IR 05000259/1981035
| ML20040G911 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/18/1981 |
| From: | Cantrell F, Chase J, Paulk G, Sullivan R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20040G891 | List: |
| References | |
| 50-259-81-35, 50-260-81-35, 50-296-81-35, NUDOCS 8202160610 | |
| Download: ML20040G911 (15) | |
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UNITED STATES
NUCLEAR REGULATORY COMMISSION
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a REGION 11
101 MARIETTA ST., N.W., SUITE 3100 o,%
Y ATLANTA, GEORGIA 30303 o
Report flos. 50-259/81-35, 50-260/81-35, and 50-296/81-35 Licensee: Tennessee Valley Authority 500A Chestnut Street Chattanooga, TN 37401 Facility Name: Browns Ferry Nuclear Plant Docket Nos. 50-259, 50-260, and 50-296 License Nos. DPR-33, DPR-52 and DPR-68 Inspection at Browns Ferry site near Athens, Alabama Inspectors:
e 2 - I 7-G I R. F. Sullivan Date Signed i z-i 7-f/
p J. W. Chase Date Signed b
12-11-21 G. L. Paulk Date Signed Approved by:
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!6Y F. S. Cantrell, Section4hief, Division of Date Signed Resident and ReactorProject Inspection SUl1 MARY Inspection on October 26 - November 25, 1981 Areas Inspected This routine inspection involved 215 resident inspector-hours in the areas of operational safety, reportable occurrences, plant physical protection, main-tenance, surveillance testing, licensee action on previous inspection findings, THI action items, bulletin review, audit program implementation, reactor trips and radiological emergency plan.
Resul ts Of the eleven areas i.1spected, no violations or deviations were identified in nine areas.
Four violations were found in two areas:
(Failure to follow radiological procedures, failure to maintain drywell-suppression chamber Delta P>1.3 psig, and failure to have CN1 alann point at 3x average background, paragraph 5; failure to take hourly samples with CN1 inoperable, paragraph 6).
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8202160610 820201 i
PDR ADOCK 05000259 l
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DETAILS 1.
Perscas Contacted Licensee Employees G. T. Jones, Power Plant Superintende.nt J. R. Bynum, Assistant Power Plant Superintendent J. R. Pittman, Assistant Power Plant Supcrintendent R. T. Smith, Quality Assurance Supervisor R. G. tietke, Engineering Section Supervisor A. L. Clement, Chemical Unit Supervisor D. C. flims, Engineering and Test Unit Supervisor A. L. Burnette, Operations Supervisor Ray Hunkapillar, Operations Section Supervisor T. L. Chinn, Plant Compliance Supervisor fl. W. Haney, flechanical Maintenance Section Supervisor J. A. Tongue, Electrical !!aintenance Section Supervisor R. E. Burns, Instrument Maintenance Section Supervisor J. E. Swindell, Field Services Supervisors A. W. Sorrell, Supervisor, Radiation Control Unit BFN R. E. Jackson, Chief Public Safety R. Cole, QA Site Representative Office of Power Other licensee employees contacted included licensed senior reactor operators and reactor operators, auxiliary operators, craftsmen, technicians, public safety officers, QA, QC and engineering personnel.
2.
flanagement Interview flanagement interviews were conducted on October 30, November 9,13 and 20, 1981, with the power plant superintendent and/or his assistant and other members of his staff.
The inspector summarized the scope and findings of the inspection activities.
The licensee was informed of four apparent violation identified during the report period.
3.
Licensee Action on Previous Inspection Findings (Closed) Violation 259/81-22-02.
Failure to provide appropriate personnel monitoring equipment. The inspector verified that the individuals involved in the incident had been instructed to fully describe work activities when initiating a Special Work Permit (SWP).
All carpenters attended a special health physics training class on September 14, 1981 which discussed the incident.
(Closed) Violation 259/81-22-01,296/81-22-02.
Failure to follow health physics procedures and surveillance instructions (SI).
The inspector verified that the corrective action taken by licensee as stated in the response to the violation had been performed.
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(Closed) Violation 260/81-22-01,296/81-22-01.
Failure to cycle the containment atmosphere dilution (CAD) solenoid valves, FCU-84-5 and 16, once a month as required by technical specifications.
The inspector verified that SI-4.7.G.a.a had been revised which will ensure the cycling of FCV-84-5 and FCV-84-16 if one' or more units is shutdown..
(Closed) Inspector Follow-up Item 259/78-SB-13.
Review to a:sure con-tainment spare penetrations are seal' welded as per the FSAR requirements.
The inspector performed a spot check of spare penetrations which traverse the primary containment and found no discrepancies.
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(Closed) Inspector Follow-up Item 259/79-11-01.
Tube of versilube dropped in reactor vessel. The inspector reviewed the safety evaluation prepared by TVA in regards to the tube of versilube dropped in Unit I reactor vessel and also reviewed the test data performed by General Electric on a similar tube at their test facilities. The inspectors had no further question in this area.
(Closed) Inspector Follew-up Item 259/78-25-01.
Followup on progress to
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identify all safety components requiring qualification documentation. Tnis item was identified in regards to IE Circular 78-08, Environmental Quali-fication of Safety-Related Electrical Equipment at Nuclear Power plants.
The inspectors concerns in this area are now being tracked via IE Bulletin 79-01, Environmental Qualification of Class IE Equipment, therefore this item is closed.
(Closed) Deviation 260/81-22-02,296/81-22-03.
Failure to maintain the CAD system at greater than 100 psi.
The inspectors verified that SI-2 had been revised to show a lower limit of 100 psi in the CAD storage tanks, and that signs had been posted on the CAD storage tank to remind operators to keep the tanks at greater than 100 psi.
(Closed) Inspector Follow-up Item, 259/79-13-01.
During the followup inspection on IE Bulletin 79-08 the inspector found that many of the
flechanical Maintenance Instructions (!fil) needed revision to clearly cover r.
ements for obtaining authorization from operations supervision before wo x. g on safety systems and to advise operations when works was completed so we systems could be returned to service.
The inspector has found that extensive revisions have been made to the if1I's to more clearly specify work control requirements in the individual procedures.
(Closed) Unresolved Item 259/81-18-04.
Fire barrier qualification. The inspector reviewed actions taken by the licensee to ensure acceptable quality of penetration repair. The inspector had no further question.
(Closed) Violation 259/81-18-06.
Ur. authorized access to protected area without excort. The inspector reviewed corrective action taken by the licensee. The inspector had no additional questions on visitor control requirements.
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(Closed) Inspector Followup Item 259/81-32-05.
Failure to complete EMI 51, quarterly inspection of dryuell electrical penetrations. The licensee obtained a moisture analyzer to verify nitrogen moisture content and completed the required quarterly maintenance.
(Closed) Violation 259/81-18-02, 260/81-18-04 and 296/81-03-03.
Failure to
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perfom a 10 CFR 50.59 review on changes to !!!1175.
The licensee has reviewed prior changes to 1911 75 for unreviewed safety questions and has determined that no unreviewed safety questions exists.
4.
Unresolved Items There were no new unresolved items identified during the report period.
5.
Operational Safety The inspectors kept informed on a daily basis of the overall plant status and any significant safety matters related to plant operations.
Daily discussions were held each morning with plant management and various members of the plant operating staff.
The inspectors made frequent visits to the control rooms such that each was visited at least daily when an inspector was on site.
Observation included instrument readings, setpoints and recordings; status of operating systems; status and alignments of emergencyor standby systems; purpose of temporary
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tags on equipment controls and switches; annunciator alarms; adherence to procedures; adherence to limiting conditions for operations; temporary alterations in effect; daily journals and data sheet entries; and control room manning.
This inspection activity also included numerous informal discussions with operators and their supervisors.
General plant tours were conducted on at least a weekly basis.
Portions of the turbine building, each reactor building and outside areas were visited.
Observations included valve positions and system alignment; snubber and hanger conditions; instrument readings; housekeeping; radiation area
controls; tag controls on equipment; work activities in progress; vital area controls; personnel badging, personnel search and escort; and vehicle search
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and escort.
Informal discussions were held with selected plant personnel in their functional areas during these tours.
In addition a complete walkdown which included valve alignment, instrument alignment, and switch positions was performed on the drywell leak detection instrumentation (Air Sampling).
This portion of the system monitors the air activity in the drywell for indications of prinary leakage.
The inspectors reviewed SI 4.2.E-2 (Air Sampling System) which calibrates the air sampling system for the drywell and found that the alarm setpoint was not set as required by the technical specifications. Table 3.2E of the
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technical specifications requires.that the alarm set point be set at three times (3x) average background.
The inspectors found the following set of conditions to exist:
a.
Unit 1 (2)
Alarm point Background 3x Background Particulate 500,000 cpm (1)
250,000 cpm (1)
750,000 cpm (1)
Gaseous 3,500 cpm (1)
375 cpm (1)
1,125 cpm (1)
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Unit 2 (2)
Particulate 500,000 cpm (1)
50,000 cpm (1)
150,000 cpm (1)
Gaseous 3,500 cpm (1)
100 cpm (1)
300 cpm (1)
Note:
(1) cpm = counts per minute (2) Unit 3 was shut down for refueling outage Based on the data above, the inspector determined that the alam setpoints for gaseous activity in Units 1 and 2 and particulate in Unit 2 was not set at the required value of Technical Specification 3.2.E.
On November 6,1981 the plant superintendent was infomed that failure to set the alarm for the drywell leak detection air sampling system at 3x background was an apparent violation of Technical Specification 3.2.E which '
requires that the limiting condition for operation for instrumentation that monitors drywell leak detection be set by Table 3.2.E.
Table 3.2.E states-that the air sampling system alarm will be set at 3x average background.
(259/81-35-01,260/81-35-01)
On October 29, 1981 while making a routine tour of Unit 2 control room, the inspectors noted that the drywell-supression chamber differential pressure (DP) was indicating 1.25 psid vice greater than or equal to 1.3 psid as required by Technical Specification 3.7.A.6.
The operator noting the inspectors observation immediately pumped the DP up to 1.3 psid.
The drywell-suppression chambers DP abnomal alam had not been received.
The inspectors subsequent investigation into this item revealed that the surveillance instruction (SI)-4.2.F-17, Drywell to Suppression Chamber Differential Pressure, perfomed on June 25, 1981 had set the drywell-suppression chamber abnomal alam at 1.26 psid as required by the SI. The automatic start feature for the drywell-suppression chamber air compressor (compressor which maintains the DP) was set at 1.31 psid, however; operations personnel had been operating the compressor in manual.
This set of conditions then relied on the operator to ensure the DP never went below 1.3 psid by frequent observation of his gauge instruments.
In September 1981, a new SI-4.2.F-17 was issued which set the alam for abnormal DP at 1.31 (-0+.01).
This revised SI had been perfomed for Unit 1
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but not for Unit 2 or 3.
On October 29, 1981 the licensee performed the revised SI on Unit 2 and increased the alam point to 1.31 psid.
Unit 3 was shut down on October 30, 1981 for a refueling outage.
On Hovember 6,1981, the Plant Superintendent was informed that failure to maintain.the drywell-suppression chamber DP at 11.3 psid was an apparent violation of Technical Specification 3.7. A.6 which requires the P to be maintained at r 1.3 psid.
(260/81-35-02)
On November 9,1981, at 850 a.m. the inspector toured the refuel floor to observe Unit 3 refueling operations. The inspector reviewed radiological controlled work in progress and examined special work permits for adherence to regulatory requirements. The inspector observed several workers perfoming general cleanup in a contaminated zone that had proper contami-nation area boundary markings posted.
The inspector could not find a special work permit for the cleanup work in progress. The inspector questioned the refuel floor health physics technician and job foreman about radiological control procedures being followed in the conduct of the work.
The technician and craftsman infomed the inspector that no special work permit had been issued.
The workers were partially dressed out in C-zcne clothing as directed by the foreman.
Radi] logical Control Instruction 10 requires that entry into an unknown contaminated area be. controlled by the issuance of a special work permit. The inspector notified the health physics supervisor immediately of the above event.
The plant superintendent was infomed during the weekly management meeting on November 9,1981, that the event was an apparent violation of Technical Specification 6.3.A.7.
The plant superintendent accepted the violation with no comment.
(259/81-35-03, 260/81-35-03and296/81-35-01).
6.
Reportable Occurrence The below listed licensee event reports (LERs) were reviewed to determine if the information provided met NRC reporting requirements.
The detemination included adequacy of event description and corrective actio'n taken or planned, existence of potential generic problems and the relative safety significance of each event. Additional in plant reviews and discussion with plant personnel as appropriate were conducted for the reports indicated by an asterisks.
LER NO.
Date Event 259/81-45 9-10-81 Operation of reactor at power with startup bus undervoltage relays out of service.
- 259/81-42 8-14-81 250V Battery Cracked 259/81-52 9-18-81 Failure to post fire watch during welding operation.
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- 259/81-46 9-11-81 Reactor building air monitor for ventilation exhaust found inoperable.
259/81-55 9-27-81 Diesel generator 1C start relay failed to close.
- 259/81-56 10-29-81 Main steam isolation valve closure time exceeded time specified.
- 259-81-57 10-30-81 Drywell and torus "A" H -0,
analyzer inoperable.
259/80-16 Rev. 1 9-25-81 RBCCW line break could degrade primary system integrity.
- 259/81-32 6-23-81 Various reactor pressure switches were found out of tolerance.
259/81-48 9-27 81 Recirculation pump reactor high pressure trip was found out of tolerance.
260/81-59 11-12-81 Reactor low pressure switches were found out of tolerance.
- 260/81-40 8/7/81 During reactor water level surveillance test, recir-culation pump B tripped.
- 260/81-43 9-1-81 Unit 2 start bus undervoltage relays inoperable during plant startup.
- 260/81-35 Rev. 2 11-11-81 Scram discharge instrument volume 3-gallon level switch inoperable.
260/81-48 10-19-81 APRM flow totalizer drifted high.
- 260/81-49 10-26-81 Containment H analyzer
inoperable because sample returned pump tripped.
- 260/81-51 10-30-81 Torus water level transmitters out of specification.
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- 260/81-53 11-06-81 PI-75-48 out of service because of improper valve lineup.
- 260/81-55 11-17-81 Reactor water level switches out of tolerance.
296/81-40 9-11-81
"R" factor out of specification.
- 296/81-49 10-9-81 Start relay for 3C diesel generator exceeded specifi-cation.
- 296/81-46 11-23-81 3EA Reactor H0V board 11G set coupling failure.
- 296/81-29 Rev. 1 11-23-81 Fire protection system sprinkler for reactor building found isolated.
- 296/81-50 10-20-81 3EA LPCI liG set inoperable due to slippage of retainer ring.
- 296/81-51 10-23-81 3EA LPCI MG set inoperable due to slippage of retainer ring.
- 296/81-53 10-27-81 Tours water level transmitter out of specification.
296/81-54 10-29-81 High drywell pressure switch out of tolerance.
296-81-55 10-30-81 Reactor water level switch out of tolerance
- 296/81-63 11-19-81 Drywell continous air monitor inoperable.
- 296/81-44 9-25-81 Scram discharge instrument volume 50 gallon level switch inoperable.
- 296/81-22 6-30-81 Containment cooling valve FCV-74-57 inoperable.
296/81-41 9-15-81
"R" factor out of specification l
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296/81-42 9-14-81 Main steam line radiation monitor out of specification.
296/81-43 9-18-81 Containment air monitor failed.
296/81-45 9-28-81
"R" factor out of specification.
296/81-39 Rev. 1 11-13-81 Turbine first stage permissive switch out of specification.
During the inspectors review of LER 259/81-46, Reactor Building ventilation exhaust monitor declared inoperable for 12.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, it was determined that the surveillance requirements were not satisfied for an inoperable continous air monitor (CAM). Technical Specification 4.8.B.a. requires that those effluent streams with continous air monitors have their activity and flow rate determined on an hourly basis. With the CAM inoperable this requirement is satisfied by the Chem Lab obtaining samples every hour with portable instruments.
The inspectors found by review of Chem Lab logs during the time period that CAfi-I-RM-90-250 was inoperable on August 16, 1981 the following time intervals occurred between the hourly samples:
Time Event Interval 1410 CAM declared inoperable 0 Min.
1625 first sample 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 15 min.
1800 second sample 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 35 min.
1920 third sample 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 20 min.
2110 fourth sample 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 50 min.
2344 fifth sample 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 34 min.
0048 sixth sample 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 4 min.
0142 seventh sample 54 min.
0244 eighth sample 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 2 min.
0341 ninth sample 57 min.
As shown by the above table, five samples exceeded the hourly (+ 25%) sample requirement. On November 23, 1981 the Plant Superintendent was informed that failure to take hourly samples was an apparent violation of Technical Specifi-cation 4.8.B.1.a. which requires the gross radioactivity to be determined on an hourly basis.
(259/81-35-02)
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7.
THI Action Items The following Till action items were reviewed by the inspectors during this report period; a.
II.E.4.2 (5.B) Containment Isolation Dependability Modifications NUREG-0660 required that the setpoint pressure for the signal that initates cor.tainment isolation be reduced to the minimum compatible with normal operations. HUREG-0737 which clarified the TMI action item
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stated that at setpoint of 1 psi above the operating maximum-containment pressure would be adequate.
Currently at Browns Ferry the containment isolation occurs at 12.5 psi and the maximum containment pressure during normal operation is 1.5 psi; therefore, Browns-Ferry meets the reo'tirements of this T111 Action item.
The inspector considers this item closed.
b.
II.B.4.
Training for liitigating Core Damage. The inspectors attended the training class for mitigating core damage given to the instrument mechanics and health physics personnel (2 separate classes). The inspectors found the training for instrument mechanics to be satis-factory. The training for health physics personnel was considered unsatisfactory by the inspectors for the following reasons:
1.
The total class session time was approximately 10 minutes and only a small portion of this time dealt with mitigating core damage (radiation level readings to expect on a core melt down).
2.
Tiethods of determining dose rate inside containment for measurements taken outside containment were not covered.
3.
Expected accuracy of radiation detectors at different locations was not covered.
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Methods for detecting radiation reading by direct measurements at detector output was not covered.
The requirements for the training course were based on Harold Denton's
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11 arch 28,1980 letter to all licensee which discussed training criteria for mitigating core damage.
The inspectors discussed their findings with the plant superintendent and training personnel. TVA has committed to reevaluating their training for health physics personnel and presenting it again in the near future. The inspector reminded the licensee that they had j
committed to having this training completed by January 1,1982 as documented by their response to this item dated December 23, 1980 l
This item remains open until satisfactory completion of the health physics training in mitigating core damage.
8.
Surceillance Testing Observation The inspector observed the performance of the below listed surveillance activities.
The inspection consisted of a review of the procedures for technical adequacy, conformance to technical specifications, verification of test instrument calibration, observation on the conduct of the test, the removal and return of the system to service, and the review of test data.
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a.
Surveillance Instruction 4.9.A.2.c, Battery Discharge Test Unit 3 b.
Surveillance Instruction 4.9. A.2.b, Battery Analysis Test Unit 3
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Surveillance Instruction 4.7.H, Containnent Atmosphere Monitor System -
Hydrogen Analyzers Unit 2.
d.
Surveillance Instruction 4.11.D.3, Fire Protection System Monthly Inspection, Unit 3.
In the above areas no violations or deviations were identified, however, one open item was noted in reference to item (d).
During the conduct of the fire protection system monthly inspection (S.I.4.11.D.3), it was noted that the procedure had omitted valve designations, incorrectly grouped valves under one generic title vice individual designations (i.e. shut all drain valves), and failed to correctly identify valves. Numerous errors of this type existed making the procedure usable only to a well experienced fire protection engineer.
This problem is generic for all units, although only Unit 3 was checked by the inspector.
The fire protection engineer agreed to revise the applicable surveillance instructions. This item will remain open.
(259/81-35-04, 260/81-35-04 and 296/81-35-02).
9.
Maintenance Observation The inspectors observed the below listed maintenance activities for procedural adequacy, adherence to procedure, and proper performance of the maintenance.
a.
System Maintenance Instruction 1976, Oxygen Gas Analyzer Calibration.
b.
Coupling alignment checks on 3EN LPCI motor generator set c.
Weather protection installation in residual heat removal service water pump rooms.
In the above areas, no violations or deviations were identified, however, one open item was noted related to item (b) and in described in paragraph 12.
10.
Bulletin Review Licensee action on the below listed bulletins was reviewed to determine if the evaluation and action taken was appropriate to satisfy the concerns described in the bulletin. The inspection consisted of records, drawings and procedure review and discussions with plant personnel.
The below listed bulletins are considered closed.
A.
IE Bulletin 80-20 " Failure of Westinghouse Type W-2 Spring Return to Neutral Control Switch."
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b.
IE Bulletin 81-02 " Failure of Gate Type Valves to Close Against Differential Pressure."
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c.
IE Bulletin 80-09 "Hydramotor Actuator Deficiencies."
d.
IE Bulletin 80-14 " Degradation of BWR Scram Discharge Volume Capabili ty."
e.
IE Bulletin 80-24. " Water Leakage Inside Containment".
f.
IE Bulletin 79-11, Westinghouse Circuit Breakers".
In the above area, no violations or deviations were identified.
11. Audit Program Implementation During this report period, the resident inspector performed a review of the licensee's quality assurance audit program performed by the Office of Power Quality Assurance group (0PQA). The scope of this review consisted of a review of recent audit reports for:
a.
scope of the audits b.
audit report findings and timely and appropriate corrective action taken c.
audits were performed by personnel not having direct responsibility in the area being audited.
In addition to the above, the audits specified in the technical specifications were reviewed to ensure they were bei'g performed at the frequency specified. The inspector also observed the OPQA staff perform an audit on Outage Activities (Audit report OPQA-BF-8100-03).
The performance of this audit conformed to the guidelines set forth in the TVA Topical Report (TVA-TR75-1).
In the above area, no violations or deviations were identified.
12. Low Pressure Coolant Injection Motor Generator Sets The low pressure coolant injection (LPCI) motor generator (M-G) sets were designed to provide a reliable power supply to the reactor MOV boards that feed the motor operators of the LPCI injection valves, the recirculation pump discharge valves, and the RHR pump minimum flow bypass valves.
The design involves the use of class 1E M-G sets as isolation devices between the auto transfer feature of the 480V reactor MOV boards and the divisional 480V chutdown boards. Redundant M-G sets are provided between the divisional 480V shutdown boards and each 480V reactor MOV board.
For example, for BFNP-1, 480V Reactor MOV Board ID will be normally supplied by M-G set 1DN, connected to 430V Shutdown Board 1A, with alternate supply to be available from M-G set IDA connected to 480V Shutdown Board 18 and etc.
The M-G sets are designated as Class 2E equipment and are designed to seismic Category 1 standards.
Each M-G set is sized to accept the load requirements of the valve operators at any time during an accident initiating event.
(i.e.LOCA). Although only one M-G set will normally i.--
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supply power to each 480V reactor 110V board both M-G sets are run at all times to assure readiness of the alternate li-G set to accept the load on auto transfers. Thus, the insertion of isolation M-G set between the Reactor f10V boards and the shutdown boards provides added assurance of independence between redundant divisions of power supply.
The LPCI li-G sets are installed on Unit 1 and 3 only.
Technical Specifi-cations for each unit require that if one IbG set is out of service, a 7-day servicing period is allowed.
Having two M-G sets out of service requires the affected unit to be in cold shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The LPCI li-G sets became operational for Unit 3 in January 1981 and for Unit 1 in September 1981.
Since the installation of the LPCI 11-G sets the inspectors have been trending the operational reliability of the system.
During the management meeting held on flovember 20, 1981 the inspectors expressed concern to the plant manager that the continued series of component failures and/or deficiencies may represent a significant condition adverse to quality and reliability of these safety system component power sources.
The component failure descriptions and deficiencies are listed on the following table 1 and are derived from licensee event reports, inspector observation, technical au tysis reports, and discussions with various plant engineers.
The problems with the H-G sets fall into two primary categories; voltage regulator and mechanical coupling failures. The licensee has been evaluating various failure mechanism during the past six months, but to date no satisfactory corrective measures have curtailed the frequency of failures.
Technical evaluation by the licensee has attributed coupling failures to lubricant deterioration, coupling seal failure, and shaft misalignments.
The manufacturer has analyzed the three voltage regulator failures to be due to the same failed component, a feedback capacitor.
Failure of the component causes the supplied M-G set voltage to vary f_5 volts, creating a constant output voltage instability.
The licensee operated the 3EA LPCI f1-G set for three months with a failed feedback capacitor before replacing regulators. The licensee considered the system fully operational during the three month period.
The inspectors are continuing their review of this item.
The licensee has established several corrective measures to reduce 11-G set l
coupling failures including monthly coupling lubrication, daily vibrational checks, and weekly visual checks. The daily vibration checks have not been as effective as expected since failures occur randomly and unpredictable.
The coupling noise spectrum may be satisfactory early 17 the day and fail by second shift. Additionally a new high performance grease (Koppers KHP)
is being used.
Corrective measures to alleviate the poor performance record of the LPCI f1-G sets is being pursued by licensee design groups.
The licensee was informed that evaluation of the coupling and voltage regulator failures will remain open for further review (259/81-35-05, 260/81-35-05and296/81-35-03)
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TABLE 1 Chronological List of Problems and Failures on New LPCI HG Sets installed on Units l'and 3 Unit Date LER #
Problem 3EN, Bearing failure on motor
11-16-81
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side of flywheel
10-3-81 81-56 3EA, Generator bearing failure, bearing replaced
09-27-81 81-52 3EA, Voltage regulator unstable, regulator replaced
09-25-81 81-51 3EA, Coupling came apart
09-22-81 81-50 3EA, Coupling camp apart
09-05-81 81-46 3EA, Coupling sheared
,
3EA Voltage regulator replaced
09-04-81
voltage unstable
,
02-24-81 81-11 3DA, Coupling damaged due to misalignment and loss of lubricant 1EN, High coupling vibration
11-24-81
on generator end 1EN, Leaking seal high
11-13-81
i vibration, coupling replaced 1EN, High coupling vibration
10-09-81
on generator end 1EN, High coupling vibration on
10-02-81
generator end
09-17-81
1EA, Voltage regulator failed, regulator replaced NOTE:
Additional Facts:
M-G set manufacturer:
Louis Allis Company Running RPft:
1800 RPM
- _ - - _ - - - - - - _ - -
,
,
..
.
.
Coupling Type:
Direct coupled Waldron 2h-P Power line Coupling Grease Recommended by Vendor:
Chevron SR-1 Voltage and Regulator flanufacturer:
Basler Electric Company 13.
Reactor Trips The inspectors reviewed activities associated with the below listed reactor trips during this report period. The review included determination of cause, safety significance, performance of personnel and systems, and corrective action. The inspector examined instrument recordings, computer printouts, operations journals entries, scram reports and had discussions with operations, maintenance and engineering support personnel as appro-pria te.
On October 30,1981, Unit 3 was manually tripped from 37% power at 11:45 p.m.
to begin a scheduled refueling and maintenance outage.
Safety systems performed as designed. No emergency core cooling was called upon nor were any main steam relief valves actuated.
No violations or deviations were identified within the area inspected.
14. Plant Physical Protection During the course of routine inspection activities, the inspectors made observations of certain plant physical protection activities.
These included personnel badging, personnel search and escort, vehicle search and escort, communications and vital area access control.
One violation was noted in this area and will be included in report 259/260/296/81-34.
15.
Radiological Emergency Plan At 2:30 a.m. on October 31, 1981, Browns Ferry declared an alert (IP-3) upon loss of the power supply to two annunciator panels in Unit I control room.
The panels (XA-55-8 B and XA-55-8C) provide information on the status of electrical power to various buses and boards. fiost of the items on the panels are common to more than one unit an'd the lost information was still available on similar panels or. the Unit 2 side of the shared control room.
On the few components which were not shared, the informaticn on electrical status was available at local panels.
There was no total loss of infor-mation en systems important to safety. The problem was found to be a failure of the 48VDC/120VAC inverter which supplied the AC power to the two panels. Repairs were made and the alert was lifted at 4:45 a.m.
Notifications required by the emergency plan were made including an infor-mation call to one of the resident inspectors. The inspectors findings were that the licensee responded in accordance with the emergency plan and no violations or deviations were identified.
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