IR 05000271/1978022
| ML19259A625 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 11/06/1978 |
| From: | Kalman G, Mccabe E, Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19259A620 | List: |
| References | |
| 50-271-78-22, NUDOCS 7901080258 | |
| Download: ML19259A625 (10) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No.
50-271/78-22 Docket No.
50-271 License No. DPR-28 Priority Category C
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Licensee:
Vermont Yankee Nuclear Power Corporation 20 Turnpike Road Westborough, Massachusetts 01581 Facility Name:
Vermont Yankee Nuclear Power Station Inspection at: Vernon, Vermont Inspection conducted: October 3-6, 1978 m
i in Inspectors:
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//[7bf T. Stetka, Reactor Inspector dat6 signed
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/P8 G. Kalman, Reactor Inspector date signed C.O. A h h date signed t
Approved by:
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.E. C. McCabe, Chief, Reactor Projects date signed Section No. 2, RO&NS Branch Inspection Summary; Inspection on October 3-6,1978 (Report No. 50-2 t/78-22)
Areas Inspected:
Routine, unannounced inspection by regional based inspectors of plant operations. refueling preparation, refueling activities, refueling outage related mair.tenance, and follow-up of outstanding items from previous inspections. The inspection involved 56 inspector-hours onsite by two NRC regional based inspectors.
Results: Of the five areas inspected, no items of noncompliaace were found in three areas; two apparent items of noncompliance were found in two areas (Infrretion - failure to barricade high radiation area; Infraction - failure to fallow rafueling procedure).
790108 52 5#
Region I Form 12 (Rev. April 77)
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D_ETAILS E
1.
Persons Contacted
- Mr. R. Burke, Engineering Support Supervisor
- Mr. W. Conway, Plant Superintendent Mr. A. Diaz, ISC Supervisor
- Mr. D. Girroir, Engineeri.g Assistant Mr. S. Jefferson, Reactor and Computer Supervisor Mr. B. Leach, Health Physicist
- Mr. W. Murphy, Assistant Plant Supervisor
- Mr. J. Pelletier, Technical Assistant Mr. W. Penniman, Security Supervisor Mr. R. Sejka, Operations Superviser Mr. G. Weyman, Chemistry and HP Supervisor The inspectors also interviewed several other licensee personnel during the course of this inspection. These employees included operations, engineering, maintenance, and health physics personnel.
- denotes those present at the exit interview.
2.
Licensee Action on Previous Inspection Findings (Closed) Unresolved item (271/78'-02-03): The licensee has revised procedure OP 4114, Standby Liquid Control System, as Revision 10 to include test firing of the squib valve charges using the installed electrical circuitry.
(Closed) Unresolved item (271/78-09-01): The licensee considers that the fire protection system is a non-nuclear safety system and therefore is not under the purview of 10 CFR Part 21.
(This matter is currently the subject of an on-going NRC review.)
Fire protection equipment will be purchased by the licensee to meet the requirements of the National Fire Protection Association (NFPA). The inspector had no further questions relative to this topic at this time.
(0 pen) Unresolved item (271/78-09-02): Procedure OP 4511 has been re-vised as Revision 8 to include calibration of the trip setpoint aiid a request for I&C Department action to reset the setpoint when required.
The procedure will be approved and issued by November 30, 197.
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(Closed) Unresolved item (271/77-22-01): Maintenance Requests 77-1123 and 77-1124 to mark all unterminated cables inside the Main Control Board were completed on November 24, 1977 and May 8,1978, respectively.
(Closed) Unresolved item (271/78-05-03): The licensee completed changeout of the flammable Contact Arm Retainers on September 28, 1978 and relay retesting is in progress. The flammable (Celcon) contact arm retainers will be disposed of and all new (non-flammable) re-tainers have been marked for permanent identification.
3.
Review of Plant Operations The inspectors reviewed selected facility operations logs, reviewed selected surveillance tests used to demonstrate component operability, and toured areas of the facility to determine that operations were ir accordance with the Technical Specifications (TS).
The facility is presently engaged in a refueling outage; therefore, the review of plant operations included refueling operations in progress and the determination that all necessary systems for this mode of operation are functional.
a.
The Shift Supervisor's Log, Control Room Operators' Logs, Re-fueling Coordinator's Logs, and Auxiliary Operators' Logs were reviewed.
This review, combined with panel observations in the control room and observation of various parameters throughout the facility during the plant tours verified that the following requirements were being adhered to:
Source Range instruments inserted and indicating a count
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rate greater than 3 cps; Senior Reactor Operators in the control room and on the
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refueling floor during core alterations; Refueling interlocks were functionally tested within one
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week of the start of core alterations and were subsequently tested on a weekly basis; Fuel movement step completions were documented on the fuel
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loading schedule and the core and spent fuel pool status boards on the refueling floor and in the control room were accurate and up-to-date;
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Required communications channels betwaen the refueling
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floor and control room were operable; Daily inspections of the refuel platform grapple and cable
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were conducted and logged in the Control Room Operators'
Log; Spent fuel pool level and temperature were within TS
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requirements; Fuel movement was not commenced until at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
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after reactor shutdown; Containment integrity was established and maintained; and,
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Standby Gas Treatment (SBGT) Systems were operable.
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No items of noncompliance were identified.
b.
The facility tour was begun on a back shift at 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br /> on October 3 and ended at 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br />.
The tour was resumed at 2330 hours0.027 days <br />0.647 hours <br />0.00385 weeks <br />8.86565e-4 months <br /> October 3 and concluded at 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> on October 4.
Further facility tours were also conducted during the day shifts.
The tours included the following areas:
All levels of the reactor building;
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Primary containment drywell;
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Refueling floor including observation of fuel movements,
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fuel channeling operations,and reactor vessel re-assenbly; and, Advanced offgas building.
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Due to the facility's refueling outage status, particular emphasis was placed on radiation area controls and plant housekeeping conditions. The tours identified the following:
(1) Plant housekeeping was observed to be effective and acceptable.
The licensee has taken actions to assure these conditions are maintained and will perfonn a general plant clean-up prior to resumption of operations.
(2) At approximately 2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br /> on October 3, the inspectors observed tnat the Control Rod Drive (CRD) repair room and CRD laydown areas, both labeled as high radiation areas, were not barricaded. A licensee representative was notified and the areas were immediately barricade.
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Subsequent to this finding, the inspector reviewed area survey sheets and toured the CRD repair room with licensee health physics personnel to verify that high radiation area posting was required for this area.
Failure to barricade a high radiation area is contrary to TS 6.5.B.1 and is considered to be an item of r.oncompliance
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(271/78-22-01).
Subsequent tours of this area and other radiation areas throughout the facility did not identify any additional problems. Since corrective action on this item was initiated prior to the completion of this inspection and that action appear ed to be effective, no response to this item is required.
4.
Refueling Preparation a.
Scope The inspector reviewed the receipt and inspection records associated with the 60 new fuel assemblies which were received for use during reload number 5.
The inspection reports of the 36 remaining re-load 5 fuel assemblies which had been stored on site were also reviewed. The refueling equipment interlock check procedure and report were examined. The findings of the pre-refueling reactor building crane inspection were reviewed and it was ascertained that related discrepancies were corrected in a timely manner.
The results of the fuel sipping operations were reviewed and the associated equipment was examined, b.
Documents Reviewed (1) A.P.1500, Revision 8, March 4,1977, Special Procedures During New Fuel Receip+ and Inspection.
(2) 0.P.1400, Revision 9, May 25,1978, Fuel Receipt and Preliminary Handling.
(3) 0.P.1401, Revision 7, June 30,1978, New Fuel Inspection and Channeling.
(4)
0.P. 5230, Revision 2, March 28,1978, Refueling Platform.
(5) VYOPF 5240.01, Revision 2, Reactor Cuilding Crane Inspection Checklist.
(6) VYOPF 1410.01, Revision 6, Fuel Loading Schedul l-
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Findings
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The inspection of the pre-refueling activities did r.at identify
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any discrepancies.
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5.
Refueling Activities
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Scope
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Fuel transfer activities were observed and related procedures were reviewed.
It was ascertained that Technical Specification
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requirements were included in the procedures and that refueling
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operations were conducted in accordance with applicable procedures.
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Follouing the final fuel move on October 4,'1978, the inspector reviewed the licensee's video tapes of the reactor core loading to verify proper core loading and fuel cell orientation.
In addition, the inspector reviewed the reretor contractor's (General
Electric) evaluation of the core loading pattern (the "as-loaded" core) and compared this evaluation to the actual fuel pattern.
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b.
Documents Reviewe'd
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A.P. 1000, Revision 4, September 15, 1978, Refueling.
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(2) 0.P.1100, Revision 7, March 3,1978, Refueling Platform Operation.
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(3) 0.P.1101, Revision 7, March 28,1978, Fuel Assembly Movement.
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(4)
0.P. 1102, Revision 7, December 22, 1977, Fuel Channel
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0.P. 1200, Revision 4, September 15, 1978, Preparation of the Reactor Vessel for Refueling.
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0.P. 1410, Revision 6, May 8, 1978, Fuel Loading.
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(7) 0.P.1411, Revision 3, April 25,1978, Core Verification.
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(8)
0.P. 4102, Revision 8, September 15, 1978, Refueling Outage Tests.
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Findings Of the areas inspected, the inspector identified one item of noncompliance.
During the transfer of a fuel bundle from the spent fuel pool to the reactor vessel, the inspector noted that the control room was requested to record the Source Range Monitor (SRM) readings after the fuel bundle was inserted into the core.
There did not appear to be any voice contact between the re-fueling floor and the control room immediately prior to the fuel bundle transfer.
Subsequent questioning of the refueling floor supervisor and other licensee personnel confirmed that the control room was not routinely alerted when core alterations were about to be made.
During the normal course of events, the operators in the control room were not aware when fuel was being moved in the reactor vessel and, as a result, did not continuously monitor the SRM's during these evolutions.
0.P.1101 require-the control room personnel to continuously monitor the SRM's whenever fuel is being moved in the reactor vessel.
Failure to follow procedures is contrary to Technical Specification 6.5.A and is considered an item of noncompliance (271/78-22-04).
E.
Maintenance a.
Scope The maintenance schedule for the refueling outage was examined and three maintenance procedures were chosen at random for review.
The work associated with the reviewed procedures was subsequently observed on a sample basis, b.
Documents Reviewed (1) 0.P.1201, Revision 4, October 2,1978, Assembly of the Reactor and Drywell Systems.
(2)
0.P. 5205, Revision 0, October 2,1978, Freeze Sealing.
(3) General Electric F0I 198/79900, Revision 0, August 3, 1977, HPCI Turbine.
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Findings No discrepancies were identifie.
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7.
Review of Terminal Boxes Located in the Containment Drywell On September 18, 1978, NRC Region I notified the licensee by telephone that during the recent outage at Peach Bottom Unit 2 (DN 50-277) it was determined that terminal block enclosures located in the contain-ment drywell were improperl mounted.
As a result of this telephone conversation, the licensee conducted an inspection of the terminal box installations in the Vermont Yankee drywell.
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The inspection identified that some terminal boxes had holes in the side and top of the box and that the conduits entering the boxes were not sealed.
The licensee instituted a program to resolve these problems.
This program checked and/or corrected the following items:
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The enclosures had cover gaskets; The ends of conduits enteriag the enclosures were sealed;
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The bottom of the box had a drainage hole;
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All other holes in the box were sealed; and,
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The terminal block mountings within.the enclosures were secure.
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Following subsequent review of these terminal boxes, the licensee deter-mined that sealing the ends of the conduit entering the boxes would not be feasible due to the number of cables within the conduit.
As an alternate plan, the licensee will verify that all terminal blocks and splices within the boxes are environmentally qualified.
Terminal blocks or splices that are not qualified will be replaced with qualified com-ponents. This action will be completed prior to plant startup.
The inspector toured the drywell and examined these terminal boxes including the corrective actions initiated by the licensee.
In addition, the inspector reviewed the licensee's corrective action program.
The inspector had no further questions on this item.
8.
Closeout of IE Bulletins and Circulars a.
Site documents related to IE Bulletins were reviewed to verify the licensee's responses were timely, accurate and adequate.
No problems were identified with the licensee's responses and planned or completed corrective action for the following IE Bulletins:
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IEB 78-03, Potential Explosive Gas Mixture Accumulations
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Associated with BWR Offgas System Operations; and, IEB 78-06, Defective Cutler-Hammer Type M Relays With
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DC Coils.
A detailed independent review of the licensee's action with respect to IEB 78-03 was conducted to verify all the stated actions were completed.
The following items were included in this review:
Inter-office memorandum detailing an evaluation of the
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hydrogen detonation that occurred on December 10, 1977 and the corrective actions taken; Revised procedures 0.P. 3123, SJAE Rupture Diaphragm Failure
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and 0.P. 4150, Off Gas Evacuation System Surveillance; and, Plant Alteration Review (PAR) Forms 78-9, 78-11, and 78-13
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that modified the advanced offgas system.
In addition the inspector held discussions with licensee repre--
sentatives regarding offgas system operations and toured the Advanced Offgas (A0G)' building to examine the installed system.
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It was determined that procedure AP 0042, Pl.nt Fire Protection Procedure, Revision 2, is being revised to include the entire A0G system including A0G piping from the condenser water box to
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the SJAE, SJAE room, and the stack rooms as Permanent Fire Control areas. This revision is not approved and is an unresolved item (271/78-22-02) pending approval and issuance of the revision.
b.
The inspector reviewed the licensee's action with respect to Circular IEC 77-13, Reactor Safety Signals Negated During Testing,.
to detennine if the licensee had received them and had taken appropriate action.
No problems were identified relative to the licensee's action on this circular.
9.
Follow-up On IE Bulletin 78-05 Concerning GE Model CR 105X Contactor Auxiliary Contacts On June 19, 1978, the Duane Arnold (DN 50-331) facility licensee reported that CR 105 auxiliary contacts used in NEMA Size 2 contactors were found to be inoperable. The problem reportedly resulted from loose plunger arm screws on the auxiliary contact.
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The licensee was apprised of this problem and as a result took action to examine their CR 105 auxiliary contacts.
This investigation identified a different auxiliary contact configuration than was reported by Duane Arnold. The licensee will verie 'ightness of the auxiliary
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contact plunger arms and add a locking co... pound to the retaining screw to assure that the screw does not loosen.
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The inspector had no further questions on this item.
10. Review of Reactor Protection System (RPS) power Supply Surveillance Program During a review of the E. I. Hatch Unit 2 nuclear power plant (DN 50-366)
reactor protection system (RPS) power supply, the NRC staff identified certain specific deficiencies in the design of the RPS motor generator voltage regulator. As a result of this finding, the NRC staff requested facilities with similar design RPS power supplies to perform a specific surveillance and calibration program on the voltage regulators to assure proper operation.
The Vermont Yankee surveillance and calibration program relative to these voltage regulators was reviewed to verify that the necessary actions were being accomplished.
The licensee has not completed the calibration of the over-voltage (0V) and under-frequency (UF) relays.
In addition, the licensee is installing a modification to provide under-voltage (UV) protection to the voltage regulators. These calibrations and the modifications will be completed prior to the resumption of plant operation. This item (271/78-22-03) is unresolved pending completion of these actions.
11.
Unresolved Items Items about which more information is required to determine accept-ability are considered unresolved.
Sections 2, 8, and 10 of this report contain unresolved items.
12.
Exit Interview An exit interview was conducted on October 6,1978 with licensee representatives (denoted in paragraph 1) at the conclusion of the inspection.
The inspectors summarizeri the purpose and scope of the inspection and discussed the findings.