IR 05000327/1988015

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Insp Repts 50-327/88-15 & 50-328/88-15 on 880208-12.No Violations or Deviations Noted.Major Areas Inspected:Ler Followup,Action on Previous Findings,Info Notices,Potential RO & Problem Identification Repts
ML20151B363
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 03/30/1988
From: Hunegs G, Jenison K, Mccoy F
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20151B357 List:
References
50-327-88-15, 50-328-88-15, IEIN-86-108, IEIN-87-051, IEIN-87-057, IEIN-87-51, IEIN-87-57, NUDOCS 8804080188
Download: ML20151B363 (31)


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u UNITED STATES

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o NUCLEAR REGULATORY COMMISSION

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j 101 MARIETTA ST REET, N.W.

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Report Nos.:

50-327/88-15, 50-328/88-15 Licensee:

Tennessee Valley Authority 6N 38A Lookout Place 1101 Market Square Chattanooga, TN 37402-2801 Docket No.:

50-327 and 50-328 License Nos.: DPR-77 and DPR-79 Facility Name:

Sequoyah Units 1 and 2 Inspection Conducted:

February 8 thru February 12, 1988 Inspectors:

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K. Jenis6n, enorR[ dent' Inspector,TeamLeader Date Sign (d N

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G. Hunegs, As si s' tant am Leader Date gn Team Members:

V. Flitton l

M. Good W. Liu L. Mellen R. Moore M. Runyan G. Walton

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Approved by b

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, p ; McCoy( Clifef, Projects Section 1

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Division of TVA Projects SUMMARY Scope: This announced inspection was conducted in the areas of licensee action on previous insptction findings, licensee event report follow-up, NRC Information Notices, potential reportable occurrences, problem identification reports, industry generated nuclear experience review items, IE Bulletins, and Condition Adverse to Quality Reports (CAQRs). The inspecticn was conducted to determine if the programs and corrective actions implemented are effective in assuring adverse conditions, including generic issues are dispositioned adequately.

Background: On October 26, 1984, Non-conformance Report (NCR) WBN NEB 8415 was initiated at the Watts Bar site documenting that certain containment pressure transmitters were not environmentally qualified.

TVA's engineering organization, the Office of Engineering, (now referred to as the Division of Nuclear Engineering (DNE)) recognized the need to review similar instrumentation at the Sequoyah site but, initiated no corrective action until nearly two 880408018e 880328 DR ADOCK 05000327 DCD

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2 months. later when NCR SQN NEB 8501 was initiated.

The Failure Evaluation / Engineering Report (FE/ER) attachment to NCR SQN NEB 8501 concluded l

that as a result of the failure mode, a reactor operator would have inaccurate

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information needed to mitigate a Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB). As a result, certain safety functions or actions would be defeated or delayed which could lead to exceeding the containment design pressure limits. An NRC review conducted from March 27-29, 1985, identified a breakdown in management controls for evaluating and reporting potentially significant safety conditions. As a result, on June 14, 1985, the NRC issued order EA 85-49 (Taylor /Parris, RIMS L44 850620001) modifying the Sequoyah site licenses.

The order was to ensure:

(1) that potentially significant safety conditions are promptly evaluated and corrected; (2) that management and procedural controls are adequate to assure that responsible levels of management are promptly made aware of potentially significant safety conditions, and; (3) that those individuals responsible for reporting significant safety conditions to the NRC are promptly apprised of potentially significant conditions which might require reporting under 10 CFR 50.72 Part 21 and realized the importance of such reporting.

Results:

No violations were identified during this inspection effort.

However, based on the issues discussed in the details section of this report, the licensee's corrective action program needs improvement in the following areas in order to ensure prompt evaluation and correction of Conditions Adverse to Quality (CAQs), Problem Identification Reports (PIRs) and Nuclear Experience Review (NER) items and it is not appropriate to consider removal of the restrictions applied to the Sequoyah site by order EA 85-49. Areas requiring additional licensee action are:

1.

Improve the speed and reliability of operability / significance determinations, and root cause determination.

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Ensure those personnel who make operability / significance determinations in the Generic CAQ, and Nuclear Experience Review (NER) processes are adequately trained and possess the correct qualifications to make operability / significance determinations.

3.

Improve the completeness and auditability of CAQ documentation re-quired for adequate management reviews prior to closure.

4.

Ensure managerial CAQR training requirements imposed by order EA 85-49 are current.

5.

Resolve specified technical questions unique to certain individual CAQRs.

The licensee's corrective actions will be reviewed in a followup inspection, 327,328/88-1.

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REPORT DETAILS

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Persons Contacted j

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Licensee Employees

H. Abercrombie, Site Director J. Anthony, Operations Group Supervisor R. Buchholz, Sequoyah Site Representative

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  • M. Burzynski, Regulatory Licensing Manager

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M. Cooper, Licensing Supervisor

  • M. Couillens, Division of Nuclear Quality Assurance
  • W. Eaton, Division of Nuclear Engineering Planning H. Elkins, Instrument Maintenance Group Manager

R. Fortenberry, Technical Support Supervisor

  • J. Hamilton, Quality Engineering Manager
  • M. Harding, Licensing Group Manager

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  • T. Jones, Quality Assurance Analyst

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  • N. Kazanas, Division of Nuclear Quality Assurance Director

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  • G. Kirk, Compliance Supervisor

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  • J. LaPoint, Deputy Site Director
  • W. Ludwig, Nuclear Experience Review Group Supervisor

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L. Martin, Site Quality Manager

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R. Olson, Modifications Branch Manager l

B. Patterson, Maintenance Superintendent i

R. Pierce, Mechanical Maintenance Supervisor

R. Prince, Radiological Control Superintendent i

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  • H. Rogers, Plant Operations Review Staff
  • A. Rosenberg, Engineering Assurance Engineer

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  • E. Sliger, Manager of Projects

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l S. Smith, Plant Manager
  • J. Sullivan, Regulatory Engineering Supervisor

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  • D. Whaley, Nuclear Experience Review Engineer i

"C. Whittemore, Licensing Engineering

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  • A. Wilkey, Site Audit Manager

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B. Willis, Operations and Engineering Superintendent

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Other licensee employees contacted included engineers, technicians, j

operators, mechanics and office personnel.

NRC Attendees

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"G. Hunegs

  • K Jenison
  • W. Liu i
  • F.

McCoy

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  • L. Mellen j

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  • R. Moore i

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  • M.

Runyan

  • G. Walton
  • Attended exit interview 2.

Exit Interview The inspection scope and findings were summarized on February 12, 1988, with those persons indicated in paragraph 1.

No violations or deviations were identified. However, the licensee committed to implement corrective actions in several areas identified in the summary paragraph of this report. Although proprietary material was reviewed during the inspection, no proprietary material is contained in this report.

Note: A list of abbreviations used in this report is contained in paragraph 12.

3.

Licensee Action on Previous Inspection Findings (92702)

(0 pen) Violation (VIO) 327,328/86-73-03, Failure to Issue Potential Generic Condition Evaluation Reports and Untimely Response by Other Plants.

This violation resulted from a team inspection performed at all TVA sites and involved various Condition Adverse to Quality Reports (CAQRs) where Potential Generic Condition Evaluation Memos (PGCEMs) were responded to in an improper or untimely manner.

TVA's response to the violation of July 10, 1987, was acknowledged and accepted by the NRC by letter dated October 6, 1987.

TVA's corrective action included focusing management attention on the resolution of open CAQs, prioritization and scheduling of open CAQs, and implementing a new reporting systems to highlight late / outstanding CAQs that require further management attention.

In addition, revisions were made to TVA's Nuclear Quality Assurance Manual (NQAM) and implementing procedures to provide more controlled and

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centralized requirements for the conduct of generic reviews.

The inspector reviewed CAQRs generated at other sites, that required a

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Sequoyah generic review, and were received at Sequoyah since September 1, 1987.

In addition the inspector reviewed CAQRs generated at Sequoyah since September 1, 1987. The reviews and generic determinations appear to be generally timely.

Significant improvement has been made in the CAQ program, however, several cases of untimely resolution were identified by both TVA and the NRC as described in later paragraphs of this report.

This item will be remain open and be reviewed by the NRC af ter TVA has completed corrective actions on the CAQ program improvements identified in the summary paragraph of this report.

(Closed) Unresolved Item (URI) 327,328/86-27-02, Long Term Unincorporated Engineered Change Notice.. )

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applied to the temporary alteration change form (TACF).

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observed was that many TACFs were implemented.in the plant without the

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same safety review, Division of Nuclear Engineering review and design support, and Plant Operations Review Committee (PORC) review that would have been affected on the actual inplant change if the actual inplant change had been implemented through the Design Change Request (DCR)/ Engineering Change Notice (ECN) process.

The current TVA procedure covering TACFs is AI-9, "Control of Temporary Modification". AI-9 requires a complete review and approval of TACFs by DNE.

The inspector did not identify any current TACFs that had not been PORC/ Qualified individual (QI) approved and no recent TACFs that did not have concurrence from DNE.

This item is closed.

(Closed) URI 327,338/86-44-01, Operating Experience Review of NRC Inspection Reports.

This issue involved the Sequoyah site not having objective evidence that it had received any original or condensed material concerning NRC inspections reports for inspections conducted at Watts Bar site (390/84-73, 85-21, 85-32, and 85-51).

In addition, site management did not have any objective evidence that they had implemented action to determine if the Watts Bar issues were present at Sequoyah, other than tnote actions prompted by several phone calls between NRC Region II management and the Sequoyah Plant Manager.

The present nuclear experience review program is implemented by TVA Standard Practice SQA 26, "Review, Implementation, and Reporting of Nuclear Experience Review Information".

This program includes NRC inspection report findings.

The inspector sampled the program to determine if information was reaching the Sequoyah site from other sites.

The inspector did not find any examples where NRC inspection report information was not delivered to the site and reviewed by Sequoyah personnel.

This item is closed.

(Closed) VIO 327,328/86-44-04, Inadequate Surveillance Instruction Review Program.

This generic issue involved an inadequate surveillance instruction review program associated with URI 327,328/86-44-01.

It was determined to be technically resolved in inspection report 327,328/87-36. However, it was left open administratively, pending resolution of escalated enforcement matters.

The administrative actions involving resolution of escalated enforcement matters are complete. The licensee's actions are complete and adequate as

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a result of the implementation of SI-1, "Surveillance Instruction Program".

This violation is closed.

(0 pen) URI 327, 328/86-53-01, Failure to Promptly Correct Conditions Adverse to Quality Audit Items.

Per letter from S.

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Ebneter (NRC) to S.

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White (TVA) dated December 29, 1987, a Notice of Violation will not be issued regarding this matter.

Because of the program implementation improvements necessary to support Unit 2 startup, this item will remain open.

4.

Unresolved Items Unresolved items are matters about which more information is required to determine whether they are acceptable or may involve violations or deviations.

No unresolved items were identified during this inspection.

5.

Licensee Event Report (LER) Followup (92700)

LERs were reviewed considering the following inspection attributes:

reporting requirements were met, causes were identified, corrective actions appeared appropriate, generic applicability was considered, the LER forms were completed, the licensee had reviewed the event, no unreviewed safety questions were involved, and no violations of regulations or Technical Specification conditions had been identified.

In addition, for those items requiring hardware, drawing and/or procedural changes in the plant the inspector verified that:

redundant components were operable when corresponding components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished using approved procedures and were inspected as appli-cable, functional testing and/or calibrations were performed prior to returning components or systems to service, and quality control records were maintained and can be retrieved.

The following LERs were adequate with respect to the inspection criteria in paragraph 5 above.

The Potential Reportable Occurrences (PR0s) that resulted in the LERs were also reviewed.

LER #

Description / NRC Status 327/87-055 Potential Loss of Auxiliary Feedwater Due to Faulty Pressure Switch Design.

327/87-061 Appendix R Associated Circuits That Share a Common Power Supply With an Appendix R Circui.'

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327/87-073 Inadequate Design of Centrifugal Charging Pumps Auxiliary Lube Oil System Could Result In the Failure of High-head Safety Injection to Start on a Manual Signal.

327/87-074 Sequoyah Emergency Procedures Do Not Adequately Address the Opening of Certain High-head Safety System Valves following a Postulated LOCA from Mode 4 Condition.

327/87-075 Inadequate Procedure and Misinterpretation of Operational Modes Resulted in a Surveil-lance Requirement Not Being Met and Condi-tion Not Being Reported.

327/87-076 A Containment Ventilation Isolation Occurred as the Result of Test Personnel Connecting Test Equipment Incorrectly.

327/87-077 Inadequate Design of the Containment Isolation System for the Hydrogen Analyzers Could Result in Bypass Leakage Following a LOCA.

327/88-002 Emergency Raw Cooling Watcr (ERCW)

Radiation Monitor Declared Inoperable Without Compliance With the Limiting Condition for Operation (LCO) as a Result of Misinterpretation of the LCO.

There were no inadequate LERs found with respect to the acceptance criteria in paragraph 5 above.

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No violations or deviations were identified.

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NRC Information Notices (ins) (92701)

The following ins were reviewed.

The inspector verified that: correc-l tive actions appeared appropriate, generic applicability had been consid-ered, the licensee had reviewed the IN and that appropriate plant personnel were knowledgeable, no unreviewed safety questions were involved, and that violations of regulations or Technical Specification conditions did not appear to occur.

M Description

'87-051 Failure of Low Pressure Safety Injection Pump Due to Seal Prchlems.87-057 Loss of Emergency Boration Capability Due to Nitrogen Gas Intrusio ~_.

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IN Description

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i 86-108, Supplement 2 Degradation of Reactor Coolant; System Pressure Boundary Resulting from Boric Acid Corrosion.

i No violations or deviations were identified.

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Potential Reportable Occurrence (PRO) (92700)

i The PRO process is a method to provide for the evaluation, notification, i

and documentation of events which are potentially reportable to the NRC.

The following PR0s were reviewed and evaluated. The inspector verified I

that: reporting and operability determinations were adequate, causes had l

been identified, corrective actions appeared appropriate, generic appli-cability had been considered, LERs or CAQRs were generated when appropri-

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ate, licensee management had reviewed the issue, no unreviewed safety questions were involved, and no violations of regulations or Technical

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Specification conditions had been identified.

In addition, for those i

items requiring hardware, drawing and/or procedural changes in the plant

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the inspector verified that: redundant components were operable when

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corrSsponding components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished using approved procedures and were inspected ss applicable, functional i

testing and/or calibrations were performed prior to returning components l

or systems to service, and quality control records were maintained and can

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be retrieved.

a.

The following PR0s were adequate with respect to the acceptance l

criteria in paragraph 7 above.

P_R_0 Description f

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1-88-012 Auxiliary Building Isolation j

2-88-005 Containment Spray Heat Exchanger

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1-87-447 Control Room Emergency Ventilation i

1-88-23 Transmitter 0-PT-67-445 Out of Calibration l

2-88-11 Deficient Procedure IMI99-RT-604B

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1-88-11

'B' Train Radiation Monitors Inoperable f

1-87-460 Centrifugal Charging Pump Flow Path l

1-87-435 FCV-63-110 Not Included On SUI 63.1 i

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PRO Description (cont'd)

1-87-439 The Hydrogen Analyzer Does Not Provide Adequate Post-LOCA Containment Isolation Upon a Loss of Auxiliary Control Air.

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2-87-114 Two 1 1/4" Diameter Bolts Specified by Drawing 48N1231R7 for Residual Heat Removal (RHR) Heat Exchangers Top Supports Were Not Installed.

2-88-027 Cold Leg Secondary Check Valves63-551, 63-553,

,"63-555, and 63-5-57 Failed Their Leak Test per SI-166.10 Part A.

2-88-016 Unit 2, Flood Mode Boration Purops 1-87-436 Containment Ventilation Isolation (CVI) on Unit 1 Occurred 2-88-019 Performance of SI-191.2 Cannot Be Located 1-87-443 Control Cable X2 and Terminal Block in The Bottom of Compartment 3 on 480V Ventilation Board 182-B Was Burnt in Half.

1-87-432 161 Out of 304 Time Delay Relays, Cycle Timers, Level Switches, and Load Controllers Calibrated as a Result of LER 327/87010 Were Found Outside the Acceptance Criteria.

1-87-445 Found Instruments Out of TI-54 Tolerance During Performance of SI-198.

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2-87-125 Received CVI on Unit 2.

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2-88-008 Data for TI-83-5002 Was Outside the Acceptance

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2-88-014 Leakage Past Flow Control Valve 2-87-513 on the Upper Head Injection System Caused Unexpected Drop in Pressurizer Level of approximately 1%.

2-88-015 Possible Inoperable Reactor Coolant System (RCS)

Temperature Channels.

1-88-020 Inoperable Penetrations

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2-88-012 During Performance of 51-106, Inspection of Flow paths in Unit 2 Ice Condenser, Revealed Blocked

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Passages in the Ice Bed Which Caused the SI to Fail, b.

The following PR0s were inadequate with respect to the acceptance criteria in paragraph 7 above.

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Description 1-87-039 Breaker 7 on the Vital Power Board I-II. This PRO was closed out prior to a root cause being determined.

A decision on operability or CAQR

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t applicability appeared to be premature.

1-88-020 Inoperable Penetration Pipe Barriers.

The root cause determination was inadequate.

1-88-033 Containn,ent Ventilation Isolation. The root cause determination was inadequate.

8.

Problem Identification Reports (PIRs)

PIRs are used by the licensee as a precursor document to CAQR, Part 21 and construction deficiency reports (10 CFR 50.55e). They are intended to be a disposition process for safety concerns.

The following PIRs were reviewed and evaluated.

The inspector verified that:

reporting and operability requirements had been met, causes had been identified, corrective actions appeared appropriate, generic applicability had been considered, administrative timeliness requirements had been met, licensee management had reviewed the event, no unreviewed safety questions were involved, and no violations of regulations or Technical Specification conditions had been identified. In addition, for those items requiring hardware, drawirg and/or procedural changes in the plant the inspector verified that: redundant components were operable when corresponding components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished using approved procedures and were inspected as applicable, functional testing and/or calibrations were performed prior to returning components or systems to service, and quality control records were maintained and can be retrieved.

a.

The following PIRs were adequate with respect to the acceptance criteria in paragraph 8 above.

PIR Description SQNCEBS795 Consideration of Differential Seismic Anchor Movements Between Buildings.

i SQNMEBS801 HVAC Cooling Load Calculation for Auxiliary Building.

SQNM BSS02 Excess Letdown Piping.

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PIR Description (cont'd)

SQNMEB8803 Reference Drawing not Updated.

SQNMEB8804 Valve Number left Off Drawing.

SQNMEB8E05 Flow Diagram for the Nitrogen System.

SQNMEB8806 Non-safety Related ANSI B 31.1 Concerns.

SQNEEB8802 Radiation Monitoring Generic Concerns.

SQNEEB8803 Condenser Vacuum Vent.

SQNEEB8804 Teflon Tape in Radiation Monitoring System.

SQNEEB8805 Flushing Process Flow Lines.

SQNSQP8801 Drawing Problems.

SQNSQP8802 Drafting Errors.

SQNMEB87111 Regenerative Cooling Calculations.

SQNMEB87104 ERCW Design Criteria, SQNMEB87086 Component Cooling Water System (CCWS) and Auxiliary Feedwater System (AFW) Calculations.

b.

The following PIRs were inadequate with respect to the acceptance criteria in paragraph 8 above.

PIR Description i

SQNCEB87069 Additional Hangers on Pipe Supports.

The generic review was incorrectly noted which was corrected after the NRC inspector identified the discrepancy.

SQNCEB87093 Epoxy Grouted Anchors.

This was not identified i

as a CAQ as required, and was corrected af ter identified by the NRC inspector.

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SQNCEB87099 Design Piping Movements Exceed Clearances.

The timeliness requirements were not met, i

SONMEB87092 Review of Containment Spray Heat Exchanger j

Ca 'ic ul ati o n s.

The generic review documentation was inconsistent.

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SQNMEB87081 Temperature Profiles for High Energy Line Break (HELB). The timeliness requirements were not met

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as the generic review was 8 months late.

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SQNMEBS7076 Test Scoping Document. The timeliness requirements for the generic response were not

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met.

The inspcctor reviewed the majority of the PIRs issued in 1988.

Although problems existed with PIRs issued early in the j

i implementation phase of the new PIR program, these problems do not I

appear to exist in the current PIR program.

Additionally the inspector reviewed a draft of SQEP-71, "PIR Procedures", and found that it adequately addressed the weaknesses identified early in the (

implementation of the new program. The licensee had also identified I

these weaknesses in the PIR program and drafted procedures prior to

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these NRC findings.

No violations or deviations were identified.

9.

Industry Generated Nuclear Experience Review (NER)

The following NERs were reviewed. The inspector verified that: reporting and operability requirements had been met, causes had been identified, corrective actions appeared appropriate, generic applicability had been considered, administrative timeliness requirements had been met, licensee management had reviewed the event, no unreviewed safety questions were involved, and no violations of regulations or Technical Specification conditions had been identified.

In addition, for those items requiring hardware, drawing and/or procedural changes in the plant the inspector verified that: redundant components were operable when corresponding i

components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished using approved procedures and were inspected as applicable, functional testing and/or

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calibrations were performed prior to returning components or systems to

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service, and quality control records were maintained and can be retrieved, j

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The following NERs were found to be adequate with respect to the

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acceptance criteria in paragraph 9 above.

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Description 87-0957-001 Control Rod Drive System Hydraulic Control Deficiencies.

87-0969-001 Emergency Diesel Testing Outside of Design Basis.

87-0965-001 Reactor Trip Due to Steam Generator Lo-Lo Level.

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NER Description

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88-0039-001 Instrument Air System Causes Reactor Trip

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Davis-Besse.

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CHS87-0103 PORC Review of Interpretations.

l BFB87-1092 Charcoal Absorber Trays.

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WBP87-0727 Installed Pipe Supports.

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BFP87-0732 Timeliness of Nonconformance.

WBQ 87-1002 Retrievability and Auditability of Quality

Assurance (QA) Documentation.

I 87-0974-001 Improper Control Room Emergency Ventilation Flow l

Testing.

b.

The following NERs were inadequate with respect to the acceptance criteria ia paragraph 9 above.

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Description 87-0978-001 Terry Turbine Overspeed Trip Mechanism Unsafe.

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This issue was identified on December 16, 1987, and was not sent to the correct site group until February 3, 1988, which exceeded the required time limits.

88-0025-001 Reactor Vessel Head "0" Ring Leakage. The Westinghouse technical bulletin affecting safety

related equipment was received on December 4, i

1987.

Corrective action on the NER evaluation was not determined until February 3, 1988.

The NER process had timeliness, adequacy of review for significance, and qualification of reviewer problems as indicated by the above two examples.

No violation or deviations were identified.

10. Conditions Adverse to Quality (CAQR)

The licensee, through its approved QA Program has implemented the CAQR process.

The approved QA program has been the subject of several NRC i

insrections (NRC inspection reports 327,328/87-25, 26, 55).

This inspection examined the implementation of the approved program througa the l

implementation of the CAQR process and other supporting sub processes.

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The CAQCs reviewed during this inspection were divided into two groups.

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The first group of CAQRs consists of those CAQRs that were generated at Sequoyan and are specific to Sequoyah.

This first group is discussed in subparagraph 10 a.

The second group of CAQRs consists of those CAQRs that were generated at TVA sites other than Sequoyah.

This second group also includes CAQRs that were generated in the Division of Nuclear Engineering and is discussed in paragraph 10 b.

a.

CAQRs Generated at Sequoyah.

The following CAQRs were reviewed.

The inspector verified that:

reporting and operability requirements had been met, causes had been

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identified, corrective actions appeared appropriate, generic applicability had been censidered, administrative timeliness requirements had been met, licensee msnagement had reviewed the event, no unreviewed safety questions were involved, and no

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violations of regulations or Technical Specification conditions had been identified.

In addition, for those items requiring hardware,

l drawing and/or procedural changes in the plant the inspector verified

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that:

redundant components were operable when corresponding

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components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished

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using approved procedures and were inspected as applicable,

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l functional testing and/or calibrations were performed prior to I

returning components or systems to service, and quality control

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records were maintained and can be retrieved.

(1) The following CAQRs were adequate with respect to the acceptance i

criteria in paragraph 10 a above SQP 87 1375 Fire Protection Breaker Failure SQP 87 1455 TACFs Do Not Have Sufficiently Detaile Unreviewed Safety Question Determination (USQD)

SQF 87 0115 Refueling Water Storage Tank (RWST) Level Transmitters SQF 87 0151 Appendix R Separation of Instrument Line SQF 87 0132 ERCW Upper Containment Setpoint and Remote Switches SQP 88 0068 480 Volt Shutdown Boards Long Time Delay Trip Settings 50P 88 0017 Pipe Chase Cooler B-B Failure Te Auto Start SQP 87 1382 Failure to Perform Quality Control (QC) Hold Points on a Work Request (WR)

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SQP 87 1286 Containment Manual Throw Switch Indication in Control Room SQP 87 1399 Level I Piping Failed (PT) Testing SQP 87 1251 Post WR Reviewers Failing to Ensure That Configuration Changes are Documented SQP 87 1235 Design Documentation Problems With the Heating, Vel.tilating and air Conditioring (HVAC) System SQP 87 1334 26 Containment Penetrations Were not Being Tested for Bypass Leakage SQP 87 1182 Primary Electrical Penetration Conductors Were Not Protected From Overload or Short Circuit SQP 87 1212 Auxiliary Control Air Compressor Was Out of Service Due to Personnel Error SQP 87 1279 High Temperature Switches Could Disable the HVAC System in the 6.9 KV Shutdown Board Room SQP 87 1276 A Fire Damper was Miswired (2) The following CAQRs were found to be adequate with respect to generic applicability only.

SQP 87 1641 Vendor Qualified Maintenance SQP 37 1600 Cable Separation Criteria SQP 87 1609 Reactor Coolant System Wiae Range Pressure Indicator SQP 87 1657 Control Building Ventilation SQP 87 1654 WR B232136 Was Not Reviewed by QA SQP 87 1605 Drawing Deviation SQP 87 1613 Diesel Generator HVAC Fans SQP 87 1606 Control Room Emergency Ventilation i

5Q0 87 1655 Nuts Were Not Hand Tightened

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SQP 87 1607 FGTS Flow Switches

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SQA 87 1642 NPDES Permit Number TN002645)

SQP 87 1650 Hydrogen Analyzer SQP 87 1656 Conduit Fitting SQP 87 1608 RWST Level Post Accident Monitor (PAM)

Indicators SQP 87 1644 VHI Setpoint Calibration (3) The following CAQRs were inadequate with respect to the accep-tance criteria in sub paragraph 10 a above.

SQP 87 1449 Failure to Follow Dasign Drawings When Implementing Construction of Support.

The CAQ was discovered on August 29, 1987, but not documented as a CAQR until September 21, 1987.

The determination of root cause end recurrence control was approximately one month late.

The CAQR was escalated on November 24, 1987, for failure to determir.e corrective action within 30 days of the required dete.

The escalation was resolved on December 2, 1987.

SQP 87 1448 Inadequate Design Basis for Containment Ventilation Isolation.

The CAQ was discovered on April 2, 1987, but not documented as a CAQR until September 4, 1987 All other time limits were exceeded by 50 to 100 percent.

The CAQR was escalated on November 3, 1987, for time limit violations, and resolved on November 19, 1987, SQP 87 1009 Category I Structures, Including Pipe Supports, Are Attached to Embedded Plates V5ich Could Not Be Located On Design Drawings.

The determination by the licensee that this CAQR was "Not Significant", was unjustified.

Classification as not significant negates management reviews and certain technical l

analyses. The licensee provid9d a response

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to the inspector which mitigated most of this concern, however, it still appears a determination of "significant" would have i

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been a more conservative position pending full problem disclosure.

Corrective action completion was due by December 15, 1987, and was not complete at the time of this inspection. The CAQR was escalated on February 1, 1988 for corrective action greater than 130 days overdue. The escalation action itself was two weeks late.

SQP 87 1126 Potential Problems With The Control Room Emergency Ventilation System.

This CAQR - received three escalations for failure to obtain PORC and Plant Manger approval of corrective action within 60 days of the.CAQR date of origination and as of this inspection was not resolved.

SQP 87 1277 Temperature Indicator Failed to Respond to Test Signals.

The CAQR required three escalation notices before PORC and Plant Manager approval of corrective action could be attained.

SQP 87 1238 Essential Safety related Motors May Trip

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Due to Transient Overvoltages When Connected to the Onsite Power Supply.

This CAQR required three escalations before required ap,nrovals of the corrective action could be attained, A required generic review was not performed, but was beg;n in

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f response to the NRC inspector's finding.

SSA 88 102807 Review of Significant CAQRs Identified at Watts Bar Are Not Accomplished in a Timely Manner at SQN.

The CAQR review for significance was not complete at the time of this inspection.

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Timeliness is an issue in this CAQR because the programmatic ability of the CAQ process itself is questioned in this CAQR.

SQP 87 3743 Diesel Generators are Incapable of Supplying Adequate Voltage to Stroke Certain Valves in Required Time _

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This CAQR was _ initiated on December 22, 1987, but not reviewed by - a mananement reviewer until January 12, 1988, exceed ng

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the 3 working-day requirement.

The root'

cause appeared to be unjustified.

SQP 87 1195 125 Volts Direct' Current (VOC) Polarity Reversed fue Two Flow Control Valves.

The scheduled-corrective action completion date was December 14, 1987, which remained

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unfinished cn February 10, 1988.

The CAQR was in escalation on February 1,1988, for

- being greater than 30 days late.

The escalated action itself was two weeks late.

SQP 88 0003 Inadequate Post-maintenance Testing Following Replacement of Diesel Hydraulic Actuators.

The CAQR was processed by the management reviewer on December 18, 1987, but not i

forwarded to the CAQR Coordinator until January 5, 1988, exceeding the 3 working day requirement.

The reviews for root cause, recurrence control, and generic concerns were late by approximately two weeks.

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SQP 87 1684 Inadequate Surveillance Instruction Failed to Include Testinc of All Combinations of Circuitry Logic Required by Technical Specifications (TS).

The root cause determination was 'merely a restatement of the problem description.

The problem was first identified as a PRO on March 25, 1987, and not as a CAQR. As a result, corrective action was incomplete.

On December 4, 1987, a CAQR was opened to

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correct this deficiency.

The use of a PRO in this case delayed the process.

SQP 87 1441 Shutdown Board Room Chiller Unit Not Wired j

per Design But Will Function.

Corrective action time limit exceeded the

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allowed 7 days, SQP 87 1704 Air System to Ensure Main Steam Isolation Valve (MSIV) Closure in 5 Seconds, Per TS j

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not Environmentally (EQ) or Seismically Qualified.

Management review for operability and corrective action exceeded'the time limits.

SQP 87 1724 Hydrogen Analyzer Lack of Redundant Power.

The time limit to determine significance was-exceeded.

SQP 87 1731 Deficiency in Red Lining of Control Room Orawings.

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The management review requirement time limit of 3 days was exceeded.

This CAQR was written December 2, 1987, and at the time of this inspection no corrective action had been proposed or approved.

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SQE 87 172802 Inadequate Processing of CAQs, TE-aanagement review and corrective action i

t limits were exceeded. The significance cae ;orization of this appears to be in error, based on repeated failures to process CAQRs.

SQF 87 0150 RHR Pipe Rupture Impact on Local Structures.

i The management review time limits were'

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No action was identified following the scheduled December 14, 1987,

completion date and no documentation of.

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completed action was made available to the inspector.

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SQP 87 1356 Inadequate Seismic Support for Unit 2 Lower

Containment Compartment Cooler.

The management review time limits were

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exceeded.

SQP 87 1672 W2 Switch Unreliability On Component Cooling Thermal Barrier Pumps.

There is no documentation that the corrective actions specified prior to switch

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change out were performed.

Intermediate corrective action requires continuity checks

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on the switch.

It appears there is no

' method of tracking intermediate corrective action in the CAQR process.

SQP 87 1404 Design Changes Without Design Control Measures.

This CAQR addressed the possibility that minor design changes viere performed without adequate control measures as a result of the-work request system.

R' commended corrective e

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action by the initiator was more in line with the potential problem, but the final approved corrective action was minimal and appear to be inadequate.

The accepted proposed disposition does not address the extent and root cause evaluation and appears to be inadequate.

Excessive time delays were experienced in precessing this CAQR.

b.

CAQRs That Were Generated by the Division of Nuclear Engineering or a site other than Sequoyah and were applicable to Sequoyah.

Note:

This category of CAQRs contains issues which predate the present Quality Assurance Program corrective action provisions.

These issues may be identified as Non-Conformance Reports (NCRs) or Significant Condition Reports (SCRs).

For the purposes of this review these issues will be treated as one group even though a redundant CAQR number may not have been assigned by the licensee.

The following CAQRs/NCRs/SCRs were reviewed. The inspector verified that:

reporting and operability requirements had been met, causes had been identified, corrective actions appeared appropriate, generic applicability had been considered, administrative timeliness requirements had been met, licensee management had reviewed the event, no unreviewed safety questions were involved, and no violations of regulations or Technical Specification conditions had been identified.

In addition, for those items requiring hardware, drawing and/or procedural changes in the plant the inspector verified that:

redundant components were operable when corresponding components or systems were removed from service, approvals were obtained prior to initiating the work, activities were accomplished using approved procedures and were inspected as applicable, functional testing and/or calibrations were performed prior to returning components or systems to service, and quality control records were maintained and can be retrieved.

(1) The following CAQRs/NCRs/SCRs were adequate with respect to the acceptance criteria in sub paragraph 10.b. abov. _ _ _ _ _ _ _ _ _ _

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l SCR WBN EEB 8590 Non QA Values of Cable Weights.

SCR WBN EEB 8612 Extreme Complexity in Using Watts Bar Environmental Data Drawings Cause Repeated Errors in Interpreting The Drawings.

SCR WBN EQP 8619 Square D Handswitches Located Inside Containment and in The Steam Valve Rooms May Come Apart During. Accident -Conditions (high-temperature).

SCR WBN 7132-S-R0 Installed Medium Voltage (VS) Cables Do Not Meet the Adjacent 3-Phase Circuit Bundle.

Spacing Requirements.

WBN 87 0308 Drawing Discrepancy.

WBP 87 0745 E.aedded Boxes.

WBN 87 0316 Drawing Discrepancy.

BLP 87 0165 QC Personnel Welding Qualifications.

BP 87 0659 Exide Emergency Battery Pack Units.

WBP 87 0721 Flex Hose Qualification.

BFN 87 0522 Main Control Room Vertical-Benchboard Type Control panels.

WBP 87 0724 Piping Analysis.

WBP 87 0691 Post Modification Functional Testing.

BFQ 87 0668 Drawing Discrepancies.

WBP 87 0760 Stress Ratios.

BP 87 0995 Limitorque Valves With Crc:ked Keyways.

WBN CEB 85 08 Incorrect Valve Operator Attachment Clamps.

WBP 871207 Potential Failure of Operator-to-Valve Engagement on Xomox Supplied Valves.

SCR SQN CEB8602 Inadequate Justification / Documentation for Design and Seismic Qualification of HVAC Ducting.

WBP 88 0040 Requirements for Establishing Qualification for Cable Tray Fitting. -.

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WBP 87 0197 Records Accountability for QA Activities.

WBP 87 0067 Chemical Volume Control System'(CVCS) Piping Reinstallation Not Inspected as Required.

WBP 87 1097 Three-Conductor Cables Installed in. Field.

Vice Two-Conductor. Cable. Required by Drawing-.

WBP 87 0597 Improper Changes Made to QA Documentation.

This item was designated as significant and required generic review, however admini-strative error resulted in the misconception that the review was sent to Sequoyah.

This mistake was discovered on January-27,1988, (audit by QA) and the error corrected. This resulted in delaying the generic review by Sequoyah.

WBQ 87 0209 Failure to Document Training of Personnel Performing Quality Related Activity.

(2) The following CAQRs/SCRs/NCRs were inadequate with respect to the acceptance criteria in sub paragraph 10.b. above.

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WBP 87 0821 CVCS Excess Letdown Piping.

This CAQR exceeded the management review and escalation time limits and is currently

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unresolved.

WBP 87 0481 Epoxy Cable.

This CAQR exceeded the corrective action time limits.

BLF 87 0141 AFW Pump Motor Circuit. Breaker Control Power.

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Sequoyah engineer disagreed with Bellefonte required corrective action.

A question was raised in the CAQR whether or not this was an NRC requirement. This CAQR.

applied to Sequo,ah but the "not applicable" box was checked.

This raises a basic programmatic question of resolution of differing technical judgements between sites.

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WBP 87 0818 Loading Conditions on Seismic Structures The review sheet was checked that it did

. apply to Sequoyah but a PIR was listed because DNE was sure Sequoyah had all aspects of the problem covered. The PIR had not been issued as of January 18, 1988, which was 157 days after initiation of the CAQR and in excess of the time limits that would have been allowed if this issue was processed using a

CAQR.

Sequoyah procedures require a CAQR to be issued if generic issues apply.

The draft answer to-the PIR does not address all items adequately. A CAQR was being written at the -

time of this inspection.

SQP 87 1648 Four Diesel and Electrical Panel Ventilation Fan Select Switches are Wrong Type EMZ vice W-2.

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The root cause/ recurrence, generic review required and operability blocks were all marked. W-2 switches are installed-but are improper in that they have 4 positions instead of 3 and the switch indication arrow does not align with the nameplate in the

"start" position.

The Plant Modification Group has initiated a WR to rework the switch p1ns in order to eliminate the 4th position.

Root cause and recurrence determination were inadequate.

WBP 87 0900 Valve Actuators Were Not Per Vendor Design.

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This CAQ was. issued at Watts Bar on September 15, 1987.

A generic review was required and issued to Sequoyah on October 21, 1987.

Sequoyah performed an evaluation and determined the issue was not applicable to Sequoyah. This evaluation was issued on January 15, 1988.

The response was 52 days overdue.

Although the licensee did not determine a j

generic problem existed at Sequoyah, i.e.,

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interchanging Limitorque motor operators with the unique valve it was matched to, they have established an acceptable maintenance program to test and correct

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A. comprehensive safety-rel'ated Motor-Operated. Valve (MOV) program for Lvisual inspection, lubrication, and testing wa s -.

developed and will 'be performed for Unit 2 during the. cycle 3 ' outage.

Motor-0perated Valve Analysis and Testing System (M0 VATS)

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equipment has' been~ purchased, and Sequoyah.

engineers have been trained by the MOVATS company. in its use.

Testing is still required, and will be' completed prior to unit 2 restart for MOVs with modifications not' complete and. for. MOVs. awaiting resolution of thrust values.

Concurrent with this testing, a historic -

data base for each valve.is being developed for future reference.

This history will consist of the M0 VATS signature in addition to other information.

With the exception of the untimely response to the generic review, the other items reviewed were found. acceptable.

SCR WBN CEB 8684 Inadequate Supports Shown on Typical Valve Support Drawings.

A Potential Generic Condition Evaluation

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i (PGCEM) memo was issued November 21, 1986.

j The evaluation at Sequoyah had not been

performed as of February 9, 1988.

This evaluation is approximately 1 year late.

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NCR W417-P Instrument Line Separation.

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The original NCR was issued at Watts Bar on June 2, 1986.

At that time it was not specified for generic review.

Subsequently a second NCR, NCR 7148 Rev. O, was issued on January 16, 1987.

NCR 7148 referenced NCR

W-417-P, Rev. 1.

It was subsequently upgraded to an SCR 7148 Rev. O and issued with an engineering report dated March 23,

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1987. The engineering report specified the generic implications would be defined in the responses to PIR WBN EEB 8713. The PGCE was

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issued January 29, 1987.

The response from i

Sequoyah, dated July 30, 1987, stated the instrument separation condition did exist and was specified on the restart

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determination form as a restart issue.

From

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the time, the issue was discovered a't Watts Bar on June 2, 1986, until it'was determined

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to be a problem 'that actually existed at Sequoyah, approximately 14 months elapsed.

The inspector evaluated the. licensee's action relative to correcting any deficient conditions found as a result of the identified problem.

A walkdown inspection was performed using CAR-87-014, Instrument In:pection Test, which identified numerous areas where the 18 inch minimum instrument line separation existed.

Acceptance criteria and. items requiring inspection were performed as identified in Special Maintenance Instruction (SMI)-0-317-61

"Instrumentation Feature Walkdown, Rework and Inspection Instructions for CAR 87-014."

The licensee's walkdown identified 36 separation violations. Sequoyah engineering personnel performed evaluations of the deficiencies and identified those which required rework prior to restart.

The licensee issued CAQR's'SQP 87 0478, SQP 87 1099, and SQP 87 1148 to establish corrective actions.

Fifteen separation discrepancies were identified as restart items.

Rework was performed by Work Plan, WP12533 and FCR5718. Completion of the work

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was documented on CAQR SQP87 0478.

The items reviewed were'found acceptable.

SCR WBN 6716-S-R0 Silicon Rubber Insulated Power and Control Cable With Overall Asbestos Braid Jacket Purchased on TVA Contracts 83999, 825997, 430134, 825018, and 822502, Did Not Meet Requirements for Use Inside Primary Containment for Class 1E Harsh Environment Applications.

This issue was resolved as a restart item outside of the normal CAQR process.

TVA staff indicated that both Sequoyah and Watts Bar personnel could not produce documentation that a generic review had been conducted at Sequoyah or that any information on the subject had been transmitted to Sequoyah.

The SCR was - a construction SCR at Watts Bar and was marked as having generic applicability.

The "yes" and

"no" block for potential generic

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condition evaluation required were left blank on the approved copy of the SCR. Four contracts were with Anaconda and one was with American Insulated Wire and Cable.

Because the generic review for this issue was not done or could not be located, TVA staff indicated that a generic review would be done. Inspectors informed TVA. staff and management that satisfactory completion of this CAQR was required prior to Sequoyah Unit 2 Startup.

CAQR WBN 87 0806 Electrical Drawings Do Not Agree On the Trip Settings of the Same Breakers.

This CAQR was initiated at Watts Bar nuclear plant on August 20, 1987 and transmitted to Sequoyah on November 20, 1987. The CAQR had generic implications and was not received at Sequoyah until 57 days after the CAQR review date. This left only 13 days to meet the 70 day requirement for a generic evaluation by Sequoyah. At the 70 day point, the CAQR had to be escalated to the Assistant Site Manager to assign a Sequoyah organization to conduct the review.

At that time, DNE,

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Electrical Modifications, Electrical

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Maintenance, and Maintenance Planning had declined the review. At the 128 day point, on December 28, 1987, a

first level escalation was sent to DNE with a required

30 day response. Discussions with TVA staff indicate the CAQR is presently undergoing a j

second level escalation nearly 6 months af ter the CAQR was written.

This was of particular concern as generic reviews can i

result in operability questions.

R Three escalations were ineffective in resolving the problem and, as of the date of this inspection, a generic review had still not been conducted.

i SCR WBN EEB 8724 Radiation Monitoring System.

This SCR incorporates several documented

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problems (SCRs, NCRs, PIRs, etc.) relating

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to numerous discrepancies in the design, application, and installation aspects of the radiation monitoring system at Watts Bar j

i nuclear plant.

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'25 The SCR was evaluated as having generic applicability and sent to Sequoyah on February 7,1987. A generic. evaluation was conducted-by Sequoyah on April 13, 1987, which indicated 'that the Radiation Monitcring System at Sequoyah had many of the same problems as the Watts Bar system.

The review indicated that numerous additional walkdowns, investigations, and-evaluations were required.

NRC Inspectors concluded the generic review conducted by TVA was incomplete in many areas.

In those areas, the generic review indicated the need for a walkdown and evaluation of sample lines, additional research, and evaluation

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of several problems that-were known to be present.

The identified problems combined with the potential for additional significant unknown problems raises serious question as to the operability of the entire radiation-monitoring system.

The Sequoyah generic review was vague in many areas and uses phases such as "do not appear to be present", "problems believed to be fixed", "maintenance section believes the problem had been fixed", "This work needs to be re-investigated and restarted", "some work was done in looking to replace this monitor", etc.

Inspector concluded that, while the review indicated that many Sequoyah problems existed, it may have been inadequate in determining or cueing CAQRs/

SCRs that could have effected system and component operability.

Therefore, the

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inspector concluded that the documentation and completeness of this generic review was inadeauate.

CAQR WBN 870892 R1 Rajiation Monitoring System.

This CAQR was written to incorporate numerous CAQs, SCRs, and PIRs on the Radiation Monitoring System for reporting, identification, and handling.

The CAQR identified 13 major areas of problems, deviations, or non-conformances with the radiation monitoring system at Watts Bar.

The Sequoyah generic review

listed the 13 areas and referenced Sequoyah

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SCRs, CAQRs, drawing deviations, etc. for the areas that applied to Sequoyah and concluded that 6 of the 13 areas were applicable to Sequoyah. The generic review listed CAQRs that had been written to document problems specific to those areas.

The generic review did not contain edequate justification for the remaining 7 areas that were considered not applicable to Sequoyah.

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This was considered significant as the 7 areas had considerable overlap with the areas needing walkdown, investigation, and evaluation in the generic review from SCR WBN EEB 8724 above.

Inspectors reviewed a.PR0 and LER status printout which indicated that about 90 PR0s and 15 LERs had been written within the last 12 months on the radiation monitoring system. At the time of the inspection 39 CAQRs and PR0s remained open on the system.

The number and type of problems indicated by PR0s, CAQRs, and LERs appears consistent with the problems initially discovered by Watts Bar and partially confirmed by generic i

reviews at Sequoyah.

The inspector concluded that system reliability is questionable.

System and component operability appears to be_ depen-dent on the detailed documantation of walkdowns specified in the generic reviews.

The extensive nature of the problems and the multiple TVA organizations having cognizance over those problems has resulted in a fragmented approach to corrective action.

CAQR WBP 870759 Safely Related Structural Steel Platforms.

The licensee in a memo (Hosmer, Abercrombie)

dated January 23, 1988, RIMS 833880114850 determined that administrative control of live loads on these structures was not acceptable.

This issue is still under DNE/ Plant review without acceptable corrective action in place.

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11.

Directed Licensee Corrective Actions as a Result of NRC Order EA 85-49 and the TVA NCR Nuclear Safety' Review Staff (NSRS) Report.

As a result of the order described in the Summary section of this report the licensee was directed to complete certain corrective actions within specific periods 'of time.

These. directed actions and their outcome are listed in sub paragraphs 11 a. through 11 d. below.

a.

Within 60 days, the licensee shall complete an evaluation of its procedures at each of its operating nuclear power plant sites and at its Office of Engineering in Knoxville, Tennessee with regard to their adequacy for ensuring that when-potentially significant safety conditions are identified by engineering management such as the Chief Nuclear Engineer (Nuclear Engineering Branch Chief), they are immediately reported to plant management, evaluated expeditiously, for appropriate action, including applicability to other plants, reported if required, and corrected.

Corrective action by the licensee appeared to comply with this requirement.

b.

Within 60 days, the licensee shall submit the evaluation to the Regional Administrator, Region II with a copy to the Director, Office of Inspection and Enforcement, along with a plan and schedule 'for promptly revising the procedures as appropriate.

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Corrective action by the licensee appeared to comply with this -

requirement, c.

Within 120 days, the licensee shall develop and submit tu the Regional Administrator, Region II, with a copy to the Director, Office of Inspection and Enforcement, a plan for training of all personnel involved in implementing the revised procedures including responsible licensee management personnel both in the Office of Engineering and the Office of Nuclear Power to ensure aat such personnel recognize potentially significant safety conditions and ensure that they are expeditiously evaluated, reported, and corrected and understand their individual responsibilities in carrying out the procedure.

The plan shall provide a schedule for when the training will be completed for all of the employees and managers.

Initial corrective actions by the licensee were adequate. However, l

recent changes in site management resulted in some key managers not having auditable records of training and/or testing in the required areas.

d.

Within 45 days, the licensee shall provide the Regional Administrator, Region II with copies of all reports, evaluations or

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other analysis that may have been prepared of the circumstances

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surrounding, including chronology of events, the qualification issue of the pressure transmitters at Sequoyah between October 1, 1984, and April 1, 1985.

If investigations have been conducted or are ongoing that have not yet been completed this shall be indicated and an expected date when the documents will be provided.

In addition, within 45 days the licensee shall survey all of its Office of Engineering (0E) employees and Nuclear Power (NUC PR) employees as well as any other appropriate employee.

Corrective action by the licensee appeared to comply with this requirement.

In summation, although the above licensee corrective actions.with respect to the order and the NSRS report were acceptable, the review conducted in this report. of the implementation of the licensee's problem identification, resolution and corrective action systems indicate that it is not appropriate to consider the removal of order EA 85-49.

Specific issues were identified to the licensee and are discussed in the summary paragraph of this report. Those issues will be inspected prior to restart of Sequoyah Unit 2.

12.

List of Abbreviations AFW Auxiliary Feed Water CCWS Component Cooling Water System

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CAQ Condition Adverse to Quality CAQR Condition Adverse to Quality Report CVI Containment Ventilation Isolation CVCS Chemical Volume Control System DCR Design Change Request DNE Division Of Nuclear Engineer ECN Engineering Change Notice EQ Environmental Qualifications ERD l Essential Raw Cooling Water FE/ J Failure Evaluation / Engineering Report HVAL Heating Ventilating and Air Conditioning HELB High Energy Line Break IN IN Information Notice LER Licensee Event Report LCO Limiting Condition for Operation L^'A Loss of Coolant Accident MSIV Main Steam Isolation Valve MSLB Main Steam Line Break MOV Motor Operated Valve MOVATS Motor-0perated Valve Analysis and Testing System NCR Non-conformance Reports NER Nuclear Experience Review NUC PR Nuclear Power NQAM Nuclear Quality Assurance Manual

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NRC Nuclear Regulatory Commission NRC-OSP Nuclear Regulatory Commission - Office Special Projects l

NSRS Nuclear Safety Review Staff OE Office of Engineering f

OSP Office Special Projects PIR Problem Identification Report PRO Potential Reportable Occurrence PORC Plant Onsite Review Committee PAM Post Accident Monitor PGCEM Potential Generic Condition Evaluation Memo PGCE Potential Generic Condition Evaluations PM Preventive Maintenance PT Dye Penetrant QA Quality Assurance QC Quality Control QI Qualified Individual RIMS Regulatory Information Management System RCS Reactor Coolant System RWST Refueling Water Storage Tank RHR Residual Heat Removal SCR Significant Condition Report SQN Sequoyah Nuclear SNP Sequoyah Nuclear Plant SCR Significant Condition Report SMI Special Maintenance Instruction SI Surveillance Instruction TS Technical Specifications TACF Temporary Alteration Change Form TVA Tennessee Valley Authority URI Unresolved Item 0500 Unreviewed Safety Question Determination VIO Violation VDC Volt Direct Current WBN Watts Bar Nuclear WR Work Request l

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