IR 05000327/1989012

Insp Repts 50-327/89-12 & 50-328/89-12 on 890406-0505. Violation Noted.Major Areas Inspected:Radiation Protection, Safeguards & Operational Safety Verification,Including Control Room Observations & Operational Performance

ML20245A273
Person / Time
Site:	Sequoyah
Issue date:	05/24/1989
From:	Brady J, Jenison K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), Office of Nuclear Reactor Regulation
To:
Shared Package
ML20245A264	List: IR 05000327/1989012 ML20245A262
References
50-327-89-12, 50-328-89-12, NUDOCS 8906210182
Download: ML20245A273 (19)

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION

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[. g .n - -  REGION li 101 MARIETTA STREET, g: j

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ATLANTA, GEORGI A 30323

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Report Nos.: 50-32"/89-12,50-328/89-12' i

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Licensee: Tennessee' Valley Authority  ; 6N 38A Lookout Place l 1101 Market Square Chattanooga,.TN 37402-2801 Docket Nos.: 50-327 and 50-328 License Nos.: DPR-77 and DPR-79 Facility Name: Sequoyah Units 1 and 2 Inspection Cond ted: Apri) 6,1989 thru May 5,1989 l

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Inspector: /  !# b K. Jenisfn, Senior Resident-Inspector Date Signed q \  ! l Inspectors: P. Harmon,. Senior Resident Inspector 4 ! D. Loveless, Resident Inspector ] y

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Accompanying Personnel: Peter A. Balmain  ; Binoy B. Desai l Approved by: / / IAd 5/df/97 g.Brady, Actin 34hief, Project Section 1

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Date Signed TVA Projects Division-Summary Scope: This announced inspection involved inspection effort by the Resident Inspectors in the area of operational safety verification' including control room observations, operations performance, system lineups, radiation protection, safeguards, and housekeeping inspections. Other areas inspected included maintenance observations, surveillance testing observations,. review of' previous inspection findings, ' follow-up of events, review of licensee identified items, and review of inspector follow-up' item Results:

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The areas of Operations and Surveillance appeared to be adequate, improving and fully capable of supporting current. plant operation Control . room operators were professional and the activities- they conducted .were well managed. .An example of these control room activitie.Cincluded the. implementation of the .feedwater control corrective' actions. A public meeting was held to discuss the'three reactor trips that occurred as_airesult of'feedwater control problem "7 PDR ADDCK89052705000J27 *' G' PNU L

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        'l The area of maintenance work planning had some licensee identified weaknesse However, the overall area of maintenance was adequate during this period. Line management involvement was acceptable and QA involvement in current issues and operations issues was aggressiv No violations or deviations were identifie Two non-cited violations were identifie NCV 327,328/89-12-05, Diesel Generator Maintenance (paragraph 3).

NCV 327,328/89-12-06, Feedwater Reactor Trips, three examples (paragraphs 9.e, 9.f, and 9.g).

Three unresolved items and one inspector follow up item were identifie URI 327,328/89-12-01, Licensed Power Indication, (paragraph 2.a) : URI 327,328/89-12-02, BIT Recirculation, (paragraph 9.a) IFI 327,328/89-12-03, RCS Identified Leakage, (paragraph 9.d) i

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URI 327,328/89-12-04, AFW Maintenance, (paragraph 4.b) l

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REPORT DETAILS . j

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1. ~ Persons Contacted

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J. Bynum, Vice President, Nuclear Power Production d

   *J. LaPoint, Site Directo *S. Smith, Plant Manager T. Arney, Quality Control Manager-
   *R.-Beecken, Maintenance' Superintendent
   *M. Cooper, Compliance Licensing Manager D. Craven, Plant Support. Superintenden *S. Crowe, Site Quality Manager
   *T. Flippo, Quality Assurance Manager      ,

R. Fortenberry, Technical Support Superviso l J. Holland, Corrective Action Program Manager l J. Patrick, Operations Superintendent > R. Pierce, Mechanical Maintenance Supervisor M. Burzynski, Site Licensing Staff Manager l A. Ritter,. Engineering Assurance Engineer ~l R. Rogers, Plant Support Superintendent 1 M. Sullivan, Radiological Controls Superintendent- 'l S. Spencer, Licensing Engineer 'I

   *C. Whittemore, Licensing Engineer      1 NRC Employees j
   *J. Brady, Acting Chief, Projects Section 1, TVA Projects Divisio ;
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   * Attended exit interview
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Acronyms and initialisms used in this report are listed in the last paragrap ' Operational Safety Verification (71707) Control Room Observations L The inspectors conducted discussions with control room operators and verified that proper control room staffing was maintaine The-inspector also verified that access to the control .. room was properly controlled, and that operator behavior was commensurate with plant , configuration and ' plant. activities' in. progress .and with . on goin j control room operations. The operators ware observe <J adhering; to i appropriate and approved procedures, including Emergency Operating Procedures, for on going activities. Operator activities.on Unit 1

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i included adequate response and followup to indications of possible reactor vessel flange o ring leakage. The operator actions and i

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review regarding this indication were deemed to be appropriate. It was determined. that the leakage indications were the result of a possible' thermocouple failure and a maintenance request was written to replace the applicable thermocoupl . . l Additionally, ~the frequency of visits to the control room by j operations line management was observed and found to be. acceptabl First line, s.econd line and Duty Operations managers were observed in the control room and appeared to be engaged in activities ' that - supported shift operations and the SO The inspector also verified that the licensee was operating the. plant in a normal plant configuration as required by TS. When .~ abnormal conditions existed, operators were found to be complying with ~the-appropriate ' LCO action statement Upon inspection of LCO action . statements, it was noted that the units _ were in .a large number of action statements requiring non-emergent actions. Licensee manage : q ment was well aware of these LCO ' action items and this issue was ' discussed on a daily basis. The licensee's TS interpretation of one action statement involving the TS 3.4.3.2 OPERABLE condition of the I PORV block valves was questioned. The licensee's interpretation .was l conservative -and this TS interpretation was later implemented by TS i amendment 101 which was issued on Apri1 ~ 3,1989. Even though the TS l

  - interpretation appeared to be conservative, it took the place of a-formal licensee 10 CFR 50.59 safety evaluation. The' adequacy of,- the !

licensee's safety evaluation program implementation is a current issue of discussion between the licensee and NRC ' managemen At least two of the licensee's staff' organizations (QA and NSRB)~have , identified weaknesses with the licensee's 10 CFR 50.59 safety evaluation processe The inspector verified that RCS leak rate calculations were performed and that the calculated leakage rates were within the'TS limits. In the specific case of the Unit-1 reactor vessel flange leak indication, total RCS leakage was determined-to be-less than 0.4 gp The inspectors reviewed control room indications associated -with safety limits and determined that none were exceeded. - The specific issue of the licensee's definition of IDENTIFIED LEAKAGE is discussed-as IFI 327,328/89-12-03, in paragraph 9.e of this. repor The inspectors sampled instrumentation and recorder traces for indication of abnormalities and verified the ' status of selected control room an.nunciators to ensure that co.ntrol room operators understood the status of the plant. Panel indications:were reviewed p for the nuclear instruments, the emergency power sources, the safety ! parametM display system and the radiation monitors to ensure-operability and operation within TS limit Control rod insertion limits were observed as specified in the TS. During this review, the use of nuclear instruments vs the use of P250 computer, program U1118, to measure thermal power _was questioned. Several ' issues' . were identified. These issues involved the following evaluations: _ _ - _ _ _ _ _ - _ _ _ _ _ _ .

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  (1) An evaluation- to determine which of the two power indications s

should be used to comply with the license condition which limits-thermal. power to 3411 M ) I (2) An evaluation of a, specific eight hour . period where ' thermal' power, as indicated by the U1118 computer program,. indicated a eight hour average in-excess of 3411 M (3) An evaluation -of the same eight hour-period'for compliance with ' axial. flux difference TS requirements-stated in TS.3. (4) An evaluation of a specific period of 52 minutes where thermal l power exceeded 100.75%- Resolution of these issues' will. be tracked u'nder ' URI 327,328/

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89-12-01. The licensee's current operating, philosophy with regard to this . issue appears' to be adequate after corrective ' actions wer taken. The licensee is currently operating using indications from the highest reading nuclear instrument to . limit . power to 100%. In-addition, the licensee is considering the use of' an eight hour average thermal power at or below 3411 MW thermal as measured by the' U1118 program. Based on discussions with.NRR and. Region II technical specialists, both. methods appear.to be. acceptable to the NRC staf No violations or deviations were observe Control Room Logs 1

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The inspectors observed control ' room operations and revie.>ed ) applicable logs including the shift logs, ' operating orders, night 6 order book, clearance hoid order book, r.nd configuration log to

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obtain information concerning operating trends and activities. The TACF log was reviewed to verify that the use of jumpers and lifted leads causing inoperabilities were clearly noted and understood. The licensee is actively pursuing corrections to ' conditions requiring TACFs through its design change progra No issues were identified with these specific log Plant chemistry reports were reviewed to confirm steam generator tube integrity in the secondary side and to verify that primary : plant chemistry was within TS limits. The implementation of the' licensee's sampling program was. observed. Plant specific monitoring . systems including seismic, meteorological and fire detection indications were reviewed for operabilit ! No violations or deviations'were observe ; _ - _ _ _ _ _ _ _ _ _ _ _ _ _

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4- , , ECCS System Alignment The inspectors . walked down portions of the following safety-related: .;

- systems to - verify operability, flow path, heat sink, water' supply,
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power supply, and proper ~ valve ~and breaker alignment: ,

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RHR (unit l'and 2) SI (unit 2).

In ' addition, the inspectors verified that selected containment isolation valves were aligned correctl No deviations or! violations were identifie Plant Tours Tours of the diesel generator, auxiliary, control, and turbine ] buildings, and exterior _ areas were - conducted to observe - plant' ' equipment - conditions, potential fire hazards, controlf of ignition sources,. fluid : leaks, excessive vibrations,. missile hazards- and plant housekeeping and cleanliness conditions. ~ The plant was observed to ' be clean and in adequate condition. . The inspectors L verified tha maintenance work orders had been submitted as . required and that followup activities and prioritization 'of work was -accomplished by

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the license The inspector visually inspected major components for leakage, proper lubrication, cooling water supply, and general conditions which could prevent fulfillment of their functional requirement The inspector observed shift turnovers and determined that necessary information concerning the plant systems statu's was addresse Turnovers were detailed, and attended by ;the outage l manager and usually the operations duty manage No violations or deviations were observe Radiation Protection The inspectors.. observed HP practices and verified the implementation: of radiation' protection contro1 On a regular basis, RWPs were i reviewed and specific work activities were monitored .to ' ensure that' activities were being - conducted in accordance with the ' applicable .-l RWP Workers. were observed for proper frisking z upon '. exiting 'l

' contaminated areas and the radiologically controlled area. Selected    !

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l radiation protection instruments were verified operable and calibration frequencies were reviewed. The following RWPs were reviewed in detail: RWP 89-20395 Seal Table RWP 89-00155 CDWE RWP 89-20409 Accumulator Room #3 The inspectors questioned the inclusion of a storm drain located on l the rear of the control building in an RCA roped off area. The HP supervisor stated that the storm drain should not have been included ! within the RCA and had the RCA boundary moved so that it was no longer include The inspector had no further question I No violations or deviations were identified, f. Safeguards Inspection , i In the course of the monthly activities, the inspectors included a j review of the licensee's physical security program. The performance i of various shifts of the security force was observed in the conduct ! of daily activities including: protected and vital area access ! controls; searching of personnel and packages; escorting of visitors; badge issuance and retrieval; and patrols and compensatory post ] In addition, the inspectors observed protected area lighting, and protected and vital area barrier integrity. The inspectors verified interfaces between the security organization and both operations and maintenanc Specifically, the resident inspectors:

 (1) observed security drills (2) observed security program response drill (3) inspected security during outages (4) reviewed licensee security event report (5) visited central alarm station (6) verified onsite/offsite communication capabilities First line management and upper level management was very responsive to the inspector's question The activities observed involving security officers were conducted in a professional manne No violations or deviations were identifie g. Hold Orders The inspectors reviewed the following H0s to verify compliance with AI-3, Revision 38, Clearance Procedure, and that the H0s contained adequate information to properly isolate the affected portions of the system being tagged. Additionally the inspectors ensured that the licensee verified that the required tags were installed on the affected equipmen The following tags were inspected to deterniine their age and that they did not impact plant operations:

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Hold Order Equipment 2-89-002 Incore Probe 2-89-003 Battery Charger No violations or deviations were identifie Conditions Adverse to Quality The inspectors reviewed selected items to determine that the licensee's problem identification system as defined in AI-12, Corrective Action, was functioning. CAQR's were routinely reviewed, for adequacy in addressing a problem or event. Additionally, a sample of the following documents were reviewed to determine that the program was functioning:

 (1) Work Requests (2) Potential Reportable Occurrences (3) Radiological Incident Reports (4) Test Deficiencies (5) Problem Reporting Documents (6) Licensee Event Reports Issues appeared to be adequately resolved in a timely manne The lower tier corrective action administrative feeder systems have a large number of issues being processed. However, no large backlog was identified, and the upper tier corrective action documents were the subject of upper level site management discussion No violations or deviations were observe Positive trends were identified in the operational safety verification are General conditions in the plant were adequate and continue to improve. Control of general reactor operations and specifically startup operations has improved and is being strongly supported by Sequoyah line managemen Radiation protection and security are adequate to continue two unit operation . Surveillance Observations and Review (61726)

Licensee activities were directly observed / reviewed to ascertain that surveillance of safety-related systems and components was being conducted in accordance with TS requirement The inspectors verified that: testing was performed in accordance with adequate procedures; test instrumentation was calibrated; LC0; were met; test results met acceptance criteria requirements and were reviewed by personnel other than the individual directing the test; deficiencies were - identified as appropriate, and any deficiencies identified during the testing were properly reviewed and resolved by management personnel; and _ ____

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system restoration was adequat For . completed tes.ts, the . inspector '. verified that testing frequencies were: met and tests. were ' performed by qualified individual ! No trends were identified in the area of surveillance performance during; this . inspection perio The area of surveillance scheduling and management was observed to be adequate and the completion of TS surveillance requirements was discussed routinely at the site director and plant manager leve The following' activities were observed / reviewed: SI-102M, Diesel Generator Mechanical Inspections SI-7, Electrical Power Systems:. Diesel Generators' MI-4.2.6,.Two Year Preventive Maintenance.of Diesel Engines This activity was performed in support of SI-102M and.SI-7. I involved the performance of preventive- and ' corrective maintenance on the diesel generators and.. supporting L systems.~ ! Maintenance performed- on valves in the diesel air start system- 1 and diesel lubrication oil system was observed and determined to not include detailed configuration con _ trol with.' respect to.~ seals and gaskets. .However, specific technician training ' had been accomplished as part of ' routine craftsman'. training _ and a i detailed drawing was provided in the maintenance package. , In j addition, the ' orientation of the lubrication oil check valves - l was not controlled during reinstallatio This - was. resolved with a procedure change to MI 4. Finally, torque: values for jointed connectior.s on the lubrication and ' air start '11nes was not addressed by this procedure. ,The licensee issued PRD SQP' 890279P to evaiuate an action plan for ' addressing ' torque requirements on diesel generator bolted connection Thes issues collectively constitute a non cited violation (NCV)- 327,328/89-12-05. The corrective actions' for these issues .are adequate and the violation is not cited because the criteria specified in Section V.G of the' Enforcement Policy- were satisfied. Therefore, NCV 327,328/89-12-05 is close . Monthly Maintenance Observations and Review (62703)- Station maintenance activities on safety-related systems and components were observed / reviewed to ascertain that they were conducted in accordance  ; with approved procedures, regulatory guides, industry. codes 'and standards, ' and in conformance with The following items were considered during this review: LCOs were met while components or systems were . removed from service; redur. dant components were operable; approvals were obtained prior to initiating the i _----___-.__--_-__-_--__..__.l- _

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   ' j work; activities .were accomplished . using approved' procedures and were - 9 inspected as applicable; procedures used were adequate to control the  :

activity; troubleshooting activities were. controlled and' the .' repair j records accurately reflected 'the activities; functional . testing and/o , calibrations- were performed prior to , returning components or systems to ' ' service; . QC records 'were. maintained; activities were accomplished by i qualified personnel; parts and materials used were properly certified; 1 radiological contr.ols were ' implemented; QC hold points were established . ; where required and were observed; fire prevention controls were -{ implemented; ' outside contractor force activities were controlled in i accordance with the approved QA program; and housekeeping was. actively : pursue Temporary Alterations (TACFs) The following TACFs were reviewed: ,

TACF 0-89-04-77,~ Pressure Switch 0-PS-77-429 on the B CST . l This TACF was implemented with WR 243097 and will be ' permanently replaced with.DCR 268 TACF 1-84-107-3, 1-FCV-3-3A MOV Fuse This TACF changed the rating on the turbine building MOV board 1A/11B fuse to the above valve from 0.8 amps to 2.0 amp This was necessary because of another FCR.' E-1756 that~ placed a heater in series with the MOV and. caused ~an overload conditio Permanent corrective action will b H accomplished with FCR 500 ,

No violations or deviations were identifie j 1 Work Requests The following work requests were evaluated: i WR B797142 - valve 2-FCV-3-148 WR B780979 - 2A AFW pump packing

 . The activity for WR B780979 was repeated'using WR B762254 and is under review to determine if the original packing  l or PMT was adequat This is identified as URI 327,328/
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89-12-0 . J l WR B790418 - repair 2-PI-002-129 WR B216758 - install temporary feedwater-instrumentation WR B282399 - FCV 003-84 valve stroke  ! WR B797201 - calibrate feedwater flow modifier l

WR B769961 - 2-PT-003-1 calibration ) i l l

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With the exception' o'f URIL 327,328/89-12-04, the activities , accomplished under the above WRs were adequately ' performed an 'l resulted. in acceptable. plant material . conditions. . First . line 1 management' involvement in the performance of the above. maintenance- ! activities was . adequate and appeared to be supported by .second. and? l third line maintenance management, outage management, and operation ,q management, j 5. Management Activities in Support of' Plant Operations TVA ' management activities were reviewed on ' a - daily basis by the NRC ' inspectors. The resident inspectors : observed that planning, scheduling, work control and other management meetings were effective' in controlling plant activities. First 1.ine- supervisors appeared.to be knowledgeable and - involved in the day-to-day activities of the plant. .First 'line supervisor involvement in the field was: observed. Management response to those plant , activities and events that . occurred during this inspection oeriod appeared: j timely and effective. ' An example - of this management ; action was its'

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involvement in . identifying and resolving the leaking .RCS RTD manifold  ; isolation valve described in paragraph '10. Following the public' meeting 1

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held on April 23, 1989, the licensee' demonstrated aggressive and-effective corrective actions with ' respect' to feedwater and steam l generator level

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control issues. However, the meeting was the culmination of approximately- ; seven months of discussions between the licensee and the' NRC over 'the root 1 cause of several steam generator level generated reactor trip ~

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6. Site Quality Assurance Activities in Support of Operations ,1 I During the inspection period, the site - QA staff performed ' audits, inspections, and reviews. Some of these issues were reviewed. by the - j inspector and found to be adequately resolved by the license The ] following audits were reviewed: QSQ-M-89-417, Operations QSQ-M-89-445, Feedwater Instrumentation .. This activity resulted in PRD SQQ-89-02-48 which addressed i procedure deficiencies and weaknesses 'in the planning of the l maintenanc , QSQ-M-89-456, Shutdown Board 2AA Maintenance i

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QSQ-M-89-462, Maintenance Activities This activity identified weaknesses in .the planning of maintenance.

' l QSQ-M-89-472,' Commitments and Corrective Action  ;

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This activity identified duplication between several procedures-including, AI-12, Corrective Action; AI-18.78, Post Trip-Review; SQA-84, Potential Reportable Occurrences; SQA-135, Commitment , Management; and SQA-186, Root Cause Assessment.- l

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10-One outstanding . issue was identified during the . review of the above- j QA surveillance activities. This issue, which- pertained to : the~

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l adequacy ' of the licensee's 10 CFR 50.59 safety evaluations, is a ' current topic of discussion between the licensee' and the NRC. The deficiencies identified by the QA' organization were being presented to line management during . this inspection . period. No violations'or 3 J deviations were identified. However, several. weaknesses were identi-fied in the' detail of maintenance planning supplied-to certain job This issue was discussed with line management during the exit of-this report and-line management is currently pursuing resolution of this issu In addition to the above audits, :the inspector discussed >several recent issues with the site QA Manager and the QA Surveillance GroupT Manager. The inspector determined that QA was deeply 11nvolved . in - these issues and had participated as a member of approximately.19 incident reviews chaired by the PORS ' organization.' This indicated complete implementation o f- the Smith / Martin memorandum o understanding concerning the inclusion of QA in the. review and root cause determination of plant event The inspector also discussed upcoming personnel reductions in the onsite QA organizatio Approximately twenty persons ' including. two managers will be' eliminated by the middle of June 1989. . This is expected to establish a stable level of personnel'on. sit . NRC Inspector Follow-up Items, Unresolved Items, Violations-(92701, 92702)

(Closed) Violation 50-327,328/88-29-04, Inadequate Procedures   '

This violation addressed-two examples of inadequate procedures.. The first- , example was procedure G-29, Radiographic Examination of Welded Joint ' This procedure was found inadequate in that ANSI B31.7 weld. standard inspection requirements for wall thickness reduction during' the qualification of field piping welds were not implemented. .The : second - example was procedure TI-89, Inservice Testing. This procedure was found ~l inadequate in that ASME Section XI testing requirements for two specific relief valves were not implemente ! In both cases the procedures .were corrected and adequate compensatory reviews were conducted. No additional or generic safety issues ~ were identified during the licensee's reviews. The inspector had no further , question This-violation is close ' 8. Licensee Event Report Followup (92700) i UNIT 1 l

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(Closed) LER 327/88-027, Radiation Monitor Calibration Procedure  '!
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The licensee discovered that'an incorrect source strength evaluation date-existed in SI-83, Channel Calibration for Radiation Monitoring System. As I a result, the. common fuel storage pool radiation monitors were affected and monitor 0-RM-90-103 was declared inoperabl The inspector reviewed the licensee's root .cause dete_rmination and - corrective actions. Corrective actions included immediate verification-o the inoperable radiation monicor, and a -revision of SI-8 The verification of the inoperable radiation monitor indicated that it' had performed within TS allowable limits. The corrective actions appek J to be adequate and the inspector had no further question ;

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This LER is close '

 (Closed) LER 327/88-028, Suspension of Fire Watches I'

As a result of a sequence of plant activities, unit operators opened the Unit 1 RCS head vents resulting in' an increase in airborne' activity in the i auxiliary buildin The operators manually initiated the auxili~ary J building gas treatment system and restricted access. The restricted , access prevented the performance of a compensatory measure for three breached fire barriers and LC0 3.7.12. The compensatory fire watches were returned within nine hour The inspector reviewed the licensee's corrective actions which also i included the repair of eight SI valves which contributed .to the sequence of events causing the airborne activity increase. The corrective actions- ,l appear to be adequate and the inspector had no further question This LER is close . Event Follow-up (93702) Unit 2 entered Mode 3 at 4i30 p.m. on April 6. At 4:55 p.m., the BIT to BAT recirculation was stopped and- the BIT recirculation valves were shut to stop back leakage from the RCS to the BIT. .This backleakage. was causing dilution of the BIT and the BAT. The inspectors questioned this procedure because the recirculation path provides the only method of ensuring that the proper BIT volume is maintained. TS 3.5.4.1 requires the BIT to be Operable in Mode with a minimum volume of 900 gallons. The action statement for this requirement:is to restore the BIT to Operable within l ' hour or be I in Hot Standby and borated to the appropriate shutdown margin within the next 6 hours. The licensee's position- was that even though no on-line verification of BIT volume exists when the BIT is off recirculation, there is no reason to suspect that the volume is being reduced. Further, the surveillance requirement is to . verify 'the proper . volume is present every 7 day The -licensee therefore concluded that taking the BIT off recirculation is allowed as long as the surveillance to verify volume is performed within 7 day The resident staff. contacted the project manager requesting an

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12' i interpretation' of the applicable T.S. ' After the resident inspector notified the plant staff of their concern regarding this issue, the BIT was placed back on recirculation at 10:02 a.m. on 10 April. .The l recirculation will be maintained until the issue is resolve Th inspector questioned the licensee's determination that taking the BIT off recirculation did not require entry into the LCO.' In addition,. the inspector questioned whether a proper safety evaluation had been completed. . Resolution of (1) BIT operability and (2) whether an .j adequate review pursuant to 10 CFR 50.59 safety evaluation' was =q performed will be tracked as URI 327,328/89-12-0 ' AT 8:40 p.m. on April 8, the licensee observed a small leak from the l high pressure seal on incore thimble K-2. .The leak wa's approximately one drop every . three minate A ' determination that a . pressure boundary leak existed was made and a Notification of Unusual Event was entered. The licensee later determined that the thimble leak did not constitute a pressure boundary leak and the NOUE was exite At 8:24 p.m. on April 11, Unit 2 was taken critical and low power testing began. Then at 12:50 p.m. on April 13, Unit 2 entered Mode 1 -) at 1% power. The inspector observed the licensees performance during i thi s . eval uati o I At 6:47 p.m. on April 13, RCS leakage.to the Pressurizer Relief Tank 1 (PRT) suddenly increased to a. value of 17.5 gpm. The leakage was i classified as IDENTIFIED LEAKAGE although the source.of the leak was ; not determined until a containment entry -and walkdown of potential ! leakage sources was performed approximately 7 hours later. Valve ) 68-530, RCS loop 3 Hot Leg RTD Bypass Manifold Outlet' isolation ' valve, was found to be not fully backseated. The valve ~has a leakof ; to the PRT after the third ring of packin RCS pre'ssure had ; apparently blown a hole in the valve packing due to the valve not ' being properly backseate The ' valve was backseated and the leak stopped at 2:05 a.m. on April 14. During the process of. identifying the leakage source, the plant staff was not initially ~ able to determine all the components and valves with leakoffs to the PRT. No single diagram or drawing represents. all potential sources to the ~ ; PR The team that entered containment was able to locate the leaking valve by backtracking from the PRT using a . heat sensitive' , infrared ' device. Since the licensee was not able to produce documentation that all inputs to the PRT were known and bounded by ~ i appropriate analysis, the classification. that.any leakage to the PRT ! is IDENTIFIED LEAKAGE may not be appropriate. Resolution of this issue will be tracked as IFI 327,328/89-12-0 e. .At 12:09 a.m. on April 15,. Unit 2 reactor tripped from 30% power during turbine generator trip testin When the turbine tripped, S.G. level program automatically reset from the ' 44% value to the no-load value of 33%. The feed controllers then began reducing _ _ _ _ _ - _ - _ - _ _ _ _ _ _ _ _ - _ _ _ _

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levels to the new value and severe level oscillations began. The f level oscillations were divergent and eventually . reached, the low-low 1 level setpoint on 'the #4 S.G. Post trip- analysis determined that the level controllers should: be placed in the manual mode in fut'ure ; cases when the turbine is trip tested to preclude the step change-in - j program icvel from initiating. a level transient. 'This is example' a ; of NCV 327,328/89-12-06. The enforcement and corrective action- ) aspects of this. and examples b and c are discussed -in paragraph ! of.this repor i f. At 12:48 a.m. on ' April 16, Unit 2. reactor tripped from 16% power on ] S.G. low-low level on #1 S.G. The trip was caused by level oscilla- .i tions that began when operators were attempting to transfer from j

level' control with the auxiliary feedwater system to control with the' main feedwater system. The post trip review of this trip determined , that the transfer from the auxiliary.feedwater system to main feed- l water system should be performed at a lower power level. The capabil- i ities of the auxiliary feedwater system to keep up with steam flow'is I at its upper limits at approximately 1% power, and the changeover to ; the main feedwater system began at approximately 1.5% power. As a ! result, levels had begun to decrease below program level by the time the main feedwater system bypasses were placed in automatic. The bypass valves began to open to restore level to program, which-introduced a large amount of cold feedwater. This caused a large level shrink and level oscillations began' which eventually resulted ! in a low-low level trip on #1 This is NCV 327,328/89-12-06' example The enforcement and correctives action aspects of this issue are discussed in paragraph 9.h of this report,

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g. At 4:47 a.m. on April 19, Unit 2 reactor tripped on low-low leve1Lon j S.G. # 2. This trip was attributed to an inability ta . m rol : levels in the manual mode on .the feedwater system bypass valve While escalating power to above 4%, automatic bypass control on loop 1 became erratic. Operators placed this bypass controller in manuti and continued the startu At approximately 10% ; power, . loop #2 controller.also became erratic and was placed in manual,' and the power escalation was resumed. While attempting to transfer. to th main feed regulating. valves from the bypasses at approximately 18% l power, level oscillations began which the operators were unable to control. Reactor power was reduced by stepping in control rods,- r which contributed to the level oscillations. The reactor eventually tripped from a power level of 1.9%. This is NCV 327,328/89-12-06 example c. The enforcement and corrective aspects of this issue are , discussed in paragraph 9.h of this repor j h. At 2:00 p.m., on April 23, a public meeting with TVA was held to discuss the three most recent Unit 2 reactor trips' and the six previous trips. TVA presented its corrective actions to ' resolve the increased number of reactor trips. These corrective actions are ___

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1 ; i i described in the NRC meeting minutes and were subsequently l inspected I by the inspector The corrective actions .were determined to be -) adequat The .three trips are identified as NCV 327,328/89-12-06, 'i examples ' a through - c for violation of TS. 6.8.1. for L failure 'to establish, implement and follow procedure These are not being cited because the criteria specified in Section V.G. of the Enforce-ment Policy were satisfied. .NCV 327, 328/89-12-06 is considered close The inspectors' observed line management and QA management involvement in the resolution of these issues. This involvement'was determined

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t a be adequate and is discussed in the ' respective section of this repor l 10. Exit Interview (30703) { i Tae inspection scope and findings were summarized. on May 4,1989, with'- those persons indicated in paragraph .1. The- Senior Resident Inspecto Mescribed the areas . inspected and discussed in detail the. inspection findings listed below. The licensee acknowledged the inspection findings and did not identify as proprietary any of the material reviewed by the inspectors during the inspectio Inspection Findings: 1' No violations, or deviations were identifie (0 pen) URI 327,328/89-12-01, " Licensed Power Indication" (Closed) NCV 327,328/89-12-05, " Diesel Generator Maintenance" 1 (0 pen) URI 327,328/89-12-04, "AFW. Maintenance" (Closed) VIO 327,328/88-29-04, " Inadequate Procedures" (Closed) LER 327,328/88-027, " Radiation Monitor Calibration Procedure" )

 (Closed) LER 327,328/88-028, " Suspension of. Fire Watches" (0 pen) URI 327,328/89-12-02, " Recirculation of BIT" (0 pen) IFI 327,328/89-12-03, "RCS Identified Leakage" (Closed) NCV 327,328/89-12-06, "Feedwater Reactor Trips" The areas of Operations and Surveillance. appeared to be adequate,-    -l improving and fully capable of supporting current plant,. operations. The observed activities of the Operations section, the control room operators in phrticular, were professional and well managed. The area of mainte-nance had some licensee identified weaknesses. . in the area of work planning. However the area of maintenance was adequate during this    .'
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-period. Line. management involvement was acceptable and QA involvement in current issues and operations is' sues was aggressiv During the reporting period, frequent discussions were held with the Site

' Director, Plant Manager and other managers concerning inspection finding ,

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I List of Acronyms and Initialisms ,. l ABGTS- Auxiliary Building Gas Treatment System ABI - Auxiliary Building Isolation ABSCE- Auxiliary Building Secondary Containment Enclosure AFW - Auxiliary Feedwater AI - Administrative Instruction AOI - Abnormal Operating Instruction AUD - Auxiliary Unit Operator ASOS - Assistant Shift Operating Supervisor ASTM - American Society of Testing and Materials BIT - Boron Injection Tank BFN - Browns Ferry Nuclear Plant C&A - Control and Auxiliary Buildings CAQR - Conditions Adverse to Quality Report CCS - Component Cooling Water System CCP - Centrifugal Charging Pump CCTS - Corporate Commitment Tracking System CFR - Code of Federal Regulations COPS - Cold Overpressure Protection System CS - Containment Spray CSSC - Critical Structures, Systems and Components CVCS - Chemical and Volume Control System CVI - Containment Ventilation Isolation DC - Direct Current J DCN - Design Change Notice i DG - Diesel Generator DNE - Division of Nuclear Engineering ECN - Engineering Change Notice ECCS - Emergency Core Cooling System EDG - Emergency Diesel Generator EI - Emergency Instructions ENS - Emergency Notification System E0P - Emergency Operating Procedure EO - Emergency Operating Instruction ERCW - Essential Raw Cooling Water ESF - Engineered Safety Feature FCV - Flow Control Valve  : FSAR - Final Safety Analysis Report  ! GDC - General Design Criteria GOI - General Operating Instruction GL - Generic Letter Heating Ventilation and Air Conditioning

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HVAC - HIC - Hand-operated Indicating Controller H0 - Hold Order HP - Health ihysics ICF - Instruction Change Form , IDI - Independent Design Inspection ' IN - NRC Information Notice r i

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16-IFI - Inspector Followup Item . IM - Instrument Maintenanc IMI - Instrument = Maintenance Instruction-IR- -

Inspection Report

KVA - Kilovolt-Amp KW - Kilowatt KV - Kilovol LER - Licensee. Event Report LCO - Limiting Condition for Operation LIV - Licensee Identified Violation LOCA - Loss of Coolant Acciden MCR - Main Control Room MI - Maintenance Instruction MR - Maintenance Report MSIV - Main Steam Isolation Valve NB - NRC Bulletin NCV - Non-cited Violation

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Notice of Violation NQAM - Nuclear Quality Assurance Manual NRC - Nuclear Regulatory Commission OSLA - Operations Section Letter . Administrativ OS LT - Operations Section Letter . Training PLS Precautions, Limitations, and Setpoints PM - Preventive Maintenance' PPM - Parts Per Million PMT - Post Modification. Test PORC - Plant Operations Review Committee l P0RS - Plant Operation Review Staff PRO - Potentially Reportable Occurrence QA - Quality Assurance J QC

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Quality. Control  ! RCDT - Reactor Coolant Drain Tank i RCP - Reactor Coolant Pump l RCS - Reactor Coolant System  ; RG - Regulatory Guide 1 RHR - Residual Heat Removal i RM - Radiation Monitor- l R0 - Reactor Operator l RPI - Rod Position Indication RPM - Revolutions Per Minute i RTD - Resistivity Temperature Device Detector  ! RWP - Radiation Work Permit i RWST - Refueling. Water Storage Tank  ; SER - Safety Evaluation Report l SG - Steam Generator i SI - Surveillance Instruction i SMI -

 'Special Maintenance Instruction   I S0I -

System Operating Instructions 1 SOS - Shift Operating Supervisor  :

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Sequoyah Standard Practice Maintenance SQRT - Seismic Qualification Review Team SR - Surveillance Requirements SRO - Senior Reactor Operator S50MI- Safety Systems Outage Modification Inspection SSQE - Safety System Quality Evaluation SSPS - Solid State Protection System STA - Shift Technical Advisor STI - Special Test Instruction TACF - Temporary Alteration Control Form TAVE - Average Reactor Coolant Temperature TDAFW- Turbine Driven Auxiliary Feedwater TI - Technical Instruction TREF - Reference Temperature TROI - Tracking Open Items TS - Technical Specifications TVA - Tennessee Valley Authority UHI - Upper Head Injection UO - Unit Operator URI - Unresolved Item USQD - Unreviewed Safety Questien Determination VDC - Volts Direct Current VAC - Volts Alternating Current WCG - Work Control Group WP - Work Plan WR - Work Request l l

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