IR 05000271/1982003
| ML20052D239 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 04/20/1982 |
| From: | Collins S, Gallo R, Raymond W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20052D231 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM 50-271-82-03, 50-271-82-3, NUDOCS 8205060392 | |
| Download: ML20052D239 (21) | |
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Docket Nos. 50-271 U.S. NUCLEAR REGULATORY COMMISSION y
820220
REGION I
820216 820223 820309 Report No.
82-03 Docket No.
50-271 Category C
License No.
DPR-28 Priority
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Licensee:
Vennont Yankee Nuclear Power Corporation
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1671 Worcester Road Framingham, Massachusetts 01701 Facility Name: Vermont Yankee Inspection at: Vernon, Vennont Inspection Conducted: February 2 - March 31, 1982 Inspectors:
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[dift l'Affut
/'/!82 W. J.' Raymond, enior Resident Inspector W E % 1 fcuita D 4/1/st S.J.ColTi'nTfReside6tInspector Approved by: 2 A
20h1 l
R.' M. Gallo, Chief, Reactor Projects I
Section IA, Projects Branch #1 Inspection Summary:
Inspection on February 2-March 31, 1982 (Report No. 50-271/82-03)_
Areas Inspected:
Routine, announced inspection on regular and backshifts by Resident
Inspectors of: actions taken on previous inspection findings; plant operations, including logs, records, equipment status and safety systems; physical security; safeguard system operability; surveillance activities; maintenance activities; plant event followup; licensee event reports, HPCI isolation setpoint; NUREG 0737 TAP requirements; and, emergency preparedness. The inspection involved 189 inspector hours by two resident inspectors.
Results:
No viblations were identified.
8205060392 820422 PDR ADOCK 05000271
PDR Region 1 torm n (Rev. April 82)
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DETAILS 1.
Persons Contacted The below listed technical and supervisory level pcrsonnel were among those contacted:
Vermont Yankee Nuclear Power Corporation Mr. R. Branch, Operations Supervisor Mr. P. Donnelly, Instrument and Control Supervisor Mr. R. Kenny, Engineer, Assessment Coordinator Mr. L. Goldthwaite, Instrument and Control Foreman Mr. R. Lopriore, Technical Assistant Mr. M. Lyster, Operations Superintendent Mr. R. Mossey, Technical Assistant
- Mr. W. Murphy, Plant Manager Mr. D. Phillips, Technical Assistant Mr. D. Reid, Engineering Support Supervisor Mr. D, Weyman, Chemistry and Health Physics Supervisor
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2.
Status of Previous Inspection Findings a.
(Closed) Followup Item (50-271/80-22-05):
RBCCW Pump Design Require-ments. The Reactor Building Closed Cooling Water Pump starting logic and normal operational status were reviewed and found to meet the design bases established in FSAR Sections 8.5 and 10.9.
No inade-quacies were identified. This item is closed.
b.
(Closed) Unresolved Item (50-271/80-18-01): Adequacy of Typical H Penetration Configuration as a 3 Hour Rated Seal. The licensee sub-mitted for NRC Staff Review the completed test results from Fire Test Procedure CTP-0404, summarized in a construction Technology Laboratory (PCA) February 5, 1982 Letter to Chemtrol Corporation.
Based on a Fire and Hose Steam Test conducted on February 4, 1982, the Typical H seal system qualified for a three hour fire-endurance period. This item is closed.
c.
(Closed) Unresolved Item (50-271/80-18-03):
Proposed Corrective
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Actions for Fire Barrier Seals. VY evaluations and corrective action
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plans were submitted by letter WVY 80-172 dated December 19, 1980.
Satisfactory completion of the fire test described in item 2.b above demonstrated a three hour fire rating for the subject penetration seals.
This item is closed.
d.
(Closed)UnresolvedItem(50-271/81-18-02): APRM Inputs to Rod Block
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l Circuits. Concerns identified by this item have been resolved, based on the inspector's review decumented in paragraph 13 of NRC Region I Inspection Report 50-271/81-19. This item is close _ -
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(Closed) Followup Item (50-271/80-10-03): Recirculation Pump B
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Cooler Leakage Alam. Repairs to the electrical circuitry for the B Recirculation Pump Motor Cooler leakage alam were completed i
during the 1980 refueling and maintenance outage. This item is
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closed.
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(Closed) Followup Item (50-271/80-17-06): OP 3013 Revisions.
OP 3013 was replaced in 1981 by procedure OP 3513, which provides
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instructions for the initial assessment and evaluation of offsite
radiological conditions. The discrepancies identified in OP 3013
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do not apply to the new procedure. This item is closed.
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3.
Shift Logs and Records a.
The inspector utilized the following plant procedures to detemine the 19ensee established administrative requirements in this area
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in p.eparation for review of various logs and records.
AP 0150, Responsibility and Authority of Operations Department
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Personnel, Revision 17, dated December 18, 1981
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f OP 0153, Operations Department Communications and Log Main-
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tenance, Revision 9 dated August 17, 1981 j
AP 0140. VY Local Control Switching Rules, Revision 5, dated I
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October 16, 1981
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AP 0020, Lifted Lead / Installed Jumper Request Procedure,
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Revision 4, dated October 16, 1980
AP 0021, Maintenance Requests, Revision 10, dated December 30, 1981
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The above procedures, Technical Specifications, ANSI N18.7-1972
" Quality Assurance Requirements for Nuclear Power Plants" and
10 CFR 50.59 were used by the inspector to determine the accepta-
bility of the logs and records reviewed.
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b.
Shift Logs and operating records were reviewed to verify that:
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Operating logs and surveillance sheets were properly completed
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and that selected Technical Specification limits were met, j
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i Control Room log entries involving abnomal conditions provided
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sufficient detail to comunicate equipment status, lockout
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status, correction and restoration.
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Log Book reviews were conducted by the staff.
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Operating and Special Orders did not conflict with Technical
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Specifications requirements.
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Jumper (Bypass) log did not contain bypassing discrepancies
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with Technical Specification requirements and that jumpers were properly approved prior to installation.
c.
The following plant logs and operating records were reviewed periodi-cally during the period of February 2-March 31, 1982:
Control Room Log
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Night Order Book Entries
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CR Information Book Jumper / Lifted Lead Log Book
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Safety Related Maintenance Requests
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Control Room Operator Round Sheet
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Auxiliary Operator Rounds Sheet
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Communications Log
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Switching Order Log
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Chemistry Log Sheet
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Shift Turnover Checklist
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Surveillance Log
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Potential Reportable Occurrence Book
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Discharge Records
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Radiochemistry Analysis Log
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No violations were identified.
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4.
Plant Tours
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Plant tours were conducted routinely during the inspection period to observe activities in progress and verify compliance with regulatory and administrative requirements.
Tours of accessible plant areas in-cluded the Control Room Building, Turbine Building, Reactor Building, Diesel Rooms, Intake Structure, Security Gate House 2 and Alann Station, l
Radwaste Building, Control Point Areas and the grounds within the Protected Area.
Inspection reviews and findings completed during the tours were as described below.
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a.
Control Room Panel Reviews The operational status of standby emergency systems and equipment /
systems aligned to support routine plant operation was confimed by direct review of control room panels. The following items were reviewed to verify adherence'to Technical Specification Limiting Conditions for Operation (LCOs) and approved procedures.
Switch and valve positions required to satisfy LC0's, where
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applicable and personnel knowledge of recent changes to proce-dures, facility configuration and existing plant conditions.
Alarms or absense of alarms. Acknowledged alarms were reviewed
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with on shift licensed personnel as to cause and corrective actions being taken, where applicable.
Review of " pulled alarm cards"with on shift personnel.
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Meter indications and recorder values.
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Status lights and power available lights.
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Front panel bypasses.
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Computer printouts.
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Comparison of redundant readings.
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No violations were identified.
b.
Radiological Controls l
Radiation controls established by the licensee, including: posting i
of radiation areas, radiological surveys, condition of step-off-pads, and disposal of protective clothing were observed for conformance with the requirements of 10 CFR 20 and AP 0503, Establishing and Posting Controlled Areas, OP 4530, Dose Rate Radiation Surveys, OP 4531, Radioactive Contamination Surveys, AP 0504, Shipment and Receipt of Radioactive Materials.
Confirmatory surveys were per-formed in areas toured to verify established posting of radiological conditions was proper. Radiation work permits were reviewed to verify confomance with procedure AP 0502, Radiation Work Permits.
No violations were identified, c.
Plant Housekeeping and Fire Prevention Plant housekeeping conditions, including general cleanliness and storage of materials to prevent fire hazards were observed in all
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areas toured for confomance with AP 0042, Plant Fire Prevention and AP 6024, Plant Housekeeping.
No violations were identified.
d.
Fluid Leaks and Piping Vibrations Systems and equipment in all areas toured were observed for the existence of fluid leaks and abnomal piping vibrations.
No inadequacies were identified.
e.
Pipe Hangers / Seismic Restraints Pipe hangers and restraints installed on various piping systems were observed for proper installation, tension, and condition.
No inadequacies were identified.
f.
Control Room Manning / Shift Turnover Control Room Manning was reviewed for confomance with the require-ments of 10 CFR 50.54 (k), Technical Specifications, AP 0152, Shift Turnover, AP 0150, Responsibility and Authority of Operations Depart-ment Personnel and AP 0036, Shift Staffing. The inspector verified, during the inspection, that appropriate licensed operators were on shift. Manning requirements were met at all times.
Several shift turnovers were observed during the course of the inspection. All were noted to be thorough and orderly.
No items of noncompliance were identified.
g.
Equipment Tagout and Controls Tagging and controls of equipment released from service were reviewed during the inspection tours to verify equipment was controlled in accordance with AP 0140. VY Local Control Switching Rules.
Controls implemented per Switching Orders (S0) 82-12,81-434, 82-125 and 82-121 were reviewed.
No inadequacies were identified.
h.
Analyses of Process Liquids and Gases Analyses results from samples of process liquids and gases were re-viewed periodically during the inspection to verify confomance with regulatory requirements. The results of isotopic analyses of rad-waste, reactor coolant, off-gas and stack samples recorded in shift l
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logs and the Plant Daily Status Report were reviewed to verify that Technical Specification limits were not exceeded and that no adverse
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trends were apparent. Boron analysis results reported for the Standby Liquid Control system on March 8 and March 30,1982, were reviewed.
No inadequacies were identified, except as discussed in paragraph 10 below.
1.
Jumpers and Lifted Leads (J/LL)
Implementation of J/LL Request No. 81-0094 was reviewed to verify that controls established by AP 0020 were met, no conflicts with the Technical Specifications were created and installation / removal was in accordance with the request.
No violations were identified.
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Conformance with Technical Specification LCOs The operational status of plant systems and equipment was reviewed to verify compliance with selected Technical Specification LCOs.
i Conditions established to meet Technical Specification 3.1.1, 3.3.B.3, Table 3.2.5 and Table 3.2.1 were verified through direct observation and/or surveillance record review.
No violations were identified.
k.
Containment Isolation System valve lineups established to maintain containment integrity
and isolation capability were reviewed on a sampling basis during inspection tours to verify confomance with the configuration specified by OP 2115. The review confimed that manual valves were shut, capped and locked as required by procedure; power was i
available to motor operated valves and no physical obstructions would block operation; and, no leakage was evident from valves, penetrations and flanges.
No inadequacies were identified.
1.
Surveillance Activities Ongoing surveillance testing of safety related equipment was re-viewed to verify the activities were conducted in accordance with approved procedures; test instruments were calibrated; redundant systems were operable and LCOs wat met; testing was conducted by L
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qualified personnel; and, test acceptance criteria were met.
Portions of the following surveillances were observed:
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OP 4334, Reactor Low Pressure, ECCS Pump Start, March 30, 1982
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OP 4313, Reactor ' Water Low Level Scram-Low Low Isolation, March 15, 1982 No inadequacies were identified.
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Radwaste Shipment Preparations and controls established for Radwaste Shipment No. 440-097 were reviewed on February 11, 1981, to verify con-fomance with the requirements of 49 CFR 173, and OP 0504.
Review included a verification of proper packaging, labeling, transporta-tion documentation and vehicle / package radiation levels.
No violations were identified.
5.
Observations of Physical Security The inspector made observations, witnessed and/or verified during regular and offshift hours that selected aspects of plant physical security were in accordance with regulatory requirements, the physical security plan and approved procedures.
a.
Physical Protection Security Organization observations indicated that a full time member of the security
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organization with authority to direct physical security actions was present as required.
manning of all shifts on various days was observed to be as
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required.
b.
Access Control identification, authorization and badging.
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access control searches, including, when applicable, the use
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of compensatory measures during periods when equipment was inoperable.
escorting,
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c.
Physical Barriers selected barriers in the protected areas and vital areas were
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observed and random monitoring of isolation zones was perfomed.
Observation of vehicle searches were mad.
inspector tours of gate house 2, the Central and Secondary
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Alam Stations were conducted at random periods.
No violations were identified.
Safeguard System Operability Reviews of the core spray, high pressure coolant injection, standby liquid control and low pressure coolant injection systems were conducted to verify the systems were properly aligned and fully operational in the standby mode. Reviews of the above systems included the following:
verification that procedure OP 2123 was technically correct as
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compared to system flow diagram G191168 and as noted by walk down of the systems; verification that each accessible valve in the flow path was in
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the correct position by either visual observation of the valve or remote position indication; verification that accessible power supplies and breakers were
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properly aligned for components that are required to actuate upon receipt of a safety injection signal; visual inspection of major components in the selected system for
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leakage, proper lubrication, cooling water supply, general condition and other factors that might prevent fulfillment of their functional requirements; and, verification that key instrumentation required for systen operation
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was functional and calibrated.
No inadequacies were identified.
7.
Surveille.nce Testing The inspector observed or reviewed portions of the following surveillance tests to verify that:
testing was perfomed in accordance with approved, technically adequate procedures by qualified personnel; test instrumenta-tion was calibrated; test data was accurate and complete, and demonstrated confomance with Technical Specification requirements; testing was completed in accordance with the established schedule; Technical Specification LCOs '
were met while testing was in progress and system restoration to service was proper; and, activities were in compliance with AP 4000, Surveillance Testing Control.
OP 4322, MS Line Area High Temperature Functional / Calibration,
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March 15, 1982 OP 4121, RCIC Valve Operability, February 15, 1982
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OP 4117, SGTS A Perfomance Check, February 15, 1982
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OP 4310, Scram Discharge Volume Functional / Calibration,
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March 31, 1982 OP 4300 Source Range Monitor Functional / Calibration,
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March 31, 1982 OP 4301, Intemediate Range Monitor Functional / Calibration,
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March 31, 1982 OP 4302, Average Power Range Monitor Functicnal/ Calibration,
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March 31, 1982 No violations were identified.
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Maintenance Activities The maintenance request log was reviewed periodically during the inspection period to detemine the scope and nature of work done on safety related equipment. The review also confimed that:
the repair of safety related equipment received priority attention; no backlog of required repairs developed on safety related systems;
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and, the perfemance of safety related systems was not impaired.
Portions of the following maintenance activity were observed / reviewed
by the inspector to verify that work was completed in accordance with approved procedures; established radiation controls were proper; personnel conducting the work were qualified; and, equipment under repair was properly returned to service, including completion of re-quired operability testing.
RP 5320. Turbine EPR MOOG Valve Replacement, Revision 4,
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March 31, 1982.
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No violations were identified.
9.
HPCI Steam Line Isolation Setpoint NRC Region I Potentially Generic Issue Report No. 82-06 concerned a problem identified at another BWR facility, where the Technical Speci-fication set point for the HPCI Steam Line High Flow isolation had been found non-conservatively high. The Instrument and Control Super-visor was interviewed and station records were reviewed to detemine whether this issue was applicable to Vemont Yankee.
Vermont Yankee Technical Specification Table 3.2.2 requires the HPCI Steam Line isolation trip setpoint to be set at less than or equal to 195 inches of water.
Procedure OP 4356 is used to establish a setpoint of 180 inches of water for differential pressure switches (steam line elbow taps) dPis 23-77 and 23-76, with corrections applied for elevation
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(head) differences. The OP 4356 setpoint agrees with the value pre-sented in VY FSAR Table 7.3.2, which corresponds to 300% of the Mode-A-Accident steam flow condition of 173,000 lbs./hr. HPCI steam line tap differential pressure measurements recorded per Startup Test Procedure STP 15 on April 5,1974, yielded 12 inches for steady condi-tions and a maximum of 60 inches of water during the HPCI startup trans-ient. Using FSAR Figure E.4-1 Ebasco Drawing 5920-784 and the calcula-tion methods of STP 15, it was shown that the setpoint of 180 inches conservatively represents 300% of the HPCI steam line flow for rated Accident mode conditions.
Based on the above, VY Technical Specification Table 3.2.2 and the existing HPCI steam flow trip setpoint are correct.
No inadequacies were identified.
10.
Inspector Followup of Events The inspector responded to events that occurred during the inspection period to verify continued safe operation of the reactor in accordance with the Technical Specifications and regulatory requirements. The
following items, as applicable, were considered during the inspector's review of operational events:
observations of plant parameters and systems important to safety
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tt confirm operation within approved operational limits; description of event, including cause, systems involved, safety
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significance, facility status and status of engineered safety features equipment; details relating to personnel injury, release of radioactive
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material and exposure to radioactive material; verification of correct operation of automatic equipment;
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verification of proper manual actions by plant personnel; and,
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verification of adherence to approved plant procedures.
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a.
Plant Shutdown from Full Power A pressure transient in the turbine control system at 8:40 P.M. on l
March 30, 1982, resulted in a slight power increase from full power j
and consequent automatic reactor scram on high flux. Plant safety equipment responded properly. No isolation or safeguards actuation occurred. The reactor remained shutdown pending completion of investigation / repairs on the turbine control system and replacement of the brushes (preventative maintenance) on the recirculation MG i
sets. The reactor resumed power operations on March 31, 1982.
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The inspector interviewed control room personnel and reviewed logs and strip chart recordings to detennine the cause and sequence of the event, along with licensee findings and corrective actions for the turbine control system. A malfunction within the electrical pressure regulator (EPR) caused a mal-positioning of the turbine governor valves and resultant spike in reactor pressure to 1020 psig.
The pressure spike in turn caused a power (flux) spike, as observed
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on all 6 APRM channels, with a peak value of 116% FP recorded on APRM channel A.
The conservative APRM trip setpoints established by OP 4302, in con.iunction with core flow of 45.5 million Ib/hr (less than 100% rated), resulted in the automatic scram from the variable power / flow trip system.
Two other lesser pressure spikes were noted by the operations crew at 6:05 P.M. on March 30, 1982.
Suspecting a malfunction in the mechanical pressure regulator (MPR), the MPR setpoint was adjusted to preclude interference with EPR operation. This action contributed to the reactor scram, as subsequent investigation showed proper MPR operation. However, no indication available on CRP 9-7 provides information to help the operator differentiate between EPR/MPR malfunctions.
Licensee investigation of the EPR yielded no definitive cause for its malfunction. Based on consultations with the system vendor, the most likely cause was postulated to be a momentary perturbation in MOOG valve output resulting from control oil debris which became trapped in its supply filter. Control oil filters were inspected and cleaned, and the MOOG valve was replaced.
Subsequent plant operation on April 1, 1982, with the EPR in control proceeded with-out incident.
The inspector witnessed portions of the plant startup in progress on March 31, 1982, and verified completion of startup prerequisites and adherence to startup procedure. A failure in the Rod Worth Minimizer to lock in on Sequence E-A-2 at 2:30 P.M. caused the operators to bypass the unit. A second licensed operator was used to verify rod pulls in accordance with Technical Specification
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3.3.B.3.
IRM channels D and E were bypassed due to noise problems and startup proceeded with the remaining IRM channels operable, in accordance with Technical Specification Table 3.2.1.
No conditions adverse to plant safety were identified during the above reviews. However, the inspector noted that 4 of 7 inadvertent scrams in the last two years have stemed from turbine control system problems. While not considered excessive, this matter and the inspector's concerns regardir.g safety system challenges and reactor cycles were discussed with licensee mancgement on March 31, 1982.
The Plant Manager noted the inspector's comments.
No violations were identified.
b.
Standby Liquid Control System Boron Concentration Vermont Yankee submitted LERs 82-03/1T and 82-07/1T in accordance with Technical Specification 6.7.B.1 to describe actions taken for Standby Liquid Control (SLC) system problems. The inspector re-viewed the event cause and the adequacy of the associated corrective actions.
(1) LER 82-03, SLC Tank Boron Concentration Following addition of water to the SLC Tank as a result of inservice inspection testing of the pumps on February 19, 1982, a sample of the solution'was taken and analyzed as required by Technical Specification 4.4.C on February 20, 1982. The shift supervisor was notified at 4:20 P.M. on February 21, 1982, that the boron concentration was 8.5 wt.%, which was less than the 9.3 wt.% limit specified by Technical Specification 3.4.C for the volume of solution in the tank. The shift supervisor took imediate action to comence a plant shutdown per Technical Specification 3.4.0 and restore tank concentration to the re-quired value. Upper plant management, the resident inspectors and ENS notifications were made.
Following a re-reading of Technical Specification Section 3.4.0, upper -ite management concluded an immediate plant shutdown was not required for out-of-spec boron concentrations and power de-crease was terminated at 96.5% FP at 5:30 P.M. on February 21, 1982. Following additions to the storage tank per OP 2114, pre-liminary analysis results at 8:15 P.M. on February 21, 1982, showed the boron concentration to be within specification. The proper concentration was confirmed by further samples and analyses at 1:35 A.M. and 8:30 A.M. on February 22, 1982.
Initial licensee evaluations of the history of water additions to the SLC tank did not explain the unexpected decrease in tank concentration. An investigation was initiated to determine the cause of the concentration reduction and a program of increased sampling was instituted until the cause fo: the reduction is known. The required SLC tank concentration / volume was administra-tively raised to provide additional margin to the Technical Speci-fication 3.4.C limit.
The following findings resulted from inspector review of the event and discussions with the Plant Manager on February 22, 1982:
(a) The inspector presented the following NRC position during the February 22, 1982, meeting and during a February 21, 1982, telecon with the Plant Manager: The Technical Specification 3.4.C volume / concentration limits are part of the operability
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requirements for the SLC system and constitute conditional requirements for the SLC system to perform its intended function. Failure to meet the Technical Specification 3.4.C limits would constitute cause to implement the Technical Specification 3.4.D action statement. The actions to be taken are discussed further below.
The Technical Specification 3.4 LCO wording is sufficiently vague to support an erroneous reading that the SLC tank concentration / volume limit is not part of the LCO. The inspector stated that VY should submit a proposed Technical Specification 3.4 change for NRC approval that would clarify the LC0 requirements regarding the SLC Tank parameters, and specify the actions to be taken when the tank requirements are not met. This item is considered unresolved pending submittal of the Technical Specification 3.4 change to the NRC(UNR 50-271/82-03-01).
(b) The need to clarify station policy regarding SLC LC0 adherence was discussed during the February 22, 1982 meeting. The following gui, dance was issued.by-the licensee in a February 22,.1982 memorandum to Operations and Chemistry Departments:
"Upon discovery of an out-of-speci-fication concentration or temperature condition in the SLC tank, actions will be initiated immediately to return the parameter to within prescribed limits; if after 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> the tank is not within limits, an orderly shutdown will be initiated such that the plant will be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the time of the initial discovery of the out-of-specification condition."
Consideration would also be given to the issuance of a M00 Directive for this item.
(c) Licensee '~nstigation of the activities associated with SLC test 0.1g and sampling during the period of February 19-February 21, 1982, detennined that the first sample of the SLC tank showing an out-of-specification condition was taken at about 10:30 P.M. on February 20, 1982. The Chemistry Technician who completed the analysis failed to immediately notify the shift supervisor as required by OP 4611. The individual responsible for the February 20, 1982, analysis, along with all other techni-cians, were re-instructed on the OP 4611 requirements and the need to follow procedures. Failure to follow OP 4611 is considered to be a licensee identified violation of Technical Specification 6.5.1 requirements.
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(d)
Inspector review of this event determined that: actions taken by the shift supervisor to initiate a plant shutdown at 4:20 P.M. on February 21, 1982, were appropriate; the Technical Specification 3.4 LCO was not violated based on an initial determination of an out-of-specification condition at 10:30 P.M. on February 20, 1982, and a restoration of the boron concentration to within limits by 8:15 P.M. on February 21, 1982; and, the SLC system could have perfomed its intended function while in the degraded condition due to the 25% margin built into the specifications.
The inspector had no further cbmments regarding LER 82-03.
(2) LER 82-07, SLC Analysis Errors Subsequent licensee review stemming from the investigation associated with LER 82-03 revealed calculational deficiencies in the procedures for controlling boron concentration in the SLC storage tank. Specifically, procedure OP 4611 failed to account for solution specific gravity in the conversion from solution weight percent to ppm-Boron. This resulted in a non-conservative 5% error in the calculated margin. Additionally, the correlation of tank concentration vs. level failed to account for the specific gravity of sodium pentaborate. This resulted in an additional nonconservative 5% error when determining solution volumes. Due to the combined errors, the plant may have unknowingly operated below the Technical Specification 3.4 limits.
Licensee evaluation of previous tank volume / concentrations concluded that the SLC system could have perfomed its in-tended function due to established margins above the Technical Specification limit. The operating limits for SLC tank boron concentration / volume were further increased as a result of LER 82-03. Corrections to OP 4611 were made by DI 82-9 dated March 10, 1982.
In addition to the abcve, the plant undertook the following corrective action as a result of LER 82-7.
A task force war established to investigate and resolve the following:
SLC boron concentration history; review of all SLC system proce-dures; level instrumentation calibration; and, design basis for SLC concentration / volume. Corr: active actions associated with the Task Force report will be submitted as a supplement to LER 82-07. This item is considered unresolved pending completion of the licensee actions listed above and subsequent review by the resident inspector (UNR 50-271/82-03-02).
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c.
Uninterruptible Power Supply (UPS) Failures VY submitted LERs 82-2 and 82-5 as a result of UPS-B failures on February 6, 1982 and February 23, 1982, respectively. UPS-B provides nonnal power to MCC 898 which in turn powers the Recirculation and RHR system valves. Both failures were attributed to failures of capacitors in the output transient suppression circuitry. As a re-sult of the February 23, 1982, failure, all UPS-B capacitors were replaced, along with those in UPS-A as preventative maintenance.
Adherence to Technical Specification 3.10.B LCO requirements during periods following the failures was verified by the inspector during routine plant status reviews.
No violations were identified.
11. Licensee Event Report Review The licensee event reports (LERs) listed below were reviewed in the NRC Resident / Regional Office. The reports were reviewed to detennine whether:
the information provided was clear in the description of the event and identification of safety significance; the event cause was identified and corrective actions taken (or planned) were appropriate; the report satis-fied requirements with respect to information provided and timeliness of submittal per NUREG 0161 and Technical Specification 6.7 criteria. Those reports annotated with an asterisk (*) concern events that required inspector followup action and inspector review / evaluation of the event is documented elsewhere, in this or other inspection reports.
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LER 82-1, Reactor Startup with inoperable IRM channels. Event date January 28, 1982; Report date February 25, 1982.
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- LER 82-2, UPS-B Failure.
Event date February 6,1982; Report date March 3, 1982.
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Event date February 20, 1982; Report date March 6, 1982.
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LER 82-4, Missed Data from Environmental Station AT1.4. Event date February 16, 1982; Report date March 16, 1982.
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- LER 82-5, UPS-B Failure.
Event date February 23 1982; Report date March 25, 1982.
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- LER 82-7, SLC Systen Procedure Calculational Errors. Event date March 9, 1982; Report date March 23, 1982.
No vilotions were identified.
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12. Review of NUREG 0737 TMI Action Plan Requirements Implementation of NUREG 0737-TMI Action Plan Requirements was reviewed to detemine the status of those items with a due date of January 1,1982.
The review consisted of establishing a licensee commitment to fulfill a requirement and a followup inspection to detemine its completion status.
References used for this review are listed below.
(1)
NUREG 0737, Clarification of TMI Action Plan Requirements, October 31, 1980 (ii)
NRC Letter to R. L. Smith, dated July 27, 1981 (iii)
Letter FVY 82-14, Item II.B.3, February 19, 1982 (iv)
Letter FVY 81-169, Item II.B.3, November 25, 1981 (v)
Letter FVY 82-1, Item II.B.3, January 5, 1982 (vi)
Letter FVY 81-171. Item II.B.3, November 25, 1981 (vii)
Letter WVY 80-170 Item II.B.3, December 15, 1980 (viii)
NRC Region I Inspection Report 50-271/82-04 and 82-05 a.
TAP Item II.B.3, Post Accident Sampling (1) Requirements: References (i) and (ii)
Provide the capability to perform reactor coolant and contain-ment atmosphere sampling and analysis following an accident, assuming a Regulatory Guide 1.4 release of fission products, without incurring excessive radiation exposures to individuals.
Provide the capability in the radiological spectrum analysis facility to promptly quantify certain radionuclides that are indicators of the degree of core damage.
Provide the capability to perfom certain chemical analyses, including that for boron and chlorides, within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from time a decision is made to take a sample. The design of the sampling facility should consider use of charcoal and HEPA filters in the ventilation exhaust.
(2) Licensee Commitments:
References (iii)through(vii)
New sampling systems required for this item will be installed.
Pre-operational testing, procedure development and personnel training are expected to be completed by February 28, 1982.
VY took exception to the staff requirement to provide for chloride sampling and analysis within three hours. Justifica-tion was provided to support completing chloride analyses within
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four days after an accident occurs. VY took exception to requirement to provide a filtered ventilation exhaust from the new sample panel, based on an evaluation that releases resulting from sample panel operation would contribute a small fraction of 10 CFR Part 100 limits.
(3)
Inspection Findings:
Reference (viii)
Modification completed under EDCR 79-51 for the post accident reactor coolant sample panel and EDCR 81-42/ PAR 81-09 for con-tainment atmosphere sampling were reviewed. This review in-cluded a verification that systems were installed in accordance with the design packages, plant procedures were revised and personnel had been trained. Additional staff reviews of these systems is also documented in Reference (viii).
Inspection findings are summarized below.
(a) The capability for obtair.fng containment atmosphere samples from a location outside the Reactor Building was provided by adding a shielded sample point in the low range (MSA)
hydrogen analyzer lines, down-stream of valves VG 39 and 40.
Lead shielding is provided for the sample location.
Three 40cc sample bombs. 2 lead pigs and a hydraulic jack are provided for obtaining the sample. The method for obtaining a sample from the new location was discussed with the plant Chemical Engineer.
No inadequacies were identified.
(b) A new reactor coolant sampling panel was installed at the East Entrance to the High Pressure Heater Bay in the Turbine Building. The panel obtains serial diluted grab samples from recirculation risers N2C and NZG to assure representative core sampling. A sample point from High Pressure heater E-1B is provided for training.
The capability to flush process lines is provided. The purge from liquid samples are directed to the torus through the HPCI turbine exhaust line. Discharge from the panel gaseous lines is directed to a Turbine Building ventilation exhaust duct located directly above the panel.
The Turbine Building ventilation system exhausts to the l
plant stack.
Electrical power for the panel and valves l
is derived from AC-PP-5, which can be powered from the emergency diesel generators.
l The panel is designed to receive reactor coolant samples under reactor conditions of 2300 psi and 6000F, Samples
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are cooled to room temperature by water bath prior to l
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degassing. Air supply for system solenoid operated valves is from the Control Air System (3 inch IA 5 and 6 headers inside the reactor building). There is no backup air supply from outside the Reactor Building, but provi-sions are included for adding a Standby nitrogen supply.
The panel design provides for handling samples with activity up to 5 ci/gm. VY estimated the worst case sample activity to be 2.5 ci/gm. Serial dilutior.s of reactor coolant water at the panel will result in a re-duction of specific activity by about three orders of magnitude. Shielded sample pigs are provided to trans-port the samples to the analysis facility.
Reference (viii)
documents additional staff reviews regarding panel proce-dures, licensee personnel training and the analytical facilities.
(c) YAEC calculations dated July 23, 1980 and August 6, 1980 (Memo Files 274/80 and 175/80, respectively) provide the engineering evaluation /detemination of the dose rates resulting from panel operation. Based on a source term assumed to contain 100% (noble gas) - 30% (halogen) -
1% (particulate) of the equilibrium core inventory one hour after the accident, the overall dose rate from the shielded panel was shown to be much less than 1 Rem /hr.
The maximum offsite exposures resulting from a single panel operation (40cc sample degassing through unfiltered ventilation system) was calculated to be 1.8 mrem whole body and 4 mR thyroid.
(d) Preoperational testing for the panel was comp 1sted on December 30, 1981. The inspcctor reviewed in detail the OP 3530 valve manipulations required to obtain a reactor coolant sample and witnessed several panel operations while VY staff training was in progress. The procedure was found adequate for obtaining samples. As of February 9, 1982, 8 of 11 Chemistry and Health Physics personnel had attended one training class on the panel.
Two rounds of training / walk throughs were scheduled for all 11 members to complete.
Reference (viii) documents additional findings in this area.
(e) The following plant drawings and procedures were reviewed to verify they had been revised to reflect the modifications:
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B191300, Sheet 51
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B191301, sheets 4, 604, 605
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G191144
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G191330
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SK 994
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590-9191
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OP 2120, DI 81-74
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OP 2143, DI 81-52
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OP 2172, DI 81-60
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RP 2185, DI 81-39
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OP 2190, DI 81-80
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OP 3530, Revision 2
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OP 2115 The valve lineups established by the above procedure changes were reviewed to verify operation of the sample panel would be provided.
Except as noted below, no inadequacies were identified.
As of February 9, 1982, OP 2115 had not been revised to reflect the new position required for recirculation riser
sample line isolation valves V -300D and V2-300C. The valves were observed to be open on February 9, 1982.
DI 82-3 was issued for OP 2115 on February 10,1982, to specify the proper positions for the valves.
No violations were identified.
13. Emergency Preparedness Appraisal and Drill The resident inspectors participated in NRC staff reviews of VY emergency preparedness. NRC staff review of the Annual Emergency Drill conducted on February 18, 1982, is documented in NRC Region I Inspection Report 50-271/82-04. Findings from the NRC staff Appraisal of Emergency Prepared-ness conducted during the period of March 15-24,1982, are summarized in NRC Region I Inspection Report 50-271/82-05.
14. Unresolved Items Unresolved items are items for which further infomation is required to detemine whether the items are acceptable or violations. Unresolved items are discussed in paragraph 10 of this repor..
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15. Management Meetings During the period of the inspection, licensee management was periodically notified of the preliminary findings by the resident inspectors. A summary was also provided at the conclusion of the inspection and prior to report issuance.
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