IR 05000309/1981020
| ML20040B285 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 10/08/1981 |
| From: | Gallo R, Lazarus W, Swetland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20040B282 | List: |
| References | |
| 50-309-81-20, NUDOCS 8201250350 | |
| Download: ML20040B285 (4) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No.
50-309/81-20 Docket No.
50-309 License No.
DPR-36 Priority
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Category C
Licensee:
Maine Yankee Atomic Power Company 1671 Worcester Road Framingham, Massachusetts 01071 Facility Name: Wiscasset, Maine Inspection conducted:
July 19-30,1981
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Gallo, Chief, Reactor Projects Section IA, Division of Resident and Project Inspection Inspection Summary:
Inspection conducted July 19-30,1981 (Report No. 50-309/81-20)
Areas Inspected:
Routine, regular and backshift inspection by the resident inspectors (65 hours7.523148e-4 days <br />0.0181 hours <br />1.074735e-4 weeks <br />2.47325e-5 months <br />).
The inspection included onsite followup of licensee events, and followup of events occurrirg during the inspection.
Results: Of the two areas inspected, no items of noncompliance were identified in one area; one item of noncompliance was identified in the other area (continued operation of the facility with equipment not meeting the least conservative aspect of a Limiting Condition for Operation).
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khhhCK0500o3o9 350 811113 Region I Form *12 G
(Rev. April 77)
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DETAILS 1.
Persons Contacted
- R. Arsenault, Operations Department Head J. Brinkler, Technical Support Department Head / Assistant Plant Manager T. Davin, Plant Shift Superintendent
- W. Paine, Technical Assistant to the Plant Manager
- D. Stevenson, Plant Shift Superintendent
- E. Wood, Plant Manager The inspectors also interviewed several plant operators, technicians and members of the engineering and administrative staffs.
- denotes those present at exit interview.
2.
Onsite Followup of Licensee Events The inspector reviewed applicable logs and records and discussed the events with licensee personnel to verify that appropriate corrective actions had been taken, that the events had been reported in accordance witn 10 CFR 50.72 and that continued operation of the facility was conducted within Technical Specification limits.
On July 18, 1981, the licensee reported via the Emergency Notification System (ENS), the failure of a 2" pipe weld in the ietdown system while operating the reactor at 2% power.
The pipe failure was isolated immediately; 100 gallons of reactor coolant were spilled in the lower level of the auxiliary building.
The inspector reviewed the operating logs and records to verify that appropriate radiological safety precautions were implemented to protect plant personnel from unnecessary exposare and that radioactivity measurements were taken to evaluate the onsite and offsite consequences.
No offsite release was identified.
The weld failure nas caused by fatigue stress resulting from oscillation of the letcown system back pressure regulating valve.
The inspector reviewed the maintenance records f or the letdown system repair (MR 81-1436)
and the welder's qualification record (Procedure YA-WP-4) dated June 22, 1981.
Testing of the letdown flow and pressure regulating systems was completed and permanent corrective measures are being evaluated by the licensee.
The letdown system will be controlled manually until corrective action is completed.
Pending engineering review of the system piping and controls, the licensee's corrective measures are considered an unresolved item.
(309/81-20-01)
3.
Followup on Events Cccurring During the Inspection The inspector reviewed the Control Room Log at 6:45 p.m. on July 19, 1981.
The reactor was critical at 2% power with coolant temperature about 540F.
The normal letdown path was isolated for repair of the weld k
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failure described in paragraph 2 above. An alternate letdown path had
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been established through the reactor coolant system drain piping to control reactor coolant volume until the normal letdown path could be restored. A log entry for 1:30 p.m., July 19, 1981 recorded the failure
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of automatic containment isolation valve DR-A-6 in the open pcsition and the repositioning of manual containment isolation valve DR-7 to the open
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position.
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DR-7 was opened to control flowrate in the alternate letdown path.
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Valves DR-A-6 and DR-7 are the containment isolation valves inside the containment for the alternate letdown path. Technical Specification (TS) 3.11.B.1 requires automatic containment isolation valves to be operable or locked closed and manual containment isolation valves to be closed
j whenever reactor coolant temperature is above 210F.
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The inspector questioned the control room operators (CRO) with respect to
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compliance with TS 3.11.B.l.
The CR0's were aware of the condition of i
the subject valves because the shift supervisor (PSS) had instructed them to close the upstream motor-operated valve (DR-M-3) in the event of an accident requiring containment isolation.
The CRO's had not identified this situation to be contrary to Technical Specifications, however, i
discussions with the on-duty PSS revealed the following information:
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the PSS knew the situation did not meet TS requirements;
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the PSS had received guidance, through his shift turnover, that the Operations Department Head had directed the operators to continue
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operating the plant at 2% power;
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the PSS acknowledged that he knew the situation technically called j
for a plant shutdown and cooldown; and,
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the PSS took no action to initiate a shutdown until the resident inspector brought the situation to the attention of the Assistant i
Plant Manager at 7:00 p.m., July 19, 1981.
The licensee commenced a plant shutdown at 7:15 p.m., July 19, 1981 at j
the direction of the Assistant Plant Manager.
l The inspector determined through subsequent discussions with the Operations
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Department Head, the PSS on duty when DR-A-6 failed, and the Shift Technical l
Advisor that:
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at the time the decision was made to open DR-7, the PSS and Operations l
Department Head discussed between themselves the condition of valves DR-A-6 and DR-7 and agreed that this condition did not meet Technical l
Spec'.fication 3.11.
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the Operations Department Head directed the PSS 'to continue operations and to instruct the CR0's to close DR-M-3 snould containment isolation i
be required because in the view of the Operations Department Head, this compensatory action met the intent of Technical Specification
3.11.
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the PSS followed the directions from the Operations Department Head j
and passed on this guidance to his relief at shift turnover.
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the Operations Department Head decided not to shutdown based on his i
judgement that the compensatory measure he prescribed sctisfied the Technical Specification intent of maintaining containment integrity.
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the Shift Technical Advisor knew that valve DR-A-6 had failed but
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was not informed of or consulted concerning the violation of Technical
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Specification 3.11 or the acceptability of the compensatory measure.
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The compensatory measure prescribed (manual remote operation of valve DR-M-3) did not provide containment isolation equivalent to the original plant design since valve DR-M-3 had not been leak-tested to containment
boundary specifications, operator action would be required immediately in
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the event of the postulated reactor coolant system break and no formal procedural requirement had been initiated to check that these verbal instructions would be carried out during an accident.
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A Region I management representative called the PSS the afternoon of July i
19, 1981 to obtain further information regarding the weld failure reported i
by the licensee the previous day.
The PSS did not inform the Region I I
representative of the off-normal operation of the alternate letdown l
system.
Failure to commence a plant shutdown af ter identification of a condition which is less conservative than the least conservative aspect of a Limiting
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' Condition for Operation constitutes a violation of the Technical Specifica-
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i tions and 10 CFR 50.36.
(309/81-20-02)
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Exit Interview
l On July 20, 1981, the in:pector met with licensee representatives (denoted j
in Details, paragraph 1) to discuss the scope and findings of the inspection.
The inspector expressed a concern regarding this violation of Technical
Specifications and stated that the existence of a misconception of operators to believe that they have the authority to institute compensatory measures rather than meet the literal requirement of any Technical Specification Limiting Condition for Operation without prior approval is unacceptable j
to the NRC.
At another meeting between the inspector and the plant manager on July 30, 1981, the licensee stated that no investigation into the events of July 19 had been conducted and that no corrective action had been taken.
The inspector expressed concern that the lack of prompt management attention
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i to take corrective action was not consistent with the severity of the violation. On August 13, 1981, the Vice President - Manager of Operations
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held a meeting with the plant operating staff to discuss operator responsi-i bilities relative to compliance with the Technical Specifications.
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