cover work in the annulus area w e, also assigned as the security escort for one of the individuals working in the annulus area. This hampered the HP technician in monitoring other workers who were in the
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3)
While in the containment, an operator performed a tash which was not part of the job description on the RWP that he had signed in on.
In this instance the inspector concluded that the protec-tive requirements for the task would not have been more restric-tive than the RWP the individual was using.
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Based on prior observation of the use of.apel air samplers for loop en-tries, item 1) above was considered an example of the need for the Radio-logical Cor.trols Department to consistently and conservatively implement the program.
Items 2) and 3) were considered examples of the need for improved support of the implementation of the Radiological Controls Pro-gram by other departments.
These areas for improvement are addressed by
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the Radiological Controls Improvement Plan which was receritly developed by the licencee. The inspector will monitor the implementation of that Plan in accordance with the routine NRC inspection program.
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Overall, maintenance was assessed as good.
For assessment or radiological controls, see detail 9 of +his report.
7.
Surwillance Testing i
The inspector observed ~ parts of tests to assess performance in accordance with approved procedures and LCOs, test results, removal and restoration of equipment, and deficiency review and resolution. The following sur-
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veillances were revieved:
Date Procedurc Number Title 9/1/89 3.1.3/3.1.6 Turbine Valve Testing / Excess Flow Check Valve Testing 9/14/89 3-6.2.2.9 Reactor Protection System Logic Matrix Test 9/21/89 3.17.5.1 Staff Building / Emergency Operating Facili-l ties Ventilation Test l
On September 14, during a Reactor Protective System (RPS) Logic Matrix
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Test, an Instrument ane' Controls technician noticed the AB matrix lights were energized with the selector switch in position 4.
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l normal indication. These lights are sometimes used as isolated circuit ground detectors, and the technician believed that a ground existed in the u
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test circuit but that it did not impact RPS operability.
The condition Wds reported to the I&C supervisor and the remainder of the logic test was I
L completed satisfactorily. The supervisor and the technician continued to l
troubleshoot the anomalous indication. They determined that the problem
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was associated with the bypass key switch for channel
'B' low steam gene-rator level.
The switch appeared to remain in the bypassed position. The L'
problem also affected the AB matrix for low steam generator level. This l
is one of the six logic matrices, and the other five assured the safety
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function remained functional.
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Cycling the switch cleared the condition. The licensee increased the
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logic matrh test frequency to once a day for a week, and then once a week for a month. There was no recurrence. The plant will soon return to i
l monthly testing. The licensee continues to investigate replacement parts and plans to replace the switch during the next available shutdown. Until
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l-that time, the licensee plans to continue additional testing of the logic u
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matrix to identify recurrence. The inspector reviewed Technical Speci-
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fication 3.9 for minimum operable channels for the RPS and concurred that the licensee was in compliance while the switch was stuck. Also, the in-spector concluoed that the I&C group showed good follow-up of an ar;o.aalous i.
indication when the results for the original logic matrix test were satis-
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factory.
Overall, surveillance performance was assessed as good.
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Observations of Physical Security 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 vital area barriers, vehicle searches and personnel identification, access control,. badging, and compensatory measures when required.
On September 7,1989, the licensee corporate offices, along with several
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other utilities and a local television news station in the State of Maine, received bomb threats to their offices. The threats centered around the corporate headquarters and transmission lines of these power companies.
The threats indicated that the placement of the ' combs was the responsi-bility of the " Boston Division of the New England Environmental Army."
The Maine Yankee Atomic Power Station, Wiscassat, Maine was specifically.
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exempted from the threats because of envirornental considerations.
The licensee evacuated and searched the corporate offices. No explosive de-vices were found.
The licensee's security supervisor at the Maine Yankee Atomic Power Sta-tion was notified of the thretts at 11:30 a.m. on September 7, 1989 and a heightened security awareness was initiated in and around the station.
Searches for explosive devices were conducted by both plant operations and security personnel in both the protected and owner-controlled areas. No explosive devices were found. The heightened security awareness posture was maintained through September 11.
The inspector observed portions of the search and the licensee's report-ability determinations.
Searches of all areas of the plant were conducted by teams consisting of a security officer and an operator, coordinated by a security supervisor. The Nuclear Safety Engineer made a repcrt to the Headquarters Operations Center using the Emergency Notification System (ENS) in accordance with 10 CFR 50.72. The inspector considered the lic-ensee's actions to sear::h the facility, knowing that it had been speci-fically exempted from the threat, to be conservative.
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Radiological Controls Radiological controls were observed on a routine basis du.
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radiation exposure control, and contamination con; ef.
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t-radiological work practices, conformance to radiological control proce-
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dures 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, i
The inspector witnessed a planning meeting for the low pressure safety injection pump overhaul. The meetirg reviewed the problems encountered in
the removal of the pump. Although the meeting partir.ipants discussed a l-number of problems that were encountered during the removal process, there was no firm direction on how those problems would be corrected as a result o f this meeting. The lack of Engineerir.g representation left a number of
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specific questions concerning the measurements of spider bearing weer open. Also, the areas of contractor training, worker and health physics coordination were discussed in general terms but no determination of
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specific actions were made.
The inspector concluded that more specific information and action items would have been appropriate.
Overall, based on the above observations and.those in Detail 6.b of this report, radiological controls performance was assessed as adequate but not significantly above the minimum requirements established by NRC regula-tions.
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periodic 10 CFR 50.59 Safety Evaluations Review The following procedures were reviewed to determine if sufficient program-matic guidance existed for the conduct of safety evaluations pursuant to 10 CFR 50.59.
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' Procedure No. 0-06-4 Rev. 2 10 CFR 50.59 Determination
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Procedure No. 17-21-2 Rev. 4 Engineering Design Change Request - Maine Yankee Procedure No. 17-21-7 Rev. 2 Safety Analysis Procedure No. 17-21-1 Rev. 2 Permanent Plant Modifications No unacceptable conditions were noted.
i The following four Engineering Design Change Request (EDCR) packages and the accompanying 10 CFR 50.59 reviews were also inspected.
EDCR 86-04 - Primary and Secondary Component Heat Exchanger
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Replacemant.
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EDCR 88-36 - Control Room Air Conditioning Cooling Upgrade EDCR 88-45 - Containment Air Compressor System Improvements.
EDCR 88-52 - PCC/ SCC Monitoring 1.97 Modifications.
No unacceptable conditions were identified.
It was concluded that the 10 CFR 50.59 process with regard to engineering modifications was acceptable, and that the associated licensee performance was good.
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State Liaison
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Periodically, the resident inspectors and the onsite representative of the State of Maine discussed their findings with each other. No unacceptable plant conditions were identified.
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Exit Interview L
Meetings were periodically held with senior facility management to discuss the inspection scope and findings. The inspector continually meet with
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the State inspector to discuss the status of inspection findings. A sur-i mary of findings for the report period was also discussed at the conclu-sion of the inspection. The licensee did not identify any 10 CFR 2.790
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material as being within the scope of the inspection.
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ATTACHMENT A Attendees at the Maine Yankee Safety System Functionel Inspection (SSFI) meet-
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ing held in the NRC Region I Office on Sentember 27, 1989.
Maine Yankee
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Name Ti tle
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J. Garrity Vice President, Licensing and Engineering J. Hebert Manager, Plant Engineering
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D. Whittier Manager, Nuclear Engineering and Licensing S Nichols Licensing Section Head
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NRC Name Title W. Hodges Director, Division of..leactor Safety J. Johnson Chief, Reactor Projects Branch 3 E. McCabe Chief, Reactor Projects Section 3B G. Kelly Chief, Technical Support Section C, Holden Senior Resident Inspector, DRP J.'Lyash Project Engineer, DRP D. Caphton Senior Techt.ical Reviewer, DRS C. Woodard Reactor Engineer, DRS A. Giancatarino ERC Environmental and Energy Services Company
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ATTACHMENT B T0
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IR 50-309/89-17 g
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INSPECTION PRACITCES
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, Resolve Concerns Before Exit
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Prioritize Remaining Items
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INTRODUCTION
' G. D. Whittier t-K'?
NRC Issees from SSFI i
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Safety Perspective Consistency J. R. Hebert
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Cross Tie of DC Buses S. E. Nichols
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Short Circuit Protection J. R. Hebert
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and Motor Thermal Protection-
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Component Cooling Heat Balance J. R. Hebert
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. Component Substitution Process J. R. Hebert i
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o Component Labeling S. E. Nichols o
Valve Maintenance Consistency S. E. Nichols
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o Surveillance Tests for Check Valves S. E. Nichols o
Resolution of SSFI Matrix Items J. R. Hebert CLOSE J. H. Garrity F
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Phasa I Interna! Events Complete
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CTMT Analysis December.1991 '
Phase II
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Offsite Consequences December,1992'
Phase III
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Application of PRA Results to Operations / Maintenance:
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Prioritization of Maintenance Activities
Component Important by Itself oo (Significance - Moderate, Large, Very Large)
Impact of Degradation Due to Other Degraded Equipment oo
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Relative Frequency of Problems with Components / Systems oo
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ADMINISTRATIVE CONTROUS FOR DC BUSES:
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Limitations en Ability to cross-Tie DC Buses:
o Done "only in cases of immediate need...."
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7 day limit whenever in Hot Standby / Power Ops (OP 1-22-2).
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Operability of DC. Buses 1 & 3 at Cold Shutdown:
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be in service."
(OP 1-22-2)
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Valve Maintenance Consister.y co'
Two Similar Check Valver, with Dissimilar PMs
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oo No Recurring Operational Concerns Identified
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Root Cause Analysis of PCC-M-43 and SCC-M-165 Failures
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Control of Instrument Setpoints
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Instrument Found Out of Calibration
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Stud' (tesults to Date
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Performance Testing of Heat Exchangers L
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Improved Instrumentation i
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Shrink and Swell L
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More Detailed / Computerized Calculations
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Air Accumulator Capacity
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Reset to.92 Psig
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Minimum Pressure for System Performance
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