IR 05000309/1989025
| ML20006G139 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 02/21/1990 |
| From: | Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20006G137 | List: |
| References | |
| 50-309-89-25, NUDOCS 9003050133 | |
| Download: ML20006G139 (11) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
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d II Report No:
50-309/89-25-License No:
DPR-36 s
- Licensee:-
Maine' Yankee Atomic Power Company a
Facility:
Maine Yankee Atomic Power Plant, Wiscasset, Maine Dates:
December 12, 1989 through January 22, 1990 Inspectors:
Cornelius F. Holden, Senior Resident Inspector Richard J, Freudenberger, Resident Inspector
~ Approved by:
SO1O b lLfttl10 i
E. C. McCabe, Chief, Reactor Projects Section 3B Date Summary: Resident Inspection Report 50-309/89-25 Scope: Routine resident inspection of special reports, events, operational
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safety, maintenance, surveillance, physical security, radiation protection, fire protection, and licensee assessment of quality improvement. The inspec-tion involved 106 inspector hours including eighteen (18) backshift and eight I
- (8). deep backshift hours.
Results:
Operator performance during the plant shutdown on January 15 (Detail 1.a.) and--
in response-to the dropped Control Element Assembly event on January 9 (Detail 3 a.).was noted'to 'e good.
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Unresolved Item 89-25-01 was.. identified on the issuance of termination radi-ation exposure information within the time required by 10 CFR 20.408-(Detail
' 2.b.).
Radiological' control program review noted good causal analysis of per--
sonnel ' contaminations during 1989 (Detail 2.a.) and the identification of sig-l nificant improvement opportunities during maintenance on radioactive equipment (Detail 3.b.).
Two design issues were identified by Maine Yankee during the report period.
- One is considered Unresolved Item 89-25-02 and refers to the alignment of the
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. containment spray building ventilation system (Detail 5.a.).
The other per-tains to adequacy of Inverter 5 installation (Detail 5.b.).
'A concern was identified regarding the licensee's method of evaluating and cor-
recting the acceptance t.riteria for a component cooling water valve cycle time.
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This is considered Unresolved Item 89-25-03 (Detail 1.a).
m y.S 9003050133 900222 U
PDR ADOCK 05000309 J
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TABLE OF CONTENTS
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1.
Operations (IP71707)................................................
2.
Radiological Control-(IP 71707)......................................
1 3.
Maintenance / Surveillance (IP 62703,61726)...........................
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Security (IP 71707)..................................................
5.
Engineering / Technical Support (IP 71707).............................
l 6.
Safety Assessment / Quality Verification (IP 71707)....................
7.
Administrative.......................................................
a..
Summary of Facility Activities..................................
b.
Interface with the State Of Maine (IP 94600)....................
c.
Exit Meeting (IP 30703).........................................
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DETAILS 1.
Plant Operations During routine daily facility tours the following were checked: manning, access control, adherence to procedures and Limiting Conditions for Opera-tion, instrumentation, recorder traces, protective systems, control room annunciators, radiation monitors, emergency power source operability, operability of the Safety Parameter Display System (SPDS), control room logs, shift supervisor logs, and operating orders.
Plant equipment condi-tion, radiological controls, and security were assessed. On a weekly basis, selected Engineered Safety Feature (ESF) trains were checked for-operability. On a biweekly frequency the inspector reviewed a safety-related tagout, chemistry sample results, shift turnovers, portions of the containment tsolation valve lineup and the posting of notices to workers.
Plant housekeeping and cleanliness were also evaluated.
NRC review concluded that the areas inspected were generally acceptable.
The following item was considered noteworthy.
a.
Component Cooling Water Isolation Valve Stroke Time Criteria On November 13, 1989, stroke time testing was performed in accordance with operations surveillance procedure 3.1.20, " Safeguards Valve Testing," on valves SCC-A-460 and SCC-A-461 as a maintenance func-tional test following a packing adjustment. These valves are pneu-matically operated, butterfly isolation valves to the non-safety-related portions of the secondary component cooling water system.
The test results were 5.5 seconds for SCC-A-460 and 2.5 seconds for SCC-A-461. The acceptance criterion listed in the operations sur-ve111ance procedure was a stroke time in the close direction of four seconds or less.
However, Plant Engineering Department Procedure 3.17.8.2, "ISI/IST Valve Tests for Discrepancy Reports or Repair Orders," contained acceptance criteria of ten (10) seconds or less.
Since the operations department relies on the Plant Engineering De-partment for valve stroke time acceptance criteria, Procedure Change Report (PCR)89-142 was written to change the operations department procedure (3.1.20) acceptance criteria to agree with the Plant Engi-neering Department procedure (3.17.8.2).
The test was declared suc-
-cessful based upon the change to the acceptance criterion.
On January 15, 1990, a plant Shif t Supervisor identified that PCR 89-142, which changed the allowed stroke time for valves SCC-A-460
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and-SCC-A-461 from four seconds to ten seconds, was incorrect based on Yankee Atomic memorandum MYP 89-306, " Review of Surveillance Pro-cedure 3.1.20, Secondary Component Cooling Water Trip Valve Testing,"
dated April 4,1989. SCC-A-460 was declared inoperable at 4:15 p.m.
and a 72-hour Technical Specification remedial action statement was entered. A plant shutdown was commenced at 7:15 p.m. to perform stroke time testing and maintenance on SCC-A-460. The NRC Operations
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p center was notified of the plant shutdown in accordance with 10 CFR 50.72 The inspector observed portions of the plant shutdown from '
l the control room and witnessed the stroke timing of the valves locally.> As-found test results were 2.7 seconds for SCC-A-460 and 2.4 seconds for'5CC-A-461. The valves were declared operable, Tech-nical Specification Remedial Action was exited,' and a plant startup
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was commenced.
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Licensee review of the discrepancy between the operations department surveillance procedure and the engineering department procedure found that both departments had received the information-included in the-above mentioned Yankee Atomic memorandum.
The operations department updated their surveillance procedure promptly. The engineering de-partment evaluated'the change in the acceptance criteria from ten seconds to four seconds and found that, historically, the valves stroked in three seconds or less.
Therefore, there were no immediate operability concerns and a procedure update prior to the next refuel-ing interval test of the valves was accepted. The determination that the procedure change could be delayed apparently did not include con-sideration of the fact that the engineering procedure is used as a reference for maintenance functional testing requirements.
The inspector considered plent management's actions to shut down the:
plant to resolve the operability question of SCC-A-460 to be appro-priate.- Plant maneuvers for the shutdown were well controlled, with shift personnel coordination notably good. However, the acceptabil-ity of the licensee's program for control of acceptance criteria for safety related valve stroke time tests is considered unresolved-
-(50-309/89-25-03).
2.
Radiological Controls Radiological controls were routinely observed. Areas reviewed included overall organization and management for external radiation exposure con-trol and contamination control.
Radiological work practices, conformance to' radiological control procedures and 10 CFR Part 20 adherence were evaluated.
a.
1989 Personnel Contamination Event Summary The inspector reviewed the 1989 Personnel Contamination Event Sum-mary. This report was prepared by the Radiological Controls Section and provided to all plant departments with personnel who work in the h
Radiological Control Area. The report contained breakdowns of all the personnel contaminations by department, individual, root cause, date, plant areas, ' dose and location of the contamination (shoes, shirt, skin,etc.) The inspector considered the report to be a potentially valuable tool for identifying areas for improvement. No unacceptable conditions were identified.
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b.
Termination Exposure Letter During the quarterly dosimetry change on December 30, 1989, dosimetry b
personnel identified a discrepancy involving a contractor employee who had terminated employment on September 28, 1989. The licensee's staff identified that the report of personnel monitoring required by 10 CFR 20.408 was not completed within the time frame established by that requirement. A verification of employment status was completed during December, but the individual's terraination had not been iden-
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tified to the dosimetry office at that time.
Immediate actions were-
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to issue the termination letters and evaluate the occurrence for re-portability.
No report to the NRC was deemed necessary by the licen-see.
Long term corrective actions included a review of all depart-ments' employment status and evaluation of the terwination process to assure that' suitable controls will prevent recurrence. By the end of
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the report period the licensee had identified several other contrac-tor employees who had terminated employment at Maine Yankee without-notification to the dosimetry office.
Revision of the termination process for contractor employees was under evaluation. This item is s
unresolved pending further inspection (UNR 50-309/89-25-01).
The inspector had no additional questions. Overall performance was acceptable.
3.
Maintenance / Surveillance L
The inspector observed and reviewed maintenance and problem investigation
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activities to verify overall plant safety and compliance with regulations, administrative and maintenance procedures,-codes and standards. The in-spector also reviewed proper QA/QC involvement, safety tag use, equipment alignment, jumper use, personnel qualifications, radiological controls for worker protection, retest requirements, and reportability per Technical l
~ Specifications.
L Also, the inspector observed parts of surveillance tests to assess per-l formance in accordance with approved procedures and Limiting Conditions l
for Operation, test results, removal and restoration of equipment, and L
deficiency review and resolution. The following activities reviewed were l
considered noteworthy:
s a.
Control Element Assembly 54 (CEA-54) Dropped On October 30, 1989, a power supply associated with CEA-48 failed.
During power supply replacement, the CEA dropped into the core. As described.in Region I Inspection Report 50-309/89-18, Detail 4.a. NRC p
inspection concluded that the licensee had acted prudently to prevent I
damage to equipment located in a high radiation area and was knowl-l edgeable of the risk of dropping the CEA prior to authorizing the wark to begin.
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A similar power supply failure affected CEA-54 on January 9,1990.
Again the licensee consulted with the vendor to generate a procedure for the replacement of the power supply while minimizing the risk of i
dropping the CEA.
The inspectors observed activities in the Control Room and at the Control Element Drive. Mechanism (CEDM) logic cabinets
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during the replacement of the control circuitry for CEA-54. Discus-sions were conducted between the operators and the I&C technicians to identify high risk steps of the evolution and the licensed operators
reviewed Abnormal Operating Procedure 2-21, "Misa11gned (Dropped)
CEA," prior to authorizing the start of work.
Communications were
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established between the work location and the Control Room. The CEA dropped subsequent to the installation of-a temporary power supply
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which was intended to hold the CEA while the faulty supply was re-placed. A power reduction was conducted and reactor parameters were a
verified to be within limits in accordance with Technical Specifica-tion requirements. The power supply was replaced and the CEA was withdrawn a short time later.
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During the previous operating cycle, Control Element Drive Mechanism (CEDM) control circuitry problems resulted in a number of inadvertent dropped Control Element Assemblies (CEAs).
The licensee identified short and long term actions to improve the reliability of the CEDM
control circuitry.
The final long term action to be completed by the licensee is to modify the CEDM control circuity to install redundant power supplies, the failure of which has been common to many of the dropped CEAs. This modification is to be installed during the next refueling outage, scheduled to begin April 7, 1990.
Although the January 9 maintenance associated with the control cir-cuitry of CEA-54 resulted in dropping the CEA and was similar to the October 30 dropped CEA, the inspector concluded that the licensee had acted prudently. The evolution was well controlled and coordinated.
b.
Letdown Prefilter Replacement
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On December 27 and 28, the inspector observed portions of the main-tenance involving replacing the Letdown System Pref 11ter (FL-35A).
j This process included the removal of the internals, transport to the storage bunker, and replacement of the filter assembly.
The used filter assembly had a contact dose rate of approximately 300 R/hr. A shielded cask was used to transport the assembly to the storage bunker, where the filter assembly was placed inte a High Integrity Container (HIC) suitable for disposal.
The evolution was conducted in an acceptable manner, but the exposure estimate (300 mrem) for the evolution was exceeded by approximately ten percent and a worker was contaminated after performing an unanti-cipated task.
Licensee observation and evaluation, documented in a radiological observation report, identified several factors that may
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have' prevented these difficulties.
These included improved preplan-ning, with a dry'run and potential equipment enhancements such as a permanent solid lifting rig to permit shield rigging time and associ-
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ated exposure reduction. Af ter the filter assembly was lowered into the housing, it was discovered that the bottom plate of the filter
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assembly was improperly installed.
The worker was contaminated dur-ing repositioning of the the now highly contaminated lower plate.
The-personnel contamination could have been prevented by improved
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contamination control techniques or more thorough verification of proper installation of the lower support plate.
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The inspector observed that, although weaknesses were evident in this evolution, the licensee was identifying opportunities to improve per-formance of the maintenance evolution.
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Turbine-Driven Auxiliary Feed Pump Monthly Surveillance f
The' inspector observed the return to service of the turbine-driven
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auxiliary feed pump (P-25B) following a monthly test run in accord-ance with operations surveillance procedure 3.1.5, " Emergency and Auxiliary Feed Pump Test." The evolution involved verification of t
proper valve and equipment alignment by a non-licensed operator. The operator appropriately performed the evolution in accordance with the procedure.
NRC observation found acceptable performance, d.
Control Room Ventilation Monthly Surveillance Test The inspector observed the performance of operations surveillance procedure 3.1.18, " Control Room Ventilation Filter Flow Verifica-tion." The test was conducted by a licensed control room operator and consisted of realignment of the control' room air conditioning system to the recirculation mode and operation of the control room breathing air system.
The operator noted that one of the two outside air' supply motor operated valves had a stroke time that was approach-ing the " Alert" value specified for Inservice Test (IST) require-ments. Although no action was necessary to comply with the proce-i dures at that-time, the operator conservatively contacted the Plant Engineering Department for evaluation of the condition of the valve.
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The inspector considered the operator's actions to be indicative of careful attention to off normal equipment performance.
4.
Security
' regulatory requirements, the physical security plan, and approved proce-
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. Inspector checks were made to determine whether security conditions met dures. Those checks included security staffing, protected and vital area barriers, vehicle searches, personnel identification, access control, badg-ing, and compensatory measures when required. No deficient conditions were identifie x
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Engineering / Technical Support a.
Containment Spray Building Ventilation The-containment spray building houses the containment spray pumps, the low pressure safety injection pumps, the residual heat removal
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(RHR) heat exchangers and the piping, valves, and instrumentation associated with these safety related systems.
On December 6, 1988, with the plant in a refueling shutdown, air flow measurements were taken with several different ventilation system damper, louver, and floor plate position configurations. Air flow measurements were taken in the "as found" condition as well as the configuration described by drawings included in the Final Safety Analysis Report (FSAR). The "as found" data indicated that ventila-tion air flow through a low pressure safety injection pump motor cubicle was less than half of the flow in the FSAR drawing. With the-ventilation system aligned as shown in the FSAR drawing, the venti-lation air flow through the pump motor cubicles was measured to be within two (2) percent of the value shown on the FSAR drawing.
At the completion of the air flow measurements, the ventilation sys-tem was returned to its "as found" configuration. There was no indi-cation that the preliminary results of the flow measurements were communicated to the Operations Department prior to startup from the refueling outage.
In a memorandum dated January 4, 1989, the engineers involved in the air-flow measurements reported their_ findings to the Vice-President, Engineering and Licensing. The memorandum contained a description of the tests performed, results, and recommended corrective actions. A copy.of the memorandum was forwarded to the operations department and entered into the department's Task Management System (TMS).
On December 29, 1989, while reviewing his TMS assignments, a Plant Shift Superintendent (PSS) identified that the containment spray building ventilation system was still aligned in the "as found" test configuration and that none of the recommendations included in the memorandum had been acted u on. He took immediate action to have the system realigned as described in the FSAR drawings and initiated an evaluation of the operability of the safety related systems in the containment spray building. The evaluation was based on the poten-tial that insufficient air flow had existed through the pump motor cubicles to prevent over heating of the motors during a design basis
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accident scenario. Mai_ne Yankee review of the FSAR design drawing for the system indicated that the flow values were based on the divi-sion of the purchase specification flow capabilities of the exhaust fans among the pump motor cubicles.
No specification of the minimum air flow (required through the motor cubicles to consider the systems operable) existed.
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As of December 29, 1989, when the system was realigned as described in the' FSAR drawing, the systems in the containment spray building were considered operable. This determination was based on test data (from the December 6, 1988 test) that showed the' air flow through the pump cubicles to be within a close tolerance.of the values specified in the FSAR. The inspector verified the alignment of the system to-be in accordance with the FSAR drawing configuration.
-The operability determination of the systems in the containment spray building prior to December 29, was assigned to the Nuclear Safety Section. An engineering evaluation of the required air flow through the pump motor cubicles to prevent overheating the motors as well as the impact on environmentally qualified equipment in the building was underway at the end of the report period. Also, the Quality Assur-ance Section has initiated a review of the Engineering Department /
Operations Department interface associated with this issue due to the delay in. implementing corrective action to realign the containment
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spray building ventilation system.
This issue is unresolved pending the' completion of the above evaluations and further NRC inspection (UNR 50-309/89-25-02),
b.
Emergency Battery 2 in Condition Outside Design Basis The 125 VDC system consists of four battery busses.
Each battery bus is continuously supplied by a separate battery charger.
If a battery charger fails, the battery supplies the bus loads.
Busses 1 and 4 or
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2 and 3 can be cross-tied. When in the cross-tied condition, train separation is eliminated.
Final Safety Analysis Report (FSAR)
assumptions include the capability for any battery to have the capac-ity.to supply the loads on its associated busses for two (2) hours when in the cross-tied condition.
Battery bus 2 orininally supplied Inverter 2 and its safety-related loads only.. A design change added an additional inverter (Inverter 5) to supply non-safety-related loads during normal operation. At the time of installation, it was intended that Inverter 5 trip during conditions that would require the safety-related loads to be supplied by the battery (i.e., station blackout).
During engineering review of options to improve reliability of the safety-related inverters, it was identified that Battery Bus 2 would not be. capable of meeting the FSAR assumptions due to inability of Inverter 5 to shed its loads from the bus as intended by the design.
Inverter 5 supplied several non-safety-related loads and was intended to trip ten (10) minutes after Battery Charger 2 stopped supplying the bus. The circuit intended to trip Inverter 5 was found to be incapable of performing this function. The resulting loads which remained connected to the battery would have resulted in Battery 2 depletion in significantly less than the two hours assumed by the FSAR.
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The loads supplied by Inverter 5 were transferred to alternate sup-plies and Inverter 5 was de-energized and administrative 1y controlled y
by tagging the equipment out of service.
The as-found condition of battery bus 2 was outside design basis assumptions, therefore the NRC was notified in accordance with 10 CFR 50.72 (b) (2).
The inspector observed that, once thr. deficiency was identified, the
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actions to correct it were technically correct and timely.
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A cause determination had not been made at the end of the report period. Maine Yankee plans to submit a Licensee Event Report on this
subject.
The inspector plans to review the cause determination and
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corrective action as a result of this issue as part of the LER review in accordance with the NRC inspection program.
6.
Safety Assessment / Quality Verification a.
Periodic Quality Improvement Assessment The Quality ~ Programs Department recently issued a report of its first attempt to assess quality improvement efforts at Maine Yankee.. The assessment covered the period from November.1988 to October-1989.
The report was orgsnized by functional areas.
It included a listing
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of internal evaluations and audits as well as external inspections conducted during the period, a summary of the findings, and an evalu-ation of future internal areas for emphasis.. The inspector consi-
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l dered the assessment to be a positive initiative.
7.
Administrative-a.
Summary of Facility Activities
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At the beginning of the report period the plant was operating at full power.
Power reductions to approximately seventy-five. percent power were accomplished on December 20 and January 9 for a generator condi-e tion monitor alarm and a dropped control element assembly (see Detail
'3.a.),respectively.
On January 15, the plant was shut down for testing of an isolation valve in the secondary component cooling water system (see Detail 1.a.) The plant operated at full power for the remainder of the report period.
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A Safety System Functional Inspection Follow-up inspection was con-ducted from January 16 to 19. -The results of the inspection will be documented in Region I Inspection Report 50-309/90-80.
A meeting was held in the Region I office on January 9 to discuss several recent plant occurrences and is documented in NRC Meeting Report No. 50-309/89-24 dated January 31, 1990.
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Interface with the State of Maine-Periodically, the resident inspectors and the on site representative of the State of Maine discussed findings and activities of their cor-
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responding organizations. No unacceptable plant conditions were
identified.
c.
Exit Meeting
Meetings were periodically held with senior facility management to
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discuss the inspection scope and findings. A summary of findings for the report period was also discussed at the conclusion of the inspec-tion.
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