IR 05000309/1979006
| ML19209B027 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 08/14/1979 |
| From: | Keimig R, Lazarus W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19209B024 | List: |
| References | |
| 50-309-79-06, 50-309-79-6, NUDOCS 7910090018 | |
| Download: ML19209B027 (12) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No. 79-06 Docket No. 50-309 Category:
C License No. DPR-36 Priority:
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Licensee: Maine Yankee At.orcic Power Company 20 Turnpike Road Westborough, Massachusetts 01581 Fat.?lity Name: Maine Yankee Atomic Power Station Inspectirn at: Wiscasset, Maine
~ Inspection conducted: My1
, 7-8, 24-26, and June 4-6, 1979 Inspectors:
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7f, yrJ. La arus date signed Approved by:
s/FG R. R. Kejdig, Chief, Reactor date signed Projects Section No. 1, RO&NS Branch Inspection Summary:
Inspection on May 1-3, 7-8, 24-26, and June 4-6, 19/9 (Report No.
50-309/79-06 Areas Inspected:
Routine unannounced inspection by a regional based inspector of licensee actions concerning selected IE Bulletins; followup on selected Licensee Event Reports (LER's); and followup on previous inspection findings. The inspection involved 80 inspector-hours on site by one NRC inspector.
Results: No items of noncompliance were identified.
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DETAILS 1.
Persons Contacted
- P. Anderson, Staff Assistant to the Plant Manager
- R. Arsenault, Assistant Operations Department Head
- M. David, Training Coordinator
- C. Frizzle, Assistant Plant Manager J. Hebert, Technical Assistant
- W. Paine, Operations Department Head G. Pillsbury, Engineering Assistant R. Prouty, Maintenance Department Head R. Radasch, I and C Department Head
- S. Sadosky, QC and Audit Coordinator D. Sturniolo, Ch3mistry and Health Physics Department Head B. Tuthill, Assistant Engineer E. Wood, Plant Manager The inspector also interviewed several licensed operators, auxiliary operators, and other members of technical and administrative staffs.
- Denotes those present at the exit interview.
2.
Followup on Previous Inspection Findings (Closed) Noncompliance (309/78-17-03): The inspector reviewed training records for several operators and verified that periodic quizzes were administered during the 1978-1979 training cycle, as required by the licensee's training program and 10 CFR 55.
(Closed) Unresolved item (309/78-17-04): The inspector reviewed I and C Department training records and verified that the records had been updated as of the most recent training conducted.
(Closed) Unresolved item (309/78-05-01): As detailed in this report, the inspector verified that valves necessary for the proper functioning of the Emergency Core Cooling Systems are included in a locked valve check list.
(Closed) Noncompliance (309/78-03-02): The inspector verified that procedure 1-13-1, "RHR Startup and Operation," had been revised to include a verification that Spray Building ventilation is realigned through the HEPA/ Charcoal filters prior to placing the P.HR system in service. This action was taken in accordance with Maine Yankee letter WMY 79-56 of May 23, 1979.
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Followup on IE Bulletin 79-06B The inspector conducted a review of the licensee's actions regard-ing IE Bulletin 79-06B, " Review of Operational Errors and System Misalignments Identified During the Three Mile Island Incident,"
to verify that the concerns identified in Bulletin had been adequately addressed and resolved as reported in licensee letters dated April 26, May 4 and 14, 1979.
A.
ONSITE REVIEW OF OPERATOR TRAINING The inspector verified the adequacy of licensee administered operator training by revie'. of training records and by discus-sion with licensed operators.. The following specific areas of training were evaluated, with findings as indicated.
That operators have received training on any procedure
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changes initiated as a result of Bulletin 79-06 and 79-06B.
That operators have been instructed on the specific
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measures which provide assurance that engineered safety features would be available if required, in particular, measures for returning such systems to operable status following maintenance and testing.
That operators have been instructed on the specific and
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detailed measures to assure that automatic actuations of emergency safety features are not overridden except as permitted in the bulletin.
That operators have reviewed plant automatic actions
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initiated by reset of engineered safety features, that could affect the control of radioactive liquids and gases.
That plant operators and supervisory personnel have been
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instructed in the provisions and directives for early NRC notification of serious events.
The inspector attended one training session held for plant operators and interviewed two licensed operators on each shift after the training had been completed.
No inadequacies in operator training in these areas were identified except for one area which concerned instruction on measures which would insure that engineered safety feature systems would be returned to service following maintenance or testing. The licensee was in the process of revising the Maintenance Request form 1119 273
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-4-to provide additional documentation of system operability following maintenance. This item is unresolved pending completion of training of operators in this area, which will be verif ad in a subsequent inspection (79-06-01).
B.
ONSITE INSPECTION OF ENGINEERED SAFETY FEATURES (ESF)
The inspector performed reviews in the following areas to verify that ESF are operable according to T.S. requirements and that licensce's procedures and administrative controls provide adequate assurance of continued operability.
(1) The valve / breaker / switch alignment procedures for all ESF systems were compared to current piping and instru-ment diagrams (P and ID's), as follows, to verify the adequacy of the alignment procedures.
P and ID's 11550-FM-92A, RHR, Containment Spray, and LP Safety Injection 115E0-FM-91A, 91B and 91C, Charging and Volume Control and HP Safety Injection 11550-FM73A, Steam Generator Feedwater Piping 11550-FM-88A, Diesel Generator Starting A.c 11550-FM-72A, Auxiliary Ste n Piping 11550-FM-70A, Main Steam Piping 11550-FM-78A,B, Secondary Component Cooling 11550-FM-94,A,B,C, Primary Component Cooling Operating Procedures 3.1.2, ECCS Routine Testing, Rev. 10 (Contains Safeguards Locked Valve List)
3.1.4, Emergency Diesel Generator Monthly Test, Rev. 6 6 3.1.5, Auxiliary Feed Pump Testing, Rev. 6 6 1-12-1, Containment Ventilation and Purging, Rev. 7.
The following valves were identified as not included in Safeguards and Auxiliary Feedwater Locked Valve Lists, but were important to system operability:
MS-185, Steam Supply to Turbine Auxiliary Feedwater Pump
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SCC-290, 297, 299, and 306, Secondary Component Cooling
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Supply and Return for Emergency Diesel Generators 1119 274
-5-Procedure changes were made by the licensee and reviewed by the inspector to add these valves to the locked valve lists in procedure 3.1.2.
No other inadequacies were identified.
(2) During preparation for starting up the plant on May 24-26, 1979, the inspector performad an independent verification of valve / breaker / switch alignments for ESF systems, including auxiliary feed system, in accordance with procedures 3.1.2 and 3.1.4.
The inspector accom-panied plant operations personnel as they performed the required alignments.
During this verification, several valves had to be positively identified by'use of P and ID's, because of missing labels.
The licersee agreed to compile a list of valves requiring labels and to have labels installed during the next refueling outage.
Completion of this action will be verified in a subse-quent inspection (79-06-02).
No items of noncompliance were identified.
(3) The inspector reviewed the following licensee procedures to verify that adequate administrative control exist to assure that ESF components are properly restored to service following maintenance and testing or extended outages:
0-07-3, Maintenance Requests 16.1, Maine Yankee Operation - Safeguard and Yellow Tag Control Log 3.1.2, ECCS Routine Testing 5-50, Replacement of LPSI Mechanical Seals 5-111, Routine Corrective Maintenance.
No inadequacies were identified.
(4) The following surveillance procedures were reviewed to verify that when completed, systems had been returned to an operable condition.
Completed test procedures were reviewed for dates indicated to verify that acceptance criteria had been met for the last test.
3.1.2, ECCS Routine Testing, Rev. 10 (5/26/79).
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3.1.14A and B, Emergency Diesel Generator Cold
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Plant Testing (7/1/78).
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-6-3.1.15.1, ECCS Operational Test - Refueling -
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Containment Isolation System (7/20/78).
3.1.15.2, ECCS Operational Test - Refueling -
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Recirculation Actuation System (7/20/78).
3.1.15.3, ECCS Operational Test - Refueling -Pump
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Flow Testing (7/19/78).
3.1.15.4, ECCS Operational Test - Refueling Safety
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Injection Tank Check Valve Testing (7/17/78).
3.1.17, Containment Spray Nozzle Test (5/4/77).
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3.1.4, Emergency Diesel Generator Routine Testing,
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1-A (5/3/79), 1-B (4/29/79).
3.1.5, Auxiliary Feed Pump Testing (3/21/79).
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3.1.12, Post Accident Purge System Testing (8/29/78).
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3-6.2.2.14, S.I.A.S. Initiation Channels (2/27/79).
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Procedure 3.1.5, " Auxiliary Feed Pump Testing," had last been performed with the plant shutdown so that insuf-ficient steam pressure was available to adequately test the turbine driven auxiliary feed pump.
The pump will be tested when sufficient steam pressure is available during the plant heatup.
This item is unresolved pending review of this data during the next routine inspection (79-06-03).
Individual surveillance procedures performed during plant outages do not realign systems for operation.
Pracer are 3.1.2 is performed prior to plant startup to assure that systems disturbed during the outage are properly returned to service.
No items of noncompliance were identified.
C.
Onsite Assessment of Operating Procedures The inspector reviewed plant procedures and interviewed plant operators to verify that the following requirements of Bulletin 79-06 had been met:
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Applicable operating procedures revised to specify conditions which must exist prior to stopping safety injection; 1119 276
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-7-Isolation of Contcinment (except for ESF systems)
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upon safety injection actuation (SIAS) to prevent inadventent transfer of radioactive liquids; Specify requirement to run reactor coolant pumps
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following SIAS as long as they are providing forced flow and continued operation will not degrade coolant system pressure boundary; Station an auxiliary feedwater operator who will
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have no concurrent duties except operation of auxiliary feedwater controls when that system is required; Specify indications available which could show that
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a power operated relief valve has stuck open and to isolate a stuck valve with the blocking valve; Provide additional information to operators to not
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rely solely upon pressurizer level indication in evaluating plant conditions; Prompt reporting procedures must require NRC notifi-
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cation within one hour of the time the reactor is not in a controlled or expected condition.
Upon such notification, maintain an open communication channel with the N2C; Operators have been trained in the above areas and
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are aware of changes in the operating procedures.
The following procedures were reviewed by the inspector to verify that the above items had been incorporated:
2-6, Steam Line Rupture, Rev. 4 2-25, Excessive Radiation Levels, Rev. 4 2-13, Major Loss of Reactor Coolant, Rev. 5 1-1, Plant Heatup, Rev. 9 2-10, Loss of System Overpressure, Rev. 4 2-12, Loss of Reactor Coolant, Rev. 4 2-11, Reactor Coolant System Leak, Rev. 4 2.50.1, Local Emergency Plan Emergency Procedure, Rev. 8 2.50.2, Site Emergency Plan Emergency Procedure, Rev. 10 2.50.3, General Emergency Plan Emergency Procedure, Rev. 10 2.50.3, General Emergency Plan Emergency Procedure, Rev. 10 0-09-2, Reportable Occurrence Reports, Rev. 6 1119 277
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_g.
9-E-2, OPMEMO:
Safeguard Valve Operation, dated 5/7/79.
Except as noted below for GPMEMO 9-E-2, no inadequacies were identified.
OPMEMO 9-E-2 lists Containment Isolation System (CIS) valves which operators are directed to close upon SIAS and other CIS valves whose control switches must be placed in the closed position prior to resetting SIAS or CIS instrumentation.
In a review of P and ID's 11550-FM-70A through 102A, the inspector identified two additional valves for inclusion in this list.
Both were verified to have been added to the OPMEMO by the inspector prior to the completion of the inspection.
D.
Tagging Practices on Control Panels The inspector reviewed the practice of tagging switches on the control panels with the operators to verify that precau-tior.s were exercised to preclude hanging equipment caution tags in a manner that could obscure valve or switch positions.
Plant procedures do not explicitly address this concern but operators indicated in their discussions with the inspector that they have always been cautious to position tags in such a manner that switch / valve positions are not obscured.
This was further reinforced by recent training received by the operators concerning the events at Three Mile Island Unit 2.
The inspector observed the control panels and Main Control Board several times during the inspection.
No tags were hung in a manner that could obscure valve / switch positions.
The inspector had no further questions in this area.
4.
IE Bulletin 79-10:
Requalification Training Program Statistics The inspector reviewed requalification exam summaries for 1975-1978 to verify that the statistics reported in the licer.see letter dated May 23, 1979 were accurate.
No inaccuracies were identified.
The inspector reviewed individual examinatiens to verify that exams were given annually as required by 16 CFR 55.
The interval between the 1978 and 1979 exams was acceptable; however,the inter-val between 1977 and 1978 exams varied between 12 and 18 months for individuals.
The licensee stated that this was due to complet-ing the training program and administering exams early in 1977 and then later in 1978, the objective in all cases was to complete the exams before June of each year.
The inspector informed the licensee that for future examinations the maximum interval will be 15 morths.
The licensee agreed to comply with this time limit.
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-9-The inspector noted that the 1979 examinations had been completed but not yet graded. The licensee agreed to complete the grading of these exams by August 5, 1979.
This will be reviewed in a subsequent inspection (79-06-04).
5.
IE Bulletin 79-07:
Seismic Stress Analysis of Safety Related Piping The issuance of this Bulletin was preceded by a Show Cause Order on March 13, 1979 which resulted in the plant shutting down until a seismic reanalysis was completed and reviewed by the NRC. The NRC staff acceptably resolved the concerr.s raised in the Show Cause Order and the Bulletin and a Termination of Order to Show Cause accompanied by a Safety Evaluation on May 24, 1979, allowing the facility to be restarted was issued.
The inspector had no further questions in this area.
6.
InOffice Review of Licensee Event Reports (LER's)
The inspector reviewed LER's received in RI office to verify that details of the event were clearly reported including the accuracy of the description of cause and adequacy of corrective action.
The inspector also detennined whether further information was required from the licensee, whether generic implications were indicated, and whether the event werra.nted onsite followup.
The following LER's were reviewed:
79-006, ECCS Valves HSI-M-54 and CS-M-66 Malfunctioned During
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Test 79-007, Broken Body-to Bonnet Stud on Pressurizer Spray Valve
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79-008, Unplanned Release of Gaseous Radioactivity
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79-009, Quarterly Release Rate for Iodine-131, and Particulates in Excess of Objectives
- -- 79-010, Reactor Coolant System Dose Equivalent Iodine-131 in Excess of 1.0pCi/ gram.79-011, Failed Power Supply for RPS Channel B Low Steam
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Generator Pressure Trip
- -- 79-012, Non-conservative Trip Setpoints for RPS Low Steam Generator Pressure.
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Except as noted below for those LER's identified by (*), the inspector had no further questions in this area.
7.
Onsite Licensee Event Followup For those LER's selected for onsite 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 Specificatio. limits.
The review included discussions with licensee perscnnel, review of PORC meeting minutes, Plant Information Reports (in-house reports), and applicable logs. The following LER's were reviewed on site.79-008, Unplanned Release of Gaseous Radioactivity, and
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79-009, Quarterly Release Rate for Iodine-131 and Particulates
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in Excess of Objectives.
Both of these LER's were required because of a release on March Ic,1979 which was a result of a body-bonnet gasket failure on
.;ontainment Spray Valve CS-M-2, Previous inspection of this event is documented in NRC Inspection Report 79-03 of May 2, 1979 and Paragraph 2 of this report.
Further investigation during this inspection concerned results of environmental monitoring following the release.
Technical Specification leble 4.8-1 requires that milk sampling be done weekly whenever plant Iodine-131 releases exceed the quarterly limit. Technical Speci#ication 3.17 is stated in teras of annual objectives for releases, with no clear means for determining a quarterly limit for Iodine-131. The licentee increased milk sampling to a weekly frequency commencing on April 2, 1979, based on the release of March 15, 1979. The inspector reviewed the data from weekly analyses of milk for dates April 2,10,17 and 23,1979.
All results indicated Iodine-131 was within the lower limits for detectability.
The licensee has agreed that the interpretation of the quarterly limit for Iodine-131, referred to in T.S. Table 4.8-1, would be the same as T.S. 3.17.A.1 (Gaseous release rate for the quarter, which would result in twice the annual objective if continued for a year).
Proposed Radiological Effluent Technical Specifications based on 10 JR S0, Appendix I have been submitted by the licensee for review and approval by the NRC Office of Nuclear Reactor Regulation.
Implementation of these specifications will clarify release limits and reporting requirements.
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The inspector had no further questions in this area.79-010, Reactor Coolant System Dose Equivalent Iodine-131 Greater
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than 1.0uC1/ gram.
Inspection of RCS Chemistry, which led to this LER, is detailed in Inspection Report 79-03. A supplemental report was submitted by the licensee, at the inspector's request to correct the statement that
"...no releases occurred during the time the coolant was above the 1.0pCf/ gram limit." A release had occurred on March 15, 1979 and was reported in LER 79-008, discussed above. The inspector had no further questions in this area.79-012, Non-conservative Trip Setpoints for RPS Low Steam Generator
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Pressure The situation in which all four low steam generator pressure reactor trip bistable setpoints were set 70 psig below the Technical Specification requirement of 485 psig, represents a licensee identified item of noncompliance. The inspector reviewed completed procedure 3-6.2.1.2,
" Protective and Safeguard Channel Calibration - Steam Generator Pressure" which had been performed August 14, 1978 and concurred with the licensee's evaluation of the reason for the error, which was a combination of personnel error and procedural inadequacy which led to transcribing the pressure transmitter voltage output for the incorrect pressure (415 vs. 485) from the QA and Calibration Form for each pressure transmitter (PT) to a data table in procedure 3-6.2.1.2.
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This data sheet was used for the subsequent monthly surveillance tests for the RPS until the error was discovered by the I and C Department.
The licensee is in the process of revising the procedure to more clearly specify the PT pressure / output voltage to be recorded in the data table.
Completion of this action will be reviewed in a subsequent inspection (79-06-05).
The inspector verified that the steam generator low pressure trip bistables had been correctly reset following discovery of the error.
Unresolved Items Unresolved items are items for which more information is needed to determine if the item is acceptable, a deviation or a noncompliance.
Detail 3 of this report contains unresolved items, lil9 2Ol
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9.
Exit Interview At the conclusion of this inspection, the inspector held a manage-ment meeting (see Detail 1 for attendees) to discuss the scope and findings of the inspection as detailed in this report.
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