IR 05000309/1980005
| ML19312E049 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 04/17/1980 |
| From: | Lazarus W, Martin T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19312E048 | List: |
| References | |
| 50-309-80-05, 50-309-80-5, NUDOCS 8006030109 | |
| Download: ML19312E049 (7) | |
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O U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION I
Report No.
80-05 Docket No.
50-309 i
License No.
'DPR-36 Priority Category C
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Licensee:
Maine Yankee Atomic Power Company 25 Research Drive Westborough, Massachusetts 01581 Facility Name:
Maine Yankee Nuclear Power Station Inspection At:
Wiscasset, Maine
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Inspection Conducted:
March 17-21, 1980 Inspectors:
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. l{zarJM, RedrJtor Inspector date
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date date Approved b :
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T. T. Marti k Chief, Reactor Projects
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Section No. 3, RO&NS Branch Inspection Summary:
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Inspection on March 17-21, 1980 (Report No. 50-309/80-05)
Areas Inspecteri:
Routine, unannounced inspection of licensee emergency pro-cedures for responding to small-break loss of coolant &ccidents; followup on Licensee Event' Reports (LERs); and review of Monthly Operating Reports. The in-spection involved 30 inspector-hours on site by one region-based inspector.
Results:
No items of noncompliance were identified.
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1 Region I Form 12 (Rev. April 1977)
8006030
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DETAILS
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1.
Persons Contacted
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P. Anderson, Administrative Department Head R. Arsenault, Plant Shift Superintendent J. Brinkler, Technical Support Department Head T. Davin, Plant Shift Superintendent W. Paine, Operations Department Head R. Radasch, I&C Supervisor
- E. Wood, Plant Manager D. Stevenson, Plant Shift Superintendent The inspector also interviewed several operators, technicians and members of the licensee's technical and administrative staffs.
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denotes those present at the exit interview.
2.
Review of Small-Break LOCA Procedures a.
Procedure Implementation The inspector reviewed the emergency proc 6dures implemented by the licensee, which would be used in the event of a small-break loss of coolant accident (LOCA), to determine if the guidelines developed by the CE Owner's Group had been incorporated.
These guidelines were submitted to NRC in report CEN-114-P (Amendment 1P) and were subse-quently approved by NRR for implementation in letters dated November 14, 1979 and December 26, 1979.
The following procedures were reviewed:
2-12, Loss of Reactor Coolant, Rev~ision 8 2-13, Major loss of Reactor Coolant, Revision 8 2-1, Emergency Shutdown, Revision 5 2-9, Natural Circulation, Revision 1 2-14, Long Term Core Cooling Realignment, Revision 2 1-7, Plant Cooldown, Revision 11 1-7-1, Plant Cooldown By Abnormal Methods, Revision 3 Except as noted below, no discrepancies were identified. CE Guidelines are referenced in parentheses.
A diagnostic chart is not included in the procedures to aid
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operators in differentiating between 1.0CA, steam line break, or steam generator tube rupture. (Chart is appended to CE Guidelines).
Operators are not required to verify adequate feedwater flow as
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one of the immediate actions. (Immediate Action No. 4).
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Realignment of safety injection to the hot legs is covered in
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procedure 2-14, but the operator is not referred to 2-14 if he is following 2-12 or 2-9. (Follow-up Action No. 2).
I Alternate methods of cooldown (other than steam dumps / atmospheric
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damps) arc not addressed. (Follow-up Action No. 4).
Operators are not cautioned concerning continuing Containment
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Spray unnecessarily. (Precaution No. 6).
PA8 radiation and sump levels are not required to be monitored
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for indication of leakage from safety injection systems after recirculation actuation (RAS). (Precaution No. 9).
Operators are not cautioned to sample reactor coolant activity
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prior to initiating shutdown cooling.
In addition to the above discrepancies with the CE Guidelines, the inspector made the following observations concerning these procedures:
"Immediate Actions" in procedure 2-12 requires the operator to
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initiate procedure 2-1, Emergency Shutdown from Power. This could be clarified by specifying performance of the immediate
=ctions of procedure 2-1.
In procedure 2-12, step 5.3.1, the operator is instructed to verify subcooling by reference to the Tsat meter. He should be cautioned to verify that meter by checking temperature and pressure against the posted saturation curve.
Location and isolation of the leak may deserve a higher priority
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in procedure 2-12.
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Based on interviews with several operators, the immediate
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action steps to isolate primary sample lines and to close loop fill and drain valves, would be more appropriate as a high priority subsequent action.
Procedures should end with a referral to the next procedure to
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be followed in getting the plant to the cold shutdown condition.
A licensee representative stated that the small-break LOCA procedures were currently being reviewed in preparation for making another set of revisions, and that the above discrepancies / comments would be evaluated for inclusion.
The discrepancies with the CE Owner's Group Guidelines will be referred to NRR for resolution.
This item will remain unresolved pending the results of the NRR review.
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b.
Trainina Requirements The inspector reviewed licensee training records and interviewed several licensed operators to verify that they had received formal training in the revised small-break LOCA procedures.
No discrepancies were identified.
c.
Operator Interviews The inspector interviewed three senior reactor operators and two reactor operators to determine their familiarity with procedure 2-12, " Loss of Reactor Coolant." Except as noted below, no discre-pancies were identified.
The immediate action steps of the emergency procedures should be performed from memory, however none of the five interviewed operators remembered to isolate the primary sample lines and the loop fill and drain valves. As these are not urgent requirements, the inspector discussed roving them to the " Subsequent Action" section of the procedure where they may be more appropriate.
The licensee is evaluating this for possible inclusion in a revised procedure. (See Detail 2.a. above).
A subsequent review of modified procedures will be performed following NRR evaluation of the differences between the licensee procedures and CE Owner's Group Guidelines. (50-309/80-05-01)
3.
In Office Review of Licensee Event Reports (LERs)
The inspector reviewed the following LERs received n the RI office to verify that details of the event &ere clearly reporned including the accuracy of the description of chose and adequacy oi' corrective action.
The inspector also determined whether further inforwation was required from the licensee, whether generic implications were indicated, and whether the event warranted en site followup.
- --80-01, Low Power Steam Line Break Safety Analys s Error
--80-02, Containment Personnel Airlock Inner Door Seal Leakage
- --80-03, Diesel Generator Output Breaker Failure
--80-05, Safety Injection Header Drain Line Crack
- --80-06, Improper Alignment of Steam Generator Level Channels Except for those LERs designated (*), which were selected for on site followup, the inspector had no further questions concerning these LER e-
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4.
On Site Followup of LERs During on site followup, the inspector verified that reporting requirements of Technical Specifications and Regulatory Guide 1.16 had been met, that appropriate corrective action had been taken, that the event was reviewed by the licensee as required, and that continued operation of the facility was conducted within Technical Specification limits. The review included discussions with licensee personnel, review of PORC meeting minutes, Plant Information Report (in-house reports), and applicable logs. The following LER was reviewed.
80-01, Low Power Steam Line Break Safety Analysis Error
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During followup on this LER, the inspector identified that the required fourteen day followup report had not been submitted. The followup report had been written by the licensee, but due to an administrative oversight, had not been mailed. The steam line break analysis error is only a concern late in core life. NRR has approved plant startup and the proposed test program proposed by the licensee in a letter dated February 26, 1980.
During startup testing the licensee will determine the required position of the feedwater regulating bypass valves to provide required flow but prevent a restart. The results of this testing and performance of modifi-cations to the feedwater regulating system will be reviewed in a subsequent inspection. (50-309/80-05-02)
80-03, Diesel Generator Output Breaker Failure
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The inspector reviewed the circumstances surrounding the breaker malfunction with the maintenance personnel involved and determined that the malfunction of the breaker is generic to GE Type AM-4.16 circuit breakers. The licensee has inspected all of these breakers in use in safety related systems and corrected the problem whera identified by reattaching the prop reset spring to another point on the prop, effectively increasing the spring tension, which should keep the prop from bouncing when the breaker closes.
This information has been forwarded to IE Headquarters for dissemination to other licensees.
80-06, Improperly Aligned Steam Generator Level Circuits
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The inspector reviewed the instruction bulletin for the Fischer and Porter 1302495 Level Transmitter and discussed the event with the I and C Supervisor. The inspector determined that the trip channels remained operable even though improper alignment of the oscillator amplifier had been performed. Failure would have been evident by a false high level turbine trip signal.
The inspector reviewed the calibration records to verify that proper realignment had been accomplished.
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5.
Observation of Plant Operations The inspector observed plant operations in the Control Room several times during the inspection verifying that required licensed operators were on watch; that the required monitoring instrumentation was operable, and that, where applicable, values were within Technical Specification limits; status of annunciators; and operability of selected safety systems.
The plant was being operated between 15 and 45 percent power.
Annunciators: Safety Injection Tanks 1, 2 and 3 Low Pressure
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The actual pressures were within T.S. limits, but be'.ow the reset point of the pressure switches. By the completion of the inspection, the pres-sure switches had been reset and the alarms cleared.
Monitoring Instrumentation
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Safety Injection Tank Levels and Pressures
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Refueling Water Storage Tank Level
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CEA Positions
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RCS Temperature and Pressure
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Symmetric Offset
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All readings were within normal acceptable limits.
Safety System Operability
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The :aspector verified that the High and Low Pressure Safety Injection Systems were aligned for proper operation based on valve and switch position indication on the Main Control Board.
No items of noncompliance were identified.
6.
In-Office Review of Monthly Reports The inspector reviewed the Monthly Operating Reports for January and February 1980, to verify that reporting requirements were met. No inade-quacies were identified.
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7.
Exit Interview The inspector met with licensee representatives at the conclusion of the inspection (see detail 1 for attendees) to discuss the scope and findings of the inspection as detailed in this report.