IR 05000029/1982015
| ML20028G056 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 01/11/1983 |
| From: | Collins S, Gallo R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20028G042 | List: |
| References | |
| 50-029-82-15, 50-29-82-15, NUDOCS 8302070357 | |
| Download: ML20028G056 (17) | |
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DCS Nos.
821012
821112 821116 U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-29/82-15 Docket No.
50-29 License No.
DPR-3 Priority Category C
Licensee:
Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701 Facility Name: Yankee Nuclear Power Station Inspection at: Rowe, Massachusetts Inspection conducted: Noverrber 16 - December 31, 1982 Inspectors: 24fo//#ni_
l/cJa%
S. 'd. Collins, Senior Resident Inspector date' signed
date si ned Approved by:
k
/ // 83 EH. Gallo, Chief,Teactor Projects datd signed Section 1A Inspection Summary:
Inspection on: November 16 - December 31, 1982 Areas Inspected:
Routine, onsite regular and backshift inspection by the resident inspector (141 hours0.00163 days <br />0.0392 hours <br />2.331349e-4 weeks <br />5.36505e-5 months <br />). Areas inspected included previous inspection items; reviews of routine plant refueling outage and plant startup operations; plant maintenance observations; plant surveillance observa'. ion; inspector re-view of plant events; reviews of events requiring one hour notification to the NRC; review of Licensee Events Reports; IE Circular followup; Plant Information Report reviews; and on-site review committee reviews.
Violations: Two:
Failure to perform safety evaluations for system modifications associated with the incore instrumentation flux thimble paths and waste disposal system instrumentation (details 6.A. and 6.B.2.) and failure to notify the NRC Operations Center within one hour of a sionificant tvent as defined by 10 CFR 50.72 (detail 6.
A.).
0302070357 830124 PDR ADOCK OS000029 G
PDR i
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DETAILS 1.
fersons Contacted Plant Operations
- H. Autio, Plant Superintendent E. Begiebing, Maintenance Supervisor W. Billings, Chemistry Manager R. Boutwell, Technical Services Supervisor E. Chatfield, Training Manager B. Drawbridge, Technical Director
- L. French, Plant Engineer T. Henderson, Reactor Engineering Manager K. Jurentkuff, Assistant Plant Operations Manager
- P. Laird, Plant Maintenance Manager
- R. Mitchell, Instrument and Control Supervisor R. Sedgwick, Security Supervisor
- N. St. Laurent, Assistant Plant Superintendent J. Trego, Radiation Protection Manager D. Vassar, Plant Operations Manager The inspector also interviewed other licensee employees during the inspection, including members of the Operations, Health Physics, Instrument and Control, Maintenance, Reactor Engineering, Security and General Office Staffs.
Quality Assurance L. Reed, Operational Quality Assurance Coordinator Yankee Atomic Electric Company P. Bergeron, Manager Transient Analysis Group J. Handschuh, Senior Engineer Transient Analysis Group L. Heider, Vice President of Operations J. Kay, Senior Engineer Licensing
- Denotes those present at exit interview on January 5, 1983.
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2.
2.
Licensee Action on Previous Inspection Findings a.
(closed) Violation (50-29/81-07-01):
Failure to Have Reviewed and Approved Procedures for SR-89, SR-90 and C-14 Vendor Analysis. The inspector reviewed IR 50-29/81-07 Appendix A Notice of Violation, dated July 16, 1981 and Yankee letter FYR 81-122, dated August 7, 1981, NRC response letter deded. October 15, 1981, Yankee letter FYR-81-147 and NRC response letter dated, November 17, 1981. The licensee noted in FYR 81-147 that although they believe a non-u.aformance does not exist; the vendor's procedures were obtained subsequent to FYR 81-122, reviewed and approved by Plant Operations Review Committee (PORC) as indicated in PORC Meeting Minutes 81-47. The inspector reviewed PORC Minutes 81-47, dated August 3,1981 and had no further questions. This item is closed.
b.
(closed) Inspector Follow Item (50-29/82-10-01): Licensee Evaluation of PS-MOV-191 Missing Bonnet Flange Bolt. This item was previously reviewed in IR 50-29/82-12, paragraph 2.g.
The licensee has subsequently doc-umented Yankee Nuclear Services Division (YNSD) analysis of the occurrence and corrective action taken in Plant Information Report (PIR) 82-10, dated December 17, 1982. This item is closed.
c.
(closed) Inspector Follow Item (50-29/82-05-01): Move Fire Trailer Office From Pump House. The inspector verified on December 30, 1982 that the perimeter of the pump house had been cleared of all trailers. This item is closed.
d.
(closed) Inspector Follow Item (50-29/82-10-02): Follow September 23, 1982 Release Calculation and System Repair. The inspector reviewed Plant Information Report (PIR) 82-14, dated, October 20,1982, which documented the licensee's review of the occurrence and corrective actions taken. The inspector had no further questions. This item is closed.
(closed)InspectorFollowItem(50-29/82-12-02): Review Licensee e.
Evaluation of Change in Refueling Boron Concentration. The inspector reviewed Plant Information Report 82-17, dated December 4,1982 for accuracy of event description and corrective actions and had no further questions. This item is closed.
f.
(closed) Inspector Follow Item (50-29/82-12-05): Main Coolant Pump No.
1 Vent Pipe Crack Followup. The licensee replaced the carbon steel vent pipe components with stainless steel and reviewed Jerguson valve applications in Safety Class 1 systems to prevent a similar occurrence.
Licensee EventReport(LER) 82-34 01T-1 was issued December 3, 1982 to document the licensee's actions. This item is closed.
3.
Review of Routine Plant Refueling Outage and Plant Startup Operations A.
Daily Inspection - The inspector verified the following by direct observation of activities, tours of the facility, discussions with plant personnel, independent verification, and facility record review:
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I 3.
1.
Control room activities were observed to verify proper manning and access control; adherence to approved procedures; adherence to Limiting Conditions for Operation (LCO's), Emergency Safe-guard Feature (ESF) status and selected value confirmation using a unit specific checklist; selected instrument and recorder trace review; control room board annunciator status and followup action review; nuclear instrumentation (N/I) and reactor protection system (RPS) operability verification; conformance with refueling shutdown margin limits; verification of containment status required for re-fueling operations; primary vent stack trace review and release followup; verification of onsite and offsite emergency power source availability; control room documents, including operator logs, main-tenance and surveillance documentation and operating orders were reviewed to note trends, apparent anomolies, routine outage operations, and establish items requiring inspector followup.
2.
During daily entry and egress from the protected area (PA) security activities were observed to verify access controls in conformance with the security plan for personnel, packages, vehicles, guard manning and conduct; selected PA barriers, and gates were examined; isolation zone conditions were observed; and licensee monitoring for radioactive materials prior to personnel, materials and equip-ment release for unrestricted use was monitored during egress from the PA.
No inadequacies were identified.
B.
Weekly System Alignment Inspection Operating confirmation was made of suected piping system trains.
Accessible valve positions in the flow path were verified correct.
Proper power supply and breaker alignment was verified. Visual inspections of major components were performed. Operability of instruments essential to system performance was verified. The following systems were checked:
-- Shutdown cooling system removal from service per OP-2162, Operation of the Shutdown Cooling System, Rev, 10, performed on December 7, 1982.
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-- Verification that actions are taken to prepare the plant for cold weather operations per OP-2115, Warm or Cold Weather Operation, Rev.8, performed on December 9, 1982.
No inadequacies were identified.
C.
Biweekly Inspection 1.
Portions of the foliowing selected ongoing ESF maintenance activities were observed to verify redundant system operability; approved pro-cedures in use; and work is being performed by qualified personnel:
-- MR 82-1100 December 7,1982, Primary Plant Instrument No. 2 AC-DC Panel, Incorrect Voltages.
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-- MR 82-1124, December 10, 1982 No. 3 Intermediate Range Nuclear Instrument Channel Start-Up-Rate Scrams at Zero DPM.
No inadequacies were identified.
2.
Portions of the following selected ESF surveillance were observed to verify that test instrumentation was calibrated; redundant system operability; approved procedures used; work perfenned by qualified personnel; and acceptance criteria met:
-- Return of Charging and Bleed Lines to Service after Maintenance, per OP-2169, Rev. 7, performed November 30, 1982.
-- Preparation of the Emergency Core Cooling System for Normal Operation, per OP-2652, Rev. 20, performed December 8, 1982.
No inadequacies were identified.
3.
The inspector independently examined the following tagouts to verify that the valve, breaker or switch was correctly positioned, that the tag was properly attached to the correct component, that the licensee's procedure. AP-0017, Switching and Tagging of Plant Equipment, was adhered to, and that equipmant removed from service conformed to T/S LC0 requirements.
-- YAEC No. 82-01561, Incore instrumentation isolation valve for path K-6, tagged closed, December 11, 1982.
-- YAEC No, 82-01568, Incore instrumentation isolation valve for path A-5, tagged closed, December 13, 1982.
-- YAEC No. 82-01588, VD-MOV-561 power supply breaker on EMCC -1 tagged open, valve verified closed. December 16, 1982.
No inadequacies were identified.
4.
A review of the licensees sampling program was conducted by monitoring results of liquid and gaseous samples during the period to verify conformance with regulatory requirements; and boric acid tank (BAT)
level and sample results were reviewed forconformance with technical specifications on the following dates:
-- Boric Acid Mix Tank Makeup per OP-2167, Rev. 7, performed on December 14, 1982.
-- Release of No. 2 Test Tank per OP-2379, Rev. 7, performed on December 23, 1982, utilizing Release Permit No.82-116.
-- Release of No. 1 Test Tank per OP-2379, Rev. 7, performed on December 24, 1982, utilizing Release Permit No.82-117.
No inadequacies were identified.
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Accessible facility areas were toured to make an independent assessment of plant and equipment. On a sampling basis the following items were observed or verified: condition of selected vital and access controlled barriers; radiation work permit com-pletion and use; protective clothing and where applicable, proper respirator use; personnel monitoring practices; operational status of selected personnel monitors, area radiation monitors and air monitors; equipmei. tagout sample to verify LCO compliance for equipment out of service; plant housekeeping and cleanliness conformance with approved programs, and communication system operability.
Inspector tours included the following areas:
Control room, turbine building, auxiliary boiler room, switchgear room, screenwell house, spent fuel pit, primary auxiliary building, safety injection building, pump and heat exchanger cubicles and
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radwaste handling complex.
No inadequacies were identified.
4.
Plant Maintenance Observation The inspector monitored portions of activities noted below to ascertain that maintenance of safety-related systems and components is being constructed per approved procedures, TS and appropriate codes and standards. Observations and reviews of records were utilized to verify conformance with LC0's; compliance with administrative and tagout procedures; use of maintenance procedures and qualifiea personnel; use of certified replacement parts; adherence to radiological controls, housekeeping requirements, and equip-ment return to service.
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-- MR 82-1159, December 17, 1982, Incore Instrumentation Flux Tube K-6 Leaking, Valved Out and Capped.
-- MR-82-1160, December 17, 1982, Incore Instrumentation Flux Tube A-5 Lecting, Valved Out and Capped.
Except for the following the inspector had no further questions.
Circumstances surrounding MR 82-1159 and 82-1160 are further discussed in section 6.B.2. of this report.
5.
Plant Surveillance Observation A.
The inspector reviewed performance of surveillance testing involving a safety-related system including: review of surveillance procedure for conformance with regulatory requirements; calibration of test equipment; system removal from service and LC0 compliance; monitoring portions of surveillance test performance; portions of system restoration to service; test data review for accuracy and completeness; confirmation
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of licensee test documentation review and discrepancy followup, test result compliance with TS criteria, personnel qualification, and adherence to surveillance' schedule..
-- Main Coolant System Leak Inspection or 2200 PSIG Pressure Test, per OP-4200, Rev. 8, performed on December 8,1982.
No inadequacies were identified.
B.
The inspector witnessed portions of the following surveillance test activities to verify testing scheduled per TS; procedures being followed; personnel qualification; LC0 compliance; correct system restoration.
-- Chemical Shutdown System Operability Check, per OP-4214. Rev. 9, performed December 20, 1982.
-- Fire System Operability Test, per OP-4210, Rev. 11, performed December 24, 1982.
No inadequacies were identified.
6.
Inspector Review of Plant Events A.
Cycle XV-XVI Refueling Operations At the beginning of the inspection perio,1 the reactor plant was in Mode 6. Refueling.
Inspector observatiens and followup on outage activities is noted below:
- Prompt Notification, November 16, 1982, pursuant to 10 CFR 50.72
"tccidental or incontrolled Radioactive Release."
On the morning of November 16, 1982, at 7:35 a.m., tl'e inspector was notified by the Plant Superintendent that a gas release had occurred between the period of November 12-16, 1982 as indicated by an unexplained reduction in Waste Gas System Surge Drum Pressure. The licensee in-dicated that the activity of the gaseous release was estimated to be low based on previous system samples and as indicated by no primary vent stack monitor increase during the suspected release period. The release was terminated at 1:30 a.m. on November 16, 1982 when WD-V-837A valve was found open and subsequently shut by plant operators. The inspector inquired as to whether notifications to the NRC and site management had been made when the event was discovered and had been evaluated by on-shift licensee personnel. The Plant Superintendent indicated that the NRC immediate notification required by 10 CFR 50.72 as implemented by 0P-MEMO No. 2A-1, June 11, 1982, and notification of Higher Plant Management (Duty Officer) required by OP-MEMO No. 2A-1, June 11,1982, had not been made following discovery of the waste gas release through the open valve WD-V-837A.
Subsequently at 7:40 a.m. the inspector monitored the 10 CFR 50.72 notification of the NRC via the RED PHONE and verifie~d that the event
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was accurately described and classified, including recognition to the NRC Operations Center Duty Officer that the event was being reported late. The licensee indicated that follow-up notifications would be made when additional infonnation was available. The inspector notified Region I Management of the event.
Documentation of inspector and licensee follow-up actions is noted below:
1.
A calculation of Waste Gas Surge Drum Discharge activity was per-fonned by the licensee and reviewed by the NRC resident and an on-site Radiation Protection Specialist Inspector. The calculation results were forwarded to the NRC Operations Center Dut via the RED PHONE, this action closed out the licensee'y Officers reporting requirements per 10 CFR 50.72 and was completed at approximately 11:55 a.m.
For calculation purposes, the release duration was estimated as being from November 13, 1982, 4:00 p.m. to November 16,1982,1:30a.m.(57.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />). The inspector verified the release Gas (WG) period estimate as being conservative and based on Waste Surge Drum pressure as indicated by hourly log readings taken by the Auxiliary Operator (AO) and recorded on the Primary A.0. Log Sheet per AP-2007, Maintenance of Operations Departmental Logs, Rev. 9.
The volume discharged was based on daily logs and graph trending of the WG Syste volume, a maximum volume recorded on November 2, 1982 (26,063 ft)minusghevolumerecordedfor 6:25 a.m., November 16,1982{20,902 ft ) results in the conservative volume discharged of 5,161 ft.
Utilizing primary vent stack (PVS)
flow of 35,100 ft3 (2 fan flow as indicate by PVS fan manometer)
and WG header sample resg}ts gr Xg (3, Krg5 (b
on-site analysis) and Ar
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,H based on October 4,1982 sample and off-site analysis by Teledyne), which indicate a total Curie content of 3.59E 1 C1 re was calculated as 1.05E-} eased, the total activity after dilution pC1/cc. The calculated release does not violate VNPS T.S. 3/4.7.7 Waste Effluent section 3.7.7.3 for maximum concentration of radioactive gaseous wastes discharged.
The inspector had no further questions in this area.
2.
The inspector reviewed the circumstances leading to the discovery of the release. Primary A.0. Log Sheets were reviewed from the period of November 12-16, 1982.
During this period the WG Surge Drum pres-sure varied from a maximum of 28 psig recorded at 3:00 p.m. on Novem-ber 12,1982; connencing at 3:00 a.m. on November 13, it gradually decreases (approx.1 psig every 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) until 5 a.m. on November 15; it then decreaser (approx.1 psig every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) to 4 psig recorded at 2:00 a.m. on November 16, when the discharge path was secured. A random review of Primary A.O. Log Sheets for the October 1982 period indicated that normal recorded readings for the WG Surge Tank Drum pressure were between 22 and 28 psig, the expected range of the WG Surge Drum pressure is MAX of 60 psig and a MIN of 5 psig as indicated by the Primary Log Sheet parameter range column.
Licensee approved procedure AP-2007, Maintenance of Operations Departmental Logs,
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Rev. 9 procedure step 8.1, requires that the Shift Supervisor be notified whenever a parameter exceeds its minimum or maximum limit.
This notification was completed by the A.0. on November 16, 1982, when the WG Surge Drum pressure approached 4 psig. AP-2002, Operations Department Personnel Shift Relief, Rev. 11, Form APF-2002.2, Nuclear Auxiliary Operator Shift Relief Checklist, requires a review of the Primary Log Sheet and verification that all parameters are within the minimum and maximum range, all off nomal conditions be listed and the Shift Supervisor be notified immediately. Technically the WG Surge Drum pressure was maintained within its MAX to MIN values; however, the per shift reviews of the log by the oncoming and off-going Primary A.0.'s should have indicated an off normal
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decreasing pressure trend prior to November 16, 1982.
It was brought to the inspectors attention that the Primary and Secondary A.0. Log Sheets are currently not required to be reviewed on a shift basis by a licensed operator. The Log Sheets encompass a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period and are ultimately reviewed by the Plant Operations Manager, Plant Supvintendent, and Asst. Plant Superintendent. A review of the logs from November 12-16, 1982, indicated that senior site management review typically took place within 1-2 days following completion of the Log. This type of review is timely for management to maintain current on plant status, but does not serve to provide for a per shift review of pertinent plant parameters by a licensed operator, similar to the requirements of AP-2002, Operations Department Personnel Shift Relief, for the Shift Supervisor to re-view control room plant logs and books prior to assuming the shift.
The inspector noted the above findings to licensee management during a...eeting on November 19, 1982 and subsequently verified that Special Order No. 700, dated November 22, 1982 directed Shift Supervisors to review all department log sheets at least once per shift and noted that licensee procedure AP-2007, Maintenance of Operating Department Logs, would be revised to include this requirement.
The inspector also reviewed Yankee Atomic at Rowe memorandum, dated November IG,1982, from H. A. Autio, Plant Superintendent, to Operations Personnel. The memorandum noted that all logs must be reviewed for trends and to establish the basis for the trend.
The inspector had no further questions in this area.
3.
During the review of the Primary A.0. Log Sheet the inspector noted that no values were recorded for W.G. Surge Drum Temp. (cF). This parameter is pertinent in that the W.G. Surge Drum (TK-36) and a major portion of the Waste Gas Header is located outside and is therefore subject to temperature changes which ultimately can effect WG Surge Drum pressure. The inspector reviewed I&C Equipment History-Data for WD-TI-300, Tank and Evaporator Temperature multi-point indicator, and noted that the unit has been declared "not operating" since an attempt to repair the instrument on November 24, 1981 had failed and Maintenance Request (MR) 81-1262, 8 Pt. Temperature
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- Indicator Waste Disposal Col. ol Panel, Erratic Indications, written November 23, 1981 remains outstanding in the MR Log. The inspector determined that Plant Alteration (PA) 82-02. Replace the Multi-Point Temperature Indicator which is Located on the Waste Disposal Control Board, dated Feburary 23, 1982, documents the intention to correct the failed indicator.
The inspector asked the licensee's represen-tative why the work received low priority. The instrument originally monitored the following points:
1.
TI-301, Primary Drain Collecting Tank - TK-30 2.
TI-302, Waste Holdup Tank - TK-31 3.
TI-303, Activity Dilution Decay Tank - TK-32 4.
TI-304, Test Tank, TK-34-1 5.
TI-305, Test Tank, TK-34-2 6.
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TI-307, Waste Gas Surge Drum, TK-36 8.
TI-308 Evaporator, EV-1 9.
Stripper Feed Presently, temporary meters are located within the back of the
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instrument cabinet to monitor the following points:
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Primary Drain Collecting Tank - TL-30 Activity Dilution Decay Tank - TK-32
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Evaporator Temperature
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The temporary 0-3000F (or 0-150 C) meters have no calibration verification attached, are suspended from the internals of the cabinet hardware and require opening of the rear cabinet door to monitor indication.
A review of completed Primary A.0. Log Sheets for the period of November 12-16, 1982, indicated that the following log points were not recorded (left blank) due to WD-TI-300 being inoperable:
WG Surge Drum Temperature
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No.1 Test Tank Temperature No. 2 Test Tank Temperature
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Waste Holdup Tank Temperature
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The following Primary A.0. Log Sheet points are being recorded utilizing the temporary meters located within the cabinet:
Primary Drain Collecting Tank
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Activity Dilution Decay Tank
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Evaporator Ternperature
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The inspector noted that hourly opening of the cabinet is required to monitor the temporary meters. This potential exposure of the operators to the energized components within the cabinet and the use of temporary uncalibrated meters, which are suspended with-in an energized cabinet, is undesirable in a personnel safety and instrument reliability sense.
It appears the licensee has not completely documented the instrument as inoperable, in that no design change was made for use of the terrporary meters and no pro-visions were made for their calibration.
Primary Log Sheets have not been revised nor notations made on the same to indicate that the temperature points presently out-of-service are not required.
Licensee approved procedure AP-0200, Design Changes and Alterations, Rev. 6, requires in Attachment A, Plant Alterations, that changes to Non-Nuclear Safety (NNS) Systems, components, and/or structures not requiring quality assurance, which are initiated and written by the plant, bc processed as a Plant Alteration (PA). AP-0200, Attachment A, section A.4. also requires that Plant Alterations which render the structure, system or component unlike their description, including drauf ngs or figures in the FHSR, shall require a sa.fety )
evaluation. This requirement is consistent with 10 CFR 50.59(a)(1 which allows the licensee to make changes in the facility as described in the safety analysis report, unless the change involves an unreviewed safety question, and requires in section (b) that records of these changes be maintained including a written safety evaluation which pro-vides the bases for the determination that the change does not in-volve an unreviewed safety question. The licensee failure to document changes to WD-TI-300, Tank and Evaporator Temperature multi-point indicator, described in the Yankee NPS Final Hazards Sumary Report (FHSR) Chapter 209, Drawing 9699-RM-41F, by utilizing a Plant Alteration (PA) and the failure to conduct ~a safety evaluation of the changes, in accordance to 10 CFR 50.59 requirements constitutes
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an apparent violation (50-29/82-12-01).
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4.
The inspector reviewed licensed operator actions resulting from the discovery of the gaseous release. The Shift Supervisor Log indicates i
that at 1:30 a.m. on November 16, 1982, a plant waste gas release path was discovered and WD-V-837A was shut to terminate the release.
Discus'sions with the on-duty Shift Supervisor indicated that following
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l recognition that an apparent release was in progress the valve WD-
V-837 was found fully open, it was determined that this provided a release path from the Waste Gas system directly to the plant ventilation l
system and allowed the covergas system to bleed down into the
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ventilation header, through the stack filter system, and discharged via stack elevated release path. The Shift Supervisor consulted the Yankee NPS Emergency Plan implementing procedures and determined that no classifiable emergency existed based on the criteria of approved procedure OP-3300, Classification of Emergencies, Rev. 2, Discussions with the operator also indicated, that at that time, he failed to recognize that the release constituted a significant event as categorized by 10 CFR 50.72 and required notification to the NRC Operations Center as soon as possible and in all cases with-in one-hour. This fact was recognized during a phone conversation with the inspector at approximately 7:35 a.m. on November 16, 1982,
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at which time the Plant Superintendent promptly completed the notifi-l cation requirement via the control room RED PHONE at approximately
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7:40 a.m.
Failure to notify the NRC Opeiations Center as soon as possible and in all cases within one-hour of significant events, as classified in 10 CFR section 50.72, constitutes an apparent violationofNRCrequirements'(50-29/82-15-02).
The inspector verified that the equirements of 10 CFR 50.72 are implemented by 0P-MEM0 No. 2A-1s dated June 11, 1982 and the on-duty shift supervisor had received training in this area during annual requalification training conducted on May 19, 1982.
5.
The licensee reviewed the circumstances which resulted in WD-V-837A l
valve (normally closed) being'mispositioned. The valve was white l
tagged in the shut position per Tagout YAEC No. 820127 on November i
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l 8, 1982. On November 13, 1982, at approximately 3:00 p.m., the
white tag was removed and the valve verified closed. The valve was then found fully open the morning of November 16, 1982 and
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ordered shut by the Shift Supervisor to terminate the release.
No record of valve operation between November 13 and November 16, 1982 can be found. The licensee has subsequently locked WD-V-837A shut.
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The inspector had no further comments.
B.
Cycle XVI Reactor Startup Operations
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1.
Preparations for Startup The inspector witnessed portions of the following licensee activities during the week of December 6, 1982, in preparation for reactor startup and physics testing:
(a) Testing and calibration of the reactivity computer per OP-7106, Reactivity Computer klibration.
(b) Establishment and completion of prerequisites for reactor criticality were reviewed by the inspector for conformance with Technical Specifications requirements and the following plant procedures as applicable:
-- OP-4611, Pre-Critical Checkoff
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-- OP-2103, Reactor Startup and Shutdown
-- OP-1701, Core XVI BOL Zero Power Physics Test
-- OP-2100, Plant Startup From Cold Shutdown
-- OP-2101, Plant Startup From Hot Standby or Hot Shutdown No inadequacies were identified.
2.
Reactor Startup and Core XVI BOL Zero Power Physics Test On December 10, 1982 the inspector witnessed the initial criticality of Core XVI conducted per OP-2103, Reactor Startup and Shutdown, and OP-1701, Core XVI BOL Zero Power Physics Test. The reactor was critical at 8:53 a.m. at a calculated boron concentration of 2013 ppm. The reactor scrammed at 9:00 a.m. due to a high Startup Rate indication caused when switching from source range to intermediate range nuclear instrumentation; subsequent operations resulted in additional reactor scrams. The reactor was again critical at 3:21 a.m. on December 11, 1982 following repairs to intermediate range (IR) channels per MR 82-1124. At 5:20 a.m. an incore instrumentation detector alarm was received and at 5:50 a.m. detector path K-6 was valved out and the instrument tube capped, an ENS call was made to NRC Operations Center at 6:45 a.m.
On December 13, 1982 at 2:25 a.m. a second incore instrumentation detector leak alarm was received, detector path A-5 was valved out and capped at 2:50 a.m. and an ENS report to NRC Operations Center was made at 5:50 a.m.
The inspector reviewed Yankee Technical Specification (T.S.) 3.4.5.
2.e. which states that Main Coolant System leakage shall be limited to a maximum of two leaking incore detection system thimbles which are valved off and not plugged. On December 16, 1982 the inspector reviewed Technical Specification Interpretation No. 28, a memorandum to H. A. Autio from L. H. Heider which interpreted T.S. 3.4.5.2.e.
as:
Plant operation in Mode 1 with more than two leaking incore detection system thimbles may continue provided the thimble is isolated by valving out-of-service and plugged downstream of the isolation valve. The inspector noted to site management on December 16, 1982 that this interpretation was contrary to the conclusion of a safety evaluation conducted by the licensee in conjunction with Plant Design Change Request (PDCR) 77-12, which was utilized to plug flux mapping path J-4 and D-3 in December of 1978. Specifically, Attachment 18, Safety Evaluation to PDCR 77-12 states: The present isolation valves for the flux mapping path tubing are located a considerable distance from the main coolant pressure boundary at the reactor vessel being mounted on the Transfer Device Frame situated on the charging floor.
Disconnecting the leaking flux mapping tube at the first fitting above the conoseal and installing a plugged fitting in its place will
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move the main coolant pressure boundary closer to the reactor vessel, thus precluding uncontrolled leakage from the primary system that could result from a missile damage to the tubes.
The inspector noted to plant management that the plugs installed on incore detection thimble tubes at positions K-6 and A-5 constituted a change to a Safety Class 2 system described in the approved Final Hazards Summary Report, Chapter 107.
The inspector also noted that tne change to the system had been accomplished without the issuance of a safety evaluation and an approved PDCR required by 10 CFR 50.59 and licensee procedure AP-0200, Design Changes and Alterations; and, that the work performed was not documented by an approved Maintenance Request as required by AP-0205, Maintenance Request. The above constitutes a second example of an apparent violation of NRC requirements as previously noted in section 6.A.3.
of this report (50-29/82-15-01, second example).
The licensee acknowledged the inspectors findings and subsequently issued approved maintenance requests MR 82-1159 and 82-1160 to document the plugging of K-6 and A-5 thimble tubes. The issue of the technical adecuacy of Yankee NPS T.S. Interpretation No. 28 remains unresolvec pending NRC-NRR review of the basis for T.S. 3.
4.5.2.e. requirements (UNR 50-29/82-15-03).
The licensee subsequently completed OP-1701, Core XVI BOL Zero Power Physics Test, on December 14, 1982 with 3 of the 7 physics test criteria exceeded. The licensee's activities in regard to performance of OP-1701, Core XVI BOL Zero Power Physics Test, and OP-1702, Core XVI Zero to Full Power Physics Test were monitored by the Senior Resident and two Region-based Specialist Inspectors.
Findings in this area are documented in IR-50-29/82-16.
At the conclusion of the inspection period the reactor was critical, operating at 95% of full reactor power with the unit on the grid
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producing 175.9 Megswatt Electric (MWe).
7.
Review of Events Requiring One Hour Notification to the NRC The circumstances surrounding the following events requiring prompt NRC (one-hour) notification via the dedicated (ENS-line) were reviewed. A suninary of the inspectors review findings follows or is documented else-where as noted below:
-- About 1:30 a.m. on November 16, 1982, a cover gas system leak was discovered resulting in an uncontrolled, unplanned gaseous release via the plant stack.
The NRC was notified via the ENS-line at 7:40 a.m.
Circumstances surrounding this event are described in section 6. A.1. of this report.
-- During the reactor startup from Cycle XV-XVI Refueling Outage several reactor scrams were initiated by the reactor protection system. The licensee informed the NRC Operations Center of these events via the ENS-line. These events are described in section 6.B.2. of this rep "+.
-- On December 11,1982, at 6:45 a.m. and on December 13, 1982, at 5:50 a.m., the licensee reported that flux thimble detector paths K-6 and A-5, respectively, had indicated leakage and were isolated.
Circumstances surrounding this event are describ:.4 in section 6.B.2, of this report.
8.
Review of Licensee Event Reports (LERs)
LERS submitted to NRC:RI were reviewed to verify that the details were clearly reported, including accuracy of the description of cause and adequacy of corrective action. The inspector determined whether further information was required from the licensee, whether generic implications were indicated, and whether the event warranted onsite followup. The following LERs were reviewed.
LER No.
Date of Event Date of Report Subject 50-29/87-37 10/16/82 11/15/82 TV-204 Failed Type "C" Test 50-29/82-39 11/06/82 12/06/82 Unplanned De-energization of Emergency MCC No. 1 50-29/82-40 11/12/82 12/10/82 Loss of Power to Vital Bus Distribution Panel No inadequacies were identified.
9.
IE Circular Followup For the following IE Circular sent to the licensee for info.mation, the inspector ascertained whether the following actions were taken by the licensee:
The Circular or Bulletin was received ' by licensee management, a review for applicability was performed, and for Circulars or Bulletins applicable to the facility, appropriate corrective actions have been taken or are scheduled to be taken.
-- IE Circular No. 81-12 Inadequate Periodic-Test Procedure of PWR i
Protection System, July 22, 1981. The inspector verified that the licensee received the IEC on July 27, 1981 and routed the document to site staff. A review of the information by the licensee indicated that the hardware concerns did not apply to Yankee NPS, in that the Reactor Trip Circuit Breakers were not provided with undervoltage trip devices.
No inadequacies were identified.
10.
Plant Information Report (PIR) Reviews The inspector reviewed PIRs prepared by the licensee per AP-0004, Plant Information Reports. The inspector determined whether the conditions were reportable as defined in the Licensee Event Reports reporting require-ments section of the Technical Specification (TS) and that the licensee's system of problem identification and corrective action is being effectively utilized. The following PIRs were reviewed:
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PIR No.
Occurrence Date Report Date Subject 82-10 9/11/82 12/17/82 PR MOV-191 Missing Bonnet Bolt 82-14 9/23/82 10/20/82 Release from Waste Gas Header Loop Seal 82-15 10/14/82 11/12/82 SI System Spill Out of CS-V-655
82-16 10/14/82 11/12/82 Fire in Turbine Building 82-18 11/16/82 12/14/82 Waste Gas Surge Drum Release 82-19 11/15/82 12/15/82 Spread of Contamination While Cleaning the Incore Z Frame Tubes 82-21 12/08/82 12/27/82 No. 1 VC Booster Pump Failure During OP-4637 Surveillance Testing Except for the following the inspector had no further coments.
-- PIR 82-10 followup is discussed in section 2.b. of this report.
-- PIR 82-14 followup is discussed in section 2.d. of this report.
-- PIR 82-18 events are discussed in section 6.A. of this report.
No inadequacies were identified.
11.
Onsite Review Comittee On the following dates the inspector observed a meeting of the Yankee NPS onsite review comittee to ascertain that the provisions of Technical Specification 6.5.1 were met.
-- December 15, 1982, PORC Meeting 82-88 conducted to review Cycle XVI low power physics test data.
No inadequacies were identified.
12.
Unreso*, ad Items Unresolved items are matters about which more information is required to determine whether they are acceptable. An unresolved item is discussed in paragraph 6.B.2 of this report.
13.
Management Meetings During the inspection period the following management meetings were conducted l
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or attended by the inspector as noted below:
-- The inspector attended an exit meeting on November 19, 1982, conducted by a region-based specialist at the conclusion of IR-50-29/82-14 Refueling-Health Physics reviews, on-site inspection.
-- The inspector attended an exit meeting on December 20, 1982, conducted by region-based specialists at the conclusion of IR 50-29/82-16, Cycle XVI Start-Up Test reviews, on-site inspection.
-- At periodic intemis during the course of the 50-29/82-15 inspection period meetings N5re held with senior facility management to discuss the inspection scope and preliminary findings of the resident inspector.
Specifically, the apparent violation of NRC requirements identified on page 10, sectic: 6.A.3 and page 13, section 6.B.2; and the apparent violation of NRC requirements identified'on page 11, section 6.A.4 were discussed with the Plant Superintendent during meetings held on November 22, December 17, and November 17, 1982 respectively. A summary of findings was alsc provided to the licensee on January 5,1983 at the conclusion of the report period (see Paragraph I for attendees).
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