IR 05000271/1979008
| ML19249F062 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 08/03/1979 |
| From: | Foley T, Mccabe E, Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19249F059 | List: |
| References | |
| 50-271-79-08, 50-271-79-8, NUDOCS 7910030895 | |
| Download: ML19249F062 (19) | |
Text
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No. 79-08 Docket No. 50-271 License No. DPR-28 Priority:
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Catt. gory:
C 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:
May 1 17 and 21-25, 1979 Inspectors:
F!T 79 T. F. Stetka, Rea tor Inspector date ' signed
0 W
Y
?i T. Foley,ReactorInspectpf datd signed (WaSA6 ebbe 5/21-25/79)
W. A. Rekito, Reactor Inspector dat'e signed (5/23-25/79)
Approved by
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.
[
J.' c McCa6eMr., Chief, d6te/ signed Peactor Projects Section No. 2 RO&NS Branch Inspection '2ummary:
Inspectior, on May 15-17 and 21-25, 1979 (Report No. 50-271/79-08)
Areas Inspected:
Special, announced inspection of the facility strike plan impiementation, the licensee's actions taken in response to IE Bulletir. 79-08, License Event Report (LER) 79-11, and the qualifications of new 'nanagement personnel.
Plant tours were conducted.
The inspection involved 139.5 inspection hours on site by three NRC regional based inspectors.
Results:
No items of noncompliance were identified.
1074 195 7910000 p
.
DETAILS 1.
Persons Contacted a.
Licensee Personnel
- Mr. R. Burke, Engineering Support Supervisor
- Mr. W. Conway, Plant Superintendent
- Mr. P. Donnelly, I&C Supervisor Mr. H. Eichenholtz, Technical Assistant
- Mr. S. Jefferson, Reactor and Computer Supervisor Mr. B. Leach, Health Physicist
- Mr. W. Murphy, Assistant Plant Superintendent
- Mr. J. Pelletier, Maintenance Supervisor Mr. W. Penniman, Security Supervisor
- Mr. J. Sinclair, Records Clerk
"Mr. R. Sojka, Operations SuperviNr
- Mr. D. Reid, Lead Technical Assistant
- Mr. G. Weyman, Chemistry and HP Supervisor b.
State of Vennont Personnel Mr. P. Paull, State Nuclear Engineer (Vermont Public Service Board)
Mr. D. Scott, Health Physicist (Vermont State Health Department)
The inspectors also interviewed several other licensee person-nel during the course of this inspection.
These employees included operations, engineering, maintenance, and health physics personnel.
- Present at the exit interview.
2.
Review of Facility Strike Plan Imolementation a.
The inspector arrived on site at approximtely 11:00 PM (2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br />) on May 15.
Upon arrival, the inspector observed security measures in effect.
The security force is composed of non-union personnel and the site has maintained the full security force.
After arrival in the control rocm the inspector observed an operation watch shift turnover and reviewed the site staffing to assure that regulatory requirements were met.
Operations logs were reviewed and a tour was conducted of the control room, reactor building, turbine building, diesel generator rooms, and rad-waste area to verify the following:
10/4 196
.
Plant operation was consistent with regulatory require-
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ments; Plant housekeeping was being maintained;
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There were no fluid leaks or abnormal piping vibrations;
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and Radiation controls were being maintained.
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No inadequacies were identified.
b.
On May IS, accompanied by the Vermont State Nuclear Engineer, the inspector reviewed the licensee's implemented olans to cope with the strike by verifyky the following:
Plant staffing and on duty hours during the strike are
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capable of meeting regulatory requirements; Refresher training of licens&d personnel who are engaged
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in conduct of licensed activities, and non-licensed personnel who perform functions which they are not normally assigned, has been conducted; Arrangements have been made to ensure adequate goods at
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the site in addition to any necessary off-site shipment of radioactive matrials; Arrangements will remain in e"fect f6e medical treatment
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of injured or contaminated persons; Provisions have been made with local law enforcement
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agencies to deal with nondocile strikers; Emergency communication equipment is available and operable;
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and, On-site and off-site personnel are sufficient to imple-
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ment the emergency plan.
The licensee is developing plans for retaining of striking personnel.
The depth of this retraining will coincide with the length of time that personnel are on strike.
These plans will be reviewed on subsequent inspections.
This item is unresolved (271/79-08-01).
c During the conduct of this inspection the inspector continually observed various plant operations in progress and discussed 10/4 197
job functions with various licensee personnel to verify the effectiveness of these personnel in performing their assigned jobs. Additional shift turnovers were observed and the licensee's watch schedule was reviewed to assure proper staffing.
No inadequacies were identified.
3.
Review of Operator Training A review of training records and discussions with licensed opera-tors on each shift were conducted to verify the adequacy of licen-see administered operator training.
The review and discussions verified the following:
That operators are aware of the specific details of the TMI-2
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incident to the extent available and have received training on any procedure changes initiated as a result of Bulletin 79-08; That operators have been instructed or the specific measures
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which provide assurance that engineered safety features would be available if required, in particular, measures for return-ing such systems to operable status following maintenance and testing; That operators have been instructed on the specific and detailed
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measures to assure that automatic actuations of emergency safety features are not overridden; That operators have reviewed plant automatic actions initiated
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by reset of engineered safety features, that could effect the control of radioactive liquids and gases; and,
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That plant operators and supervisory personnel have been instructed in the provisions and directives for early NRC notification of serious events.
All operators interviewed appeared knowledgable of the events described in the Bulletin and of plant changes made as a result.
The licensee revised AP 0150, Responsibilities and Authorities of Operations Department Personnel, as Revision 10 on May 22, 1979, to provide guidance to the operators for establishing and maintaining a continuous communication channel with the NRC.
This AP and training sign-off sheets, demonstrating that operators had received training, were included in the inspector's review.
Glik h00
No inadequacies were identified.
4.
Review of Engineered Safety Features (ESF)
A detailed review of the ESF was conducted to verify by independent examination of records, procedures and equipment that ESF are operable according to T.S. requirements and that the licensee's procedures and administrative controls provide adequate assurance of continued operability.
a.
Valve /brecker/ switch lineups were reviewed for the following systems using the system procedures noted below.
These system procedures were compared to current system diagrams to verify adequacy of the lineups.
(1) Emergency Core Cooling Systems (ECCS)
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OP2120, High Pressure Coolant Injection System, Rev.
9 (HPCI);
OP2121, Reactor Core Isolation Cooling System, Rev.
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9 (RCIC);
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OP-2123, Core Spray System, Rev. 8 (CS);
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OP-2124, Residual Heat Removal System, Rev. 11 (RHR); and,
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OP2185, Condensate and Demineralized Water Transfer System, Rev. 5.
The actual valve positions for accessible valves that cuuld affect operation of the emergency systems were verified by observation on May 23, 1979.
(2)
Service Water System (SWS)
OP2181, Service Water / Alternate Coolir.g Ooerating Proce-dure, Rev. 8.
This procedure's valve lineup was reviewed to check valves supplying cooling water to the Emergency Diesel Generators and the Uninterruptable Power Supplies.
The actual valve positions for accessible valves that could affect system operation were verified by observation on May 22 and 23, 1979.
} i) / 4
(3) Standby Liouid Control System (SBLCS)
OP2114, Standby Liquid Control System, Rev. 7.
ihe actual valve positions for accessible valves that could affect system operation were verified by observa-tion on May 23, 1979.
(4) Standby Gas Treatment System (SGTS)
OP2117, Standby Gas Treatment, Rev. 5.
The actual valve and damper positions for accessible components that could affect system operation were verified by observation on May 23, 1979.
(5) Emergency Diesel Generators (EDG)
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OP2126, Diesel Generators, Rev. 6; and, RP2195, Fuel Oil Transfer System, Rev. 7.
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The fuel oil transfer system was reviewed to check valves supplying fuel oil to the EDG's.
The actual valve positions for accessible valves that could affect system operation were verified by observation on May 22, 1979.
(6) Emergency Electrical Systems
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OP2142, 4 KV Electrical System, Rev. 5;
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OP2143, 480 VAC System, Rev. 7; and, OP2145, Normal & Emergency 125 VDC Operation, Rev.
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4.
The actual breaker / switch positions for all applicable safety system circuitry were verified by observation on May 23, 1979.
b.
Review of the valve / breaker lineups identified in item A revealed a number of inadequacies.
Examples of some of these inadequacies follow.
) G { l) A0
(1) Emergency Core Cooling Systems i.
OP 2120 HPCI 808, the valve lineup requires the valve
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to be closed and capped.
The valve has a gage installed and is open to allow gage usage; HPCI 842, incorrectly listed on pages 4 and 5
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of the valve lineup as SGT-14;
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HPCI 39, on valve lineup but not shown on system drawing; and,
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Miscellaneous errors giving valves incorrect numbers, names, or locations.
ii.
OP 2123 CS 828 A&B, CS 829 A&B, the valve lineup requires
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these valves to be open.
The actual status of these valves is closed.
Since these valves are test connection isolation valves, their correct position is closed.
It was noted that the licensee's control room copy of the valve lineup showed this status change.
The diagram for the CS system is incorrect with regard to this piping; CS 819, the valve lineup requires the valve to
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be open.
Since the valve is a drain isolation, it should be and was identified as being closed.
This status was noted on the licensee's control room copy of the valve lineup; and, Miscellaneous problems such as valves with
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broken o, no identification tags and incor-rectly labeled instrument racks.
iii.
OP 2124 RHR 168, labeled as "RHR 'A&C' Pumps Minimum
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Flow", should be "RHR 'B&D' Pumps Minimum Flow"; and,
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RHR 199 A&B, the valve lineup requires the valves to be closed.
The valves are presently 10/4 201
connected via tubing to the sample sink and are used for sampling and are therefore open.
This status was noted on the licensee's control room copy of the valve lineup.
(2)
Standby Liquid Control System 0P 2114 SLC 18 does not appear on the valve lineup.
This
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valve is written in on the control room valve lineup and was correctly positioned.
(3)
Standby Gas Treatment System OP 2117 HVAC 14 & 15, these valve designations do not exist.
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They are apparently also designated as HPCI 842 and RCIC 817 respectively.
This problem was identified on the control room copy of the valve lineup.
(4)
Emergency Diesel Generators i.
RP 2195
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F0 29 A&B, incorrectly described in valve lineup as transfer pump suction valves in lieu of pump discharge valves;
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FO 4, the valve lineup requires the valve to be closed.
For correct system operation the valve
'should be and was identified to be open.
This status was noted on the licensee's control room copy of the valve lineup; F0 41 A&B, these valves are listed as duplex
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strainer outlets and signad as open on the control room valve lineups.
These valves could not be located for a valve position check and may not exist; and, F0 30 A&B, these valves, located between the
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fuel oil storage tank and EDG day tanks, were not listed on the valve lineup, however, they were verified to be in the correct position.
1074 202
11.
OP 2126 Root valves for flow instruments 28 A&B are not
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included on the valve lineup.
(5) Emergency Electrical System OP 2143 MCC 89A and 898 breakers, located on MCC 9B ana 8B
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respectively were not locked as designated on the breaker lineup.
This status was identified on the control room breaker lineup;
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V10-20 breaker on MCC 90 was closed instead of open as designed on the breaker lineup.
This status was identified on the control room breaker lineup; and,
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MCC-90 power panel breakers are not shown on the system drawings.
The inadequacies discussed above were verified, by observation (see Paragraph 4.a), to assure that proper system operations were not affected.
In most cases, the inadequacies were already previously identifie.1 by the licensee by pen-and-ink changes and notations on the control room valve / breaker lineup master copies in the control room.
The licensee will revise the valve / breaker / switch lineups to be consistent with actual system status.
These revisions will be completed by July 25, 1979.
This item is unresolved (271/79-08-02).
The licensee will check valves and breakers to assure that all items are properly tagged.
This tagging will be completed by July 25, 1979 and is considered unresolved (271/79-08-03).
The status of the F0-41 A&B valves (item (4)i) will be reviewed and appropriate corrective action taken.
This will be com-pleted by July 25, 1979 and is considered unresolved (271/79-08-04).
The valve lineups reviewed do not include instrument isolation valves and many of these valves are not tagged or identified.
The licensee has already identified what instrument valves require tagging and has developed a preliminary instrument valve lineup to be used when returning the plant to service following an outage.
The tagging of valves, issuance of the instrument valve lineup anc verification that all accessible 1074 203
valves are correctly positioned will be completed by July 25, 1979.
This item is unresolved (271/79-08-05).
c.
Review of system drawings to verify the adequacy of the valve /
breaker lineups identified the following inadequacies;
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RCIC valves 801,815, and 816 are not on the system drawing; SGTS valves 14 and 15 do not appear on the drawing (these
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valves may be incorrectly identified as discussed in Paragraph 4.b(3));
CS drawing 191168 shows valve V-25B in two places and
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shows an incorrect flow instrument 45A&B; Most system drawings do not shcw the instrument valves
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(though the root valves are shown);
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The drawings for the EDG lube oil, air start, and cooling are not complete.
The only drawings available are those by the diesel manufacturer which are not complete with respect to valve numbers, functions, or existence; and, Drawing inadequacies as identified in Paragraph 4.b.
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The licensee will review and revise drawings to be consistent with system configuration by July 25, 1979.
This item is unresolved (271/79-08-06).
5.
Surveillance Test / Maintenance Procedure Review a.
The inspectors reviewed the following Surveillance Tests /
Maintenance Procedures to assure that a system / component is returned to an operational lineup.
Proc. No.
Title Rev.
Date OP 4114 Standby Limid Control System
7/76/78
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OP 4115 Primary Containment System
5/8/78
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OP 4116 Secondary Containment System
5/8/78
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OP 4117 Standby Gas Treatment System
5/8/78 OP 4120 High Pressure Coolant Injection
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System
7/26/78 OP 4121 Reactor Core Isolation Cooling
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System
7/26/78 OP 4122 Auto Blowdown System
7/26/78
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10/4 204
Proc. No.
Title Rev.
Date OP 4123 Core Spray System
9/28/77
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CP 4124 Residual Heat Removal System
7/26/78 OP 4126 Diesel Generators Surveillance 10 4/12/79
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OP 4144 120/240 VAC Uninteruptable (Vital) MG Set Surv.
5/8/78 OP 4181 Service Wtr/ Alt. Cooling System
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Sury.
9/28/77 OP 4195 Fuel Oil Transfer System
9/11/78
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OP 4210 Testing the 125V Main Station
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Batteries, 125V Switchyard Batteries & 24V Reactor Protection System Batteries &
the UPS Batt.
7/13/78 OP 4211 Station Battery Discharge Test
5/8/78
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OP 4212 Transfer Trip, Primary &
Secondary Carrier Inservice Testing
5/8/78
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OP 4214 Core Spray & LPCI Aux Power Monitor Calibration
8/17/77 OP 4302 APRM Functional
12/6/77
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OP 4303 Rod Block Monitor Functional Test
5/25/78
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OP 4304 Rod Block Monitor Calibration
11/22/77 OP 4306 Control Rod Block System "A"
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Logic Test
5/25/78
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OP 4307 Control Rod Block System "B" Logic Test
5/25/78
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OP 4308 Average Power Range Monitor Calibration
5/25/78 OP 4310 Scram Disch Vol High Water
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Level Func/ Calibration
11/22/77 OP 4311 Drywell Hi Press Scram Cont
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Isol Func/ Calibration
5/8/78 OP 4312 Reactor Vessel Hi Press Scram
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Func/ Calibration
7/13/78 OP 4313 Reactor Water lo Level Scram -
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High/Lo Lo Water Isolation Functional / Calibration
10/12/78
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OP 4314 Generator Load Reject - Turbine Control Valve Fast Closure Scram Func/ Calibration
12/6/77
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OP 4315 Main Steam Line Radition Monitor Scram - Isolation Fun'tional/ Calibration
3/28/78 1074 20b
Proc. No.
Title Rev.
Date OP 4316 Reactor Manual Scram Functional
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Test
3/28/78
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OP 4317 Scram Test Switch Functional Test
5/8/78 OP 4318 Reactor Mode Switch in
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Shutdown Func Test
3/28/78
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OP 4319 RPS - First Stage Turb Press Func/Calib
5/25/78 OP 4320 RPS Response Time Check
5/8/78
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OP 4322 Main Steam Line Area High Temp
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Func/Calib
3/28/78 OP 4323 Main Steam Line High Flow
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Func/Calib
9/28/77
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OP 4324 Main Steam Line Low Pressure Func/Calib
9/28/77
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OP 4325 Main Condenser Low Vacuum Isol Func/Calib
5/25/78
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OP 4326 Reactor Bldg Vent & Refueling Floor Radiation Monitor Func/Calib
3/28/78
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OP 4332 Reactor Bldg Vent & SGTS Subsys "A" Logic Test
3/28/78 OP 4333 Reactor Bldg Vent & SGTS
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Subsys "B" Logic Test
3/28/78 OP 4334 Automatic Initiation Test of
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PCIS Valves
3/28/78
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OP 4335 Reactor Bldg Vent & SGTS Logic Power Monitor Functional Test
5/25/78 OP 4337 Reactor Water Level ECCS
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Initiation - Isol Func/Calib
3/7/79 OP 4338 Drywell High Pressure ECCS
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Func/Calib
5/8/78
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OP 4339 Reac Law Press ECCS Pump Start Func/Calib
5/25/78 OP 4340 Reac Low Press ECCS Valve
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Purm Func/Calib
3/7/79
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OP 4341 High Drywell Press - Cont Spray Interlock Func/Calib
5/8/78
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OP 4343 ADS System "A" Logic Test
7/26/78 OP 4344 ADS System "B" Logic Test
8/17/77
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OP 4345 ADS Power Monitor Functional
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Test
3/28/78 OP 4346 Core Spray Pump Disch Press
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Func/Calib
5/25/78
)0lh N
Proc. No.
Title Rev.
Date OP 4347 Core Spray Header Diff Press
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Func/Calib
9/28/77
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OP 4348 Core Spray Power Monitor Functional
5/25/78
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OP 4349 Core Spray Subsystem "A" Logic Test
9/28/77 OP 4350 Core Spray Subsystem "B"
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Logic Test
9/28/77
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OP 4351 RHR Low Pressure - RHR Inter-lock Value 10-15A&B Func/
Calib
12/6/77
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OP 4352 RHR Pump Discharge Pressure Func/Calib
5/25/78
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OP 4353 RHR System Power Monitor Functional
3/28/78 OP 4354 RHR Subsystem "A" Logic Test
4/12/79
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OP 4355 RHR Subsystem "B" Logic Test
4/12/79
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OP 4356 HPCI Steam Line High Flow Func/Calib
9/28/77
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OP 4357 HPCI Steam Low Press Func/Calib
5/8/78
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OP 4358 HPCI Steam Line & Space High Temp Func/Calib
8/17/77
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OP 4359 HPCI System Power Monitor Functional
5/25/78
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OP 4360 HPCI System Logic Test
5/8/78
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OP 4361 HPCI System Isolation "A" Logic Test
11/22/77 OP 4362 HPCI System Isolation "B"
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Logic Test
5/25/78
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OP 4363 HPCI/ CST Water Level Func/
Calib
5/8/78
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OP 4364 RCIC Steam Line High Flow Func/Calib
9/28/77
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OP 4365 RCIC Steam Line Low Pressure Func/Calib
11/22/77 OP 4366 RCIC Steam Line Tunnel & Space
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High Temp Func/Calib
//13/78
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OP 4367 RCIC System Power Monitor Func Test
5/25/78
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OP 4368 RCIC System Isolation "A" Logic Test
5/25/78
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OP 4369 RCIC System Isolation "B" Logic Test
11/22/77
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OP 4371 Drywell & Torus Press Trans-mitter Calib
7/26/78 10/4 20/
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Proc. No.
Title Rev.
Date OP 4372 Drywell & Torus Atmospheric
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Temp Calib
12/6/77 OP 4373 Torus Water Temperature
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Calibration
2/9/78 OP 4374 HPCI-Torus Water Level Func
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Test /Calib
5/8/78 OP 4375 Reactor Pressure Calibration
9/28/77
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OP 4376 Torus-Reactor Bldg Vacuum Breaker Diff Press Calib/
Func
5/25/78
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OP 4378 Excess Flow Check Valve Functional
7/13/78 OP A379 Orywell/ Torus Differential
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Press Calib
12/22/77 OP 4398 RPS Scram Reset Delay
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Func/Calib
6/12/78
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OP 5221 480 VAC Circuit Bkrs inspection, Testing and Calibration
5/8/78 OP 5222 4KV AC Circuit Bkr Inspection,
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Calibration and Testing
7/13/78
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OP 5223 Emergency Diesel Generator Maintenance
6/6/78
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OP 5310 Repair of Safety Related Inst &
Components
8/3/77
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OP 5311 Calibration of SLC Instr"-
mentation
12/6/77 OP 5312 Calibration of Core Spray
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System Balance of Plant Instrumentation
12/22/77
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OP 5313 Calib of RHR/LPCI Sys Balance of Plant Inst.
12/6/77
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OP 5314 Calib of HPCI System Bal of Plant Instr.
7/13/78 OP 5315 Calib of RCIC Sys Balance of
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Plant Inst.
5/8/78
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OP 5316 Calib of Station Service Water Instr. Sys.
5/8/78
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OP 5317 Calib of RHR Serv Water Balance of Plant Inst.
12/6/77
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OP 5318 Calib of CRD Hydraulic Control Unit Inst.
6/6/78
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OP 5319 Calib of Rx Jet Pump Flow Inst.
12/6/77
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OP 5326 Calib of Diesel Fuel Oil Storana Inst
5/8/78 i0/4 208
.
Proc. 14o.
Title Rev.
Date OP 5329 Calib of SGTS Balance of Plant
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Instruments
6/6/78 b.
Findings (1) The inspectors identified the following discrepancies with respect to the return-to-normal criteria during the review of the preceding procedures:
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OP 4121, Rev 10, " Reactor Core Isolation Cooling System"; Step B-6 positions valve RCIC-27, and the procedure does not restore the valve back to its original position.
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OP 4124, Rev 10, " Residual heat Removal System";
Step B-11 pcsitions valves RHR-39 ano RHR-89 and the procedure does not restore these valves back to their original position, however, the procedure does require that the system be returned to the standby status in accordance with the operational valve lineup of the RHR system.
This is inconsistent with other procedures of this type which in addition to including a generic, " return to normal" statement, also insure that all valves have been previously returned to the normal valve lineup positon by the procedural steps.
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OP 4368, Rev. 8, "RCIC System Isolation ' A' Logic Test"; Valves 13-18, 13-34, 13-35 are initially checked closed during the procedure prerequisites.
The end of the procedt. e leaves these valves in the open position, apparently not returning the valves to the ncrmal positions.
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OP 4378, Rev. 7, " Excess Flow Check Valve Functional Test"; The procedural steps isolate instruments from op2 ration and do not return the instruments back to service.
The licensee will revise these procedures by July 25, 1979.
Until these procedures are revised, the licensee will independently verify that the valves are returned to their correct positions following performance of these tests.
These items are unresolved (271/79-08-07).
)Glh
.
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(2) The preceding review also identified the following procedural discrepancies:
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OP 4341, Rev. 8, "Hign Drywell Pressure Containment Spray Interlock Func/Calib." The step sequence on the data sheet is not consistant with the appropri-ate steps in the procedure body.
OP 5313, Rev. O, " Calibration of RHR/LPCI System
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Balance of Plant Instruments." The acceptance criteria for D/P cell 10-148A is not consistant with the acceptance criteria of D/P cells of the same type, however, the recorded surveillance data for D/P cell 10-148A was within the acceptance range of the corresponding typeset D/P cells.
The licensee will revise these procedures to assure the data sheet is consistent with the procedures and the DP cell acceptance criteria is correct.
These items are unresolved (271/79-08-08).
c.
The latest surveillance test conducted for each ESF system (identified in paragraph Sa) was reviewed to verify that the acceptance criteria were met.
No additional discrepancies were identified.
6.
Review of Administrative Controls for System Taqqing and Return to Service The following procedures were reviewed to verify the adequacy of the administrative controls developed to assure systems are properly returned to service and whether tagging practices provide the potential for obscuring various indicators:
A.P. 0140, Vermont Yankee Local Control Switching Rules (DI
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79-19);
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A.P. 0020, Lifted Lead / Installed Jumper Request Procedure, Rev. 3;
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A.P. 0021, Maintenance Requests, Rev. 7;
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A.P. 0025, Plant Equipment Contre'
(DI 79-5);
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A.P. 4000, Surveillance Testing Control, Rev. 5; and, it)/4 210
,
A.P. 0153, Maintenance of Operations Departmental Logs Rev. 6.
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During the inspection, the licensee was in process of revising facility procedures to emphasize and/or clarify the return to normal criteria and the verification of operability of redundant safety related systems prior to removing a system from service.
This effort was about 90 percent complete and included a revision to A.P. 0153.
The licensee will complete these revisions within 30 days.
This item is unresolved (271/79-08-09).
7.
Independent Verification of Valve / Breaker / Switch Alignments The inspector discussed the use of an " independent verification" of valve, breaker, or switch alignments by a second operator with licensee representatives.
These representatives stated that they do not perform such verifications and felt that they would not improve the system lineup confirmation.
The inspector had no further questions on this item at this time.
8.
Activation of RCIC or HPCI for Reactor Vessel Level Control The inspector queried a lice.see representative to determine if actuation of RCIC or HPCI is required to assist in level control of the reactor during a routine coerational level transient.
The licensee representative stated that either RCIC or HPCI (or both systems together, if necessary) are required to be actuated to assist with reactor vessel pressure and level control.
The following emergency procedures require system actuation:
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0.P. 3100, Reactor Scram Emergency Procedure; 0.P. 3103, Loss of Normal Power Emergency Procedure;
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0.P. 3111, Loss of Condenser Vacuum Emergency Procedure; and, 0.P. 3112, Loss of Feedwater Emergency Procedures.
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The inspector had no further questions on this item at this time.
9.
Station Personnel Change A personnel change has been made in the I&C Supervisor position.
The individual assuming this position was the Technical Assistant to the I&C Supervisor.
10/4 211
.
The inspector reviewed this individual's qualifications to verify that these qualifications were consistent with facility Technica:
Specifications and in conformance with the requirements of ANSI 18.1-1971.
No inadequacies were identified.
10.
DC Power System Review Review of drawings related to the DC power busses and batteries identified a possible problem with the battery output breaker.
If this breaker were open or tripped, the battery would be out of service.
Review of the system design indicates that this occur-rence would be undetected.
The licensee conducts a weekly surveillance test of the batteries.
This test would detect this breaker condition if battery discharg-ing were occurring (an expected result over a period of time),
however, it may be a number of weeks before this condition was determined.
This item was discussed with a licensee representative who acknow-ledged the inspector's findings.
The licensee revised the daily round sheet (that is conducted each shift) that is an appendix to A.P. 0150, Responsibilities and Authorities of Operations Depart-ment Personnel, as DI 79-8, to include a check of the batteries'
output breakers.
11.
Facility Tour During the course of this inspection, numerous tours were conducted of the reactor building, turbine building, control room, screen house, rad-waste building, cable spreading room, switchgear room, diesel generator rooms, and the torus areas.
On one of these tours, the inspectors noted that the bullet proof door providing access from the torus area to the RCIC room would not latch.
Entrance into this room from the torus was prevented by an additional locked cage style door.
The reason the bullet proof door would not latch was due to incomplete work on the latch.
The licensee will complete the work on this latch to assure the bullet proof door will lock.
This item is unresolved (271/79-08-10).
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12.
Review of Licensee Response to IE Bulletin 79-08 The licensee's response to the IE Bulletin 79-08, Events Relevant to Boiling Water Power Reactors Identified During Three Mile Island Incident, was reviewed and evaluated both in office and on site.
This review verified that the licensee's response was timely, accurate, and adequate and that actions taken by the licensee as the result of his review of the Bulletin were accomplisned as stated.
No items of noncompliance were identified.
13.
Review of Licensee Event Report The inspector reviewed Licensee Event Report (LER 79-11) to verify that:
The report accurately describe the event;
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The safety significance is as reported;
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The report is accurate as to cause;
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The report satisfies requirements with respect to information provided and timing of submittal;
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Corrective action is appropriate;
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Action has been taken; and
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The event was reviewed and evaluted W the Plant Operations Review Committee (PORC).
No inadequacies were identified.
14.
Unresolved Items Unresolved items are those items for which further information is required to determine whether they are acceptable or items of noncompliance.
Unresolved items are contained in Paragraphs 2.b, 4.b., 4.c., 5.b., 6 and 11 of this report.
15.
Exit Interview The inspectors met with licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on May 25, 1979 and summarized the scope and findings of the, inspection as they are detailed in this report.
During this meeting, the unresolved items were identified.
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