ML20244C391

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Insp Repts 50-327/88-50 & 50-328/88-50 on 881212-890126. Violations Noted.Major Areas Inspected:Plant Operations, Surveillance & Sys Outage Control,Corrective Action Program, Maint Activities & Qualified Reviewer Process
ML20244C391
Person / Time
Site: Sequoyah  Tennessee Valley Authority icon.png
Issue date: 04/04/1989
From: Brady J, Elrod S
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20244C347 List:
References
50-327-88-50, 50-328-88-50, NUDOCS 8904200179
Download: ML20244C391 (61)


See also: IR 05000327/1988050

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- n REGION 11 - p 5" L ,j 101 MARIETTA STREET.N.W. E '2 - ATLANTA, GEORGIA 30323 - _ ._; ,o ' ..... -Report Nos. 50-327/88-50,' 50-328/88-50 ' Licensee: ' Tennessee' Valley Authority ~ .6N 38A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

Docket Nos.

50-327 and 50-328 License Nos.: DPR-77 and DPR-79 Seq'oyah Units 1 and 2 Facility Name: u Inspection Conducted: December 12-16, 1988 and January 9-26, 1989 Lead Inspector: } 8LA - />1 h Mn 4/f/89 f.;A.JElro Tdm L(ader Date Signed Team Members: W. C. Bearden; Resident Inspector. B. R.-Bonser, Project Engineer J. N. Donohew, Senior Project Manager . 'G. E. Gears, Senior. Project Manager R. D. Gibbs, Reactor Inspector G. T. Hubbard, Branch Chief D. P. Loveless,' Resident Inspector . 'G. A. Walton, Senior Resident Inspector ' Approved by: M[ag[ 9M[87 @f B. Brady, ddting Chief, Date Signed ~ ~ TVA Projects Section 1 TVA Projects Division Summary Scope: This routine, announced inspection involved inspection onsite in the areas of. pla'nt operations; surveillance control, system outage control, and work control prccesses; the corrective action program; maintenance activities; plant modifications process; implementation 'g - and verification of commitments made to the NRC; QA, QC, and QM activities; the qualified reviewer process; and followup of events. The inspection reviewed quality and quality verification in relation H to' the ability of line management to get workers to accept responsibility for doinc quality work, the ability of the quality assurance and quality monitoring organizations to verify quality by audits and surveillance, and the ability of the organization as a l 8904200179 89041o , gDR ADOCK 05000327 l PDC l ___-_ -

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whole' to' _verifyf and , accept" quality at the interfaces between groups ~g .when deliverables.are transferred. " Results: Onei violation was identified for. failure to take adequate corrective caction to preclude repetition of previous violation 327, 328/87-30-01 . involving lack of control over ' plant . evolutions', and system and equipment status 'in the radioactive waste area, . paragraph- 5 (VIO !327,~328/88-50-01). .Four unresolved items * were identified. 1. Trending within ACPs and the appropriate . thresholds for entering .the CAQR; process, paragraph 7 (URI 327, 328/88-50-02). 2L Failure to include vendor torque requirements. in maintenance - ' instructions,. paragraph 8 (URI 327, 328/88-50-04). 3. . Completion of workplan review and the reporting of repairs and replacements made under the- requirements of ASME Code Section XI, paragraph 10 (URI 327, 328/88-50-05). 4. Engineering evaluations of vendor manuals, paragraph 5 (URI 327, 328/88-50-07). It was concluded that site line management was strongly' dedicated to quality and was convincing workers that quality work was what was expected. One exception was found in the radwaste processing area. This was revealed by the resin transfer problem discussed herein. The events indicated that management attention had been lacking and that overall site procedure upgrades had not had an affect on upgrading quality in . this area. SQN has not yet completed the indoctrination of personnel regarding accepting the responsibility of doing quality work; however, thi s inspection found substantial improvements, compared to past years, in the quality awareness of personnel. The function of the quality monitoring organization was to assist site management in meeting quality objectives by iden ti fyi ng conditions adverse to quality on a real-time basis before they impacted on nuclear safety, reliability, or component operability. This inspection concluded that the quality monitoring organization was a well qualified, adequately staffed organization which was performing its function well. , Unresolved Items are matters about which more information is required '

to determine whether they are acceptable or may involve violations or deviations. ' ! , ' ' ' ' ' ' ' ' ' ' L_-_-_a-____ - .

, ._. _ - _ __ __ _ _ _ _ _ _ _ ' . A:, .: . ' 3 { 7 v ~ The use of interfaces between groups and by the organization as a whole, to verify -and accept quality when deliverables were trans- ferred ' was not emphasized- as 'a quality verification tool. For example,..the Maintenance Department was using an interface organization between the shops 'and QA to ensure thatLcompleted surveillance tests represented quality work prior to their transfer to - QA for review, however, some of the problems that were being identified for correction had resulted because procedure changes had. not bee'n adequately, communicated to the shop organization responsible for performing them. An interface problem was also identified between engineering and the plant in relation to vendora manuals having conflicting data ' and resulted from a lack of communication between the two organizations. These examples are discussed L further in the. report details. Although interface problems between engineering and the plant were identified by the NRC staff in an earlier . inspection report (327, 328/87-52),. interfaces were still ~ not . actively 'used by site or corporate management for quality verification purposes. , In order to reduce the enormous amount of intense upper management effort necessary to make the CAQR system work, the licensee developed a change to the CAQR process and implemented it in September 1988, c immediately prior to the restart of Unit 1. The change provided . ' several ACPs- to act as corrective action screening processes.. Those - issues not meeting the acceptance criteria for being a CAQR stayed in the ACPs for resolution. This inspection concluded that the changes were adequately implemented and strongly supported by senior line management. The changes appeared to have the desired effect of forcing insignificant and less significant issues down to the proper level for resolution, while keeping safety significant items at the senior management level. The inspectors found SQN's process of plant surveillance control, system outage control and work control to be well established, controlled by procedures, and working. The process provided for the identification of work needing to be performed, establishment of procedures to do the work, scheduling of'the work, performance of the work, and tracking of the work to completion, r = - - _ _ - _ - _ _ _ _ _ . _ - _ - _ _ - _ _ _ - _ _ _ _ _ - - _ _ _ _ . _ _ - _ - _ _ _ _ _ - _ . _ _ _ _ - _ . _ _ _ _ _ . _ _ _ . _ - _ _ _ _ _ - _ _ _ _ _ _ - - _ _ _ _ _ _ _ _

_ . _ - - _ , l i .. . , ~ } 1 REPORT DETAILS I 1. Persons Contacted Licensee Employees 4

  • J.

LaPoint, Site Director j S. Smith, Plant Manager i J. Anthony, Operations Group Supervisor, POTC K. Allen, Periodic Test Coordinator

  • T. Arney, Quality Assurance Manager
  • W. Aslinger, Asst. Site Rep. Employee Concern Program
  • R. Beecken, Maintenance Superintendent

H. Birch, Unit 1 Work Control Supervisor

  • J. Blackburn, Special Projects Engineer

G. Boles, Manager, Maintenance Planning and Technical

  • E. Boyles, Site Rep. Employee Concern Program

L. Bryant, Program Support Manager S. Chapman, Supervisor, Document Closure

  • M. Cooper, Compliance Licensing Manager

D. Craven, Plant Support Superintendent ! I. DiBase, Environmental Qualification and Preventive Maintenance Program Supervisor H. Elkins, Instrument Maintenance Group Manager R. Fortenberry, Technical Support Supervisor H. Gammage, Site Procedures Manager J. Hamilton, Quality Engineering Manager

  • R. Hays, Radioactive Waste Processing Supervisor
  • J. Holland, Corrective Action Program Manager

T. Howard, Quality Assurance Surveillance Supervisor S. Johnson, Quality Assurance Technical Support Supervisor D. Jones, Shift Outage Manager N. Kazanas, Vice President, Nuclear Quality Assurance

  • D. Kelley, Waste Water Processing Group Manager
  • J. Klein, Maintenance Trending and NPRDS Supervisor

R. Lewis, Modifications Engineer C. Lonas, System Evaluator

  • J. Maddox, Engineering Assurance Lead Engineer
  • L. Martin, Site Quality Manager

T. Matthews, Licensing Project Manager

  • D. Michlink, Coordinating Project Engineer

R. Miles, Modifications Manager L. Parscale, Manager of Licensing Performance J. Patrick, Operations Group Manager

  • J. Petty, Shift Technical Advisor Supervisor

R. Pierce, Mechanical Maintenance Supervisor G. Putt, Work Control / Outage Group Superintendent M. Ray, Site Licensing Staff Manager

  • J. Robinson, Modifications Manager, Engineering

F. Roemer, Nuclear Quality Assurance Engineer . , _ _ _ . . _ _ . _ _ _ _ . _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . . _ . _ . - _ - _ _ . _ . _ _ _ _ . _ . - _ . _ _ . _ _ _ _ _

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  • R. Rogers, Plant Support Superintendent

R. Shell, Compliance and Services Manager 1

  • S. Spencer, Licensing Engineer

T. Spink,l Replacement Items Program Manager M..Sullivan, Radiological Controls Superintendent { D. Thomas,' Impact Evaluator l

  • P. Wallace, Site Programs Manager

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  • J. Walker, Manager of Operations . Support.and Procedures

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  • J. Ward, Outage Scheduling Supervisor

J.. Wilder, Senior Engineering Specialist. '*L. Wheeler, Materials Manager C. Whittemore, Licensing Engineer

  • A."Wilkey,'Sequoyah Quality Audit Group Manager

NRC Employees

  • F. McCoy, Assistant Director for TVA Inspection-Programs
  • L. Watson, Section Chief

K. Jenison, Senior Resident InspectorL

P. Harmon, Senior _ Resident Inspector L 'P. Humphrey, Resident Inspector

  • J. Brady, Project Engineer
  • Attended exit interview

' NOTE: Acronyms and.initialisms used in this report are listed in the last paragraph. 2. Introduction - This special, announced NRC team inspection at the Sequoyah Nuclear Plant was performed to evaluate the acceptability of the line and QV. organizations' activities and . management's support of these activities. This inspection was performed primarily under the guidance of NRC ' Inspection Manual Chapter 35702, " Inspection of Quality Verification Function". The inspection consisted of personnel interviews, direct observation'of in progress activities, and review of work documents. Line organizations must be aware of quality requirements and must be confident of their ability to perform to the required level of quality. If the QV organizations are technically credible, they can and should help define identified deficiencies, provide insight into the root cause '3 ._ of deficiencies, and approve and confirm the resolution of deficiencies in 'a ; technically meaningful way. The inspection also assessed line

management's ability to ensure that identified deficiencies are dealt with promptly and completely. Quality Verification Function Inspections are not intended to verify licensee compliance with administrative controls; they are intended to verify'the technical adequacy of safety-related activities. However, if deficiencies are found in these activities, the underlying procedures and

_-_ - -__ - -__ I ' .E.. gn t 4 qy 1 "'#,- . y o , administrative controls 'are reviewed.' The. intent of these inspections is- > to . improve plant operational safety through inspection-processes that' are

focused on activities'that. affect plant safety.and reliability.

r 7 _ The QVFI at Sequoyah primarily focused ~ on plant operations; corrective'and. ~ ' preventive. maintenance 'of . plant systems and components; and the . control. , system 'for' surveillance, system outages, and other work. A ' secondary . ' focus.was -the completion of and verification of commitments :made to the -

' LNRC. -Several program areas, such as ' the corrective action program, .were ' Jalso': inspected. . The inspectors reviewed selected documented examples- in . these and closely associated areas to _ identify safety-significant problems (if any) ~ to .be used as ' vehicles for evaluating the - effectiveness - of- qualityJachievement, _ self-verification, and other verification. The results'of,this review.are discussed in'this report. . L3.

In spection .0bjec'tives':

' The primary. objectives of this inspection were: Assess ~the effectiveness of the licensee's various line organizations- . 'in achieving and self-verifying quality in their. functions. Assess ' the . " quality verification effectiveness" of the 'other .QV -- organizations: in identifying,- resolving, and preventing safety-significant ' problems and deficiencies 'in various functional areas. I? For this assessment, " quality verification effectiveness" is defined as the . ability of the ~ 1icensee to verify quality and to identify, 1 - correct, and prevent problems It is not limited tc the licensee's -Quality Assurance Organization, but is the aggregate of all efforts to verify quality results and take corrective action when a quality result was not obtained. It specifically includes the line organizations. Assess the effectiveness of the _line management in ensuring that - safety-significant problems and deficiencies are dealt with promptly and. completely, in response to input from the QV organizations. - Assess the effectiveness of communication of generic issues within the QV organizations and line management. Assess the effectiveness of communication of lessons learned from - other TVA sites. - Assess the ef fecti /eness of engineering support to correct ' deficiencies. { A secondary objective, closely related to site performance and verifica- ) tion, was to assess the performance and verification effectiveness related 1 , - - - _ - - - - . _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ - _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ . _ _ _

n A > 1a Yl.' ! 4 , toEthe commitments made to. the NRC which were prerequisites for ' allowing - the return to operation. 4. Inspection Scope: The inspection was divided into the following functional areas: Plant Operations page 5 -- Surveillance Control / System Outage Control / Work Control Processes - -- page 16- Corrective-Action Program page 21

' - , Maintenance' Activities page 26- - ~ - Safety Information Management. System / Corporate Commitment Tracking -System - page 35- Plant Modification Process page 38 - -QA Routine? Audits / QC Activities / and Special Surveillance (or - monitoring: activities) in Support of Operations page 47 Independent-Qualified Reviewer Process page 50 - l5. Plant Operations At the beginning of the operations inspection, both reactor units were operating' at power. To better grasp the workings of the Operations Department. and the QA/QM programs related to operations, the inspectors met briefly with Operations and QA management. To detect previous problems and identify potential repetitive problems or- problem trends,1988 documentation from various problem tracking systems was reviewed. The documentation review included TROI printouts - both open and closed items, PRO printouts, selected QM reports, and the NRC Open Items List. The documentation review revealed there had been a signi f.i cant number of configuration control and operational status problems prior to and duriag the restart of Unit 2 in late 1987 and the first half of 1988. The problems had the following characteristics: - These problems centered around maintaining cognizance of the opera- tional. status of critical systems, structures, and components. This included status records of valve power supplies, instrumentation, penetrations, and structural components. The problems had occurred during normal plant operation, during the - performance of surveillance instructions, and during maintenance. I ~ ______-__________-_-_Q

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_ -5- . O , ' .Some:of the'more significant problems found, which the licensee indicated .

were corrected, were:

' > . , J Then-current: procedures. and system checklists did not_ always match - existing as-configured prints and existing system, configurations. J. Procedures which' called for closing valves with the plant in Mode 4 would' prevent the centrifugal charging _ pumps.from having their normal - ' cold leg : injection path to ' the .RCS. (See CAQR SQP 880151 discussed 'below). Some of' the ERCW- valves. maintained by- SI-682 were be'ing moved from - - 2 ' their required positions without 'any documented reason. (See CAQR: . SQP 880213 discussed below)'

Conflicting' valve positions between' an SI for system 67 and a GOI,

c- . coupled with J the valves being found out of positioni during SI- performance.'(See CAQR SQP 880490 discussed below) , ' ' Valves-_ _ being omitted- from a valve- . check 11st could cause an: - inappropriate valve alignment. ' 1, . Not- correctly maintaining configuration control logs in the control '- , , room. . Failures -of the system ' alignment corrective action program to -- eliminate SOI checklist inadequacies prior to restarting system alignments. l . A check of NRC inspection reports showed that some of ~ the above-listed ~ problems were identified as a result of NRC findings: 47 .NRC inspectors had reviewed the licensee's control of system - configuration status by auditing logs and recently-completed SOI- checklists and walking down' the. RHR and UHI systems. Four examples of failure to properly implement procedures' associated with controlling plant configuration were identified. (IR 327,328/88-26) l. - Spills of primary coolant water occurred as a result of misconfigured systems. (IR 327, 328/87-24 and 87-30) [. ' Examples of components being out of position. (IR 327, 328/87-66 and ' - ib' 88-06) (See CAQR: SQP 880245 discussed below) u L Specific CAQRs were examined more closely for details and corrective l actions: p. L SQP 880151 - - Procedures GOI-1 and G01-3 had BIT isolation valves - LL 63-25, '-26, -39, & -40 isolated while the plant was in Mode 4 - ! preventing a (requi~ed for Mode 4) flow path from the centrifugal charging pumps to the normal, cold-leg, injection path to the RCS. e , j: ' - _ ,- ..

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.- l . . I , The corrective actfons were checked and appeared to be sufficient and complete. SQP 830213 - Many performances of SI-682 over the past several years - l~ had shown unexplained deficiencies. ERCW FCVs were being moved from l required positions. The corrective actions given in the CAQR were checked and appeared to be complete and satisfactory. . a w SQP 880245 - This CAQR was written in response to NRC violation 327, - 328/88-06-02. The CAQR involved a failure of the system alignment corrective action program to eliminate SOI ' checklist inadequacies prior to restarting the system alignment process. ' Corrective actions for some.of the specific examples cited were checked. The corrective actions appeared to..be sufficient and complete. - SQP 880414 - This CAQR. involved.the use of status boards to maintain configuration control . The corrective action included revision of AI-58 to delete the use of stat 0s boards for configuration control and require the use of a configura} ion log. - , . .ve . .; SQP 880490: - This CAQR - involved conflicts between procedures - pertaining to the positions for certain ERCW valves. The corrective action was checked and appeared to be sufficient and complete. - SQP 880504 - This CAQR was written because the waste disposal system (liquid) checklist . did . not contain certain system valves. This review;found thev1isted valvesnto inow be included in the system checklist; The review of these - CAQRs indicated to the inspector that previous configuration control problems had received significant appropriate attention and that the corrective actions given in the CAQRs were sufficient and had been completed. A review of recent licensee quality reports- in the configuration / status' control area ' identified no significant problems. Having completed the past problem review for background 5 information, current activities were reviewed. As part of . the. review of this area, the NRC inspectors reviewed the conduct of operations activities. NRC inspectors spent some time in most plant locations including the control room. This effort included normal working hoursp backshift, and shift turnover. This time was primarily spent directly observing operational activities during power operation on both units and included routine shift activities and the performance of surveillance testing. Since configuration control had been a previous problem, this review was sensitive to that subject. . <i. w n , t Control room activities were generally conducted in an efficient and professional mannere Formal and clear communications were observed in most cases. Personnel traffic and. noise levels in the control room were _ _ - - _ _ _ - _ -_ - - _ . _ _ _ . i

_, _ _ - - . (- ... m , L ... 7 m controlled such that 'they did not interfere with routine shift' activities. ' . Operators were attentive, aware of, plant conditions, and remained in their. -designated areas. ' Shift manning was in accordance with TS requirements and proper relief.was observed prior to shift ~ personnel being discharged from their duties. Major scheduled activities appeared to' receive the , proper amount of. management review and planning prior to being included in [ the daily scheduled list of work activities. l' ' The 111censee. had; established ' systems to maintain cognizance of. system . status and maintain configuration control. The effectiveness - of the present systems and their conformance to the appropriate Als was reviewed. A System Status File was ' set up in the control room for Units 1 and 2. Critical systems were aligned as required by appropriate valve and power availability' checklists following the last outage. The existence and . maintenance of these files was verified. As evidenced by their corrective- action- in the .CAQRs, the licensee had expended significant effort in assuring that the checklists included all valves and electrical components. However, some problems still existed -as evidenced by .the finding, while transferring spent resin, of a waste disposal- system interface valve that was not listed in a checklist' er shown on approved drawings. .When conditions required deviating from the normal system ' alignment, the deviations were to be entered in a Configuration Log. The Configuration Log and the System Status File should reflect the current status of critical systems. The Units 1 and 2 control room Configuration Logs were checked. No deficiencies were noted. As described in AI-58, Maintaining Cognizance of Operational Status - Configuration Status Control, the log was indexed by system numbers corresponding to the numbers given in the - AI, which was. maintained in the unit horseshoe area. The log entries appeared to conform to the AI. The Test Awareness Log was reviewed. The Test Awareness Log is used for ongoing activities controlled by approved procedures that provide both configuration changes and return-to-normal within the procedure. These activities were not entered in the Configuration Control Log. A review of the entries in the Unit 1 Test Awareness ' Log showed one deficiency. Maintenance Instruction MI-20.12, Calibration of Bourdon Tube Pressure Indicators, had been started on January 15, 1989, at 7:55 am, had not been closed out. The Unit 1 SOS indicated it was an error and the log entry should have been closed out. The. Critical Valve Summary Checklists in the Unit 1 and 2 control room areas were reviewed. The list is utilized to provide verification of ' proper valve alignment each shif t on safety related systems to ensure system operability in the applicable modes. The inspector's review of the checklist revealed no deficiencies. The Unit 1 UD log was reviewed for configuration control practices. One minor deficiency was noted on the 2300 - 0700 log for January 16-17, 1989. t i % . _ - _ _ . - _ . _ _ _ - - _ . .- _ _ _ _ . , - - _ _ _ - _ _ = _ . - _ _ _ , - - - _ _ _ . - - - . . . _ _ _


_,-___---._______-__-___________-__-.--_-_--_-.____.__.__.--..____._---.._---_-____..-w

y I rw. - . . . . kn; ' n . 8. F An LCOLwhich was entered and logged.for changeout of'a radiation monitor , = filter.was not logged when;the changeout was complete, however,' completion of the. activity.was-logged. Elapsed time from beginning to end.was about' eight' minutes. . .The Unit 1 LCO Action' Log was reviewed. The LCO. Ac' tion Log was not n, treated as' a .QA record but was treated as an operator. aid. The UO log discussed'above provides the QA record. No deficiencies were noted during " this review. The inspector's. overal1Li.mpression of the configuration / status-control systems established by Lthe licensee was. good. However, even in the small- sample size reviewed by the . inspector, minor deficiencies were noted. . .The111censee. was : developing a pilot program .for use during the Unit. 2, l Cycle 3 refueling outage. The program will utilize a TS Condition Report'- to track out-of-service TS: components or systems when the plant is:in an operational mode where that TS would not be presently applicable or where g - the. TS .would ' become more conservative later in a different- mode of operation.. The. program effectiveness in ensuring TS compliance will be evaluated during future inspection activities. Tours of .the Unit 1 and'2 Auxiliary, Control, and Turbine Buildings were ~ conducted to observe plant conditions', including general- cleanliness / housekeeping, potential fire hazards, leaks, and adverse equipment operation. The NRC inspector noted that cleanliness and housekeeping were excellent. No significant material or equipment problems were noted, however, one concern was noted in that all six'AFW pumps appeared to have excessive packing leakage while the pumps were idle. This concern was discussed with licensee management who stated that no acceptance criteria for leakage exists and :the packing glands were intentionally maintained loose to increase packing life and decrease shaft sleeve wear DCR 2565 has been written to replace these pumps' packings with mechanical seal assemblies. Since the first pump is scheduled to be modified during the upcoming Unit 2, Cycle 3 outage, the NRC inspector believes that the licensee has adequately addressed this concern- l In' addition to the routine system configuration control set-up, control must be maintained during the performance of sis which involve manipulation of the plant. To prevent recurrence of mispositioneo valves, the licensee formed a dedic.ated operations SI team. The formation of this team limited the number of persons performing operations department SI tests, increased each person's exposure to the sis and enhanced the internal communications within the group. The inspector specifically observed the functioning of the SI team during surveillance SI-5 and interviewed team members. The SI team concept appeared to be effective in improving efficiency and control. The NRC inspectors observed portions of the following surveillance tests: - . _ _ . _ _ _ _ - -_..__-- _ - _ _ -

_ _ _ _ - _ _ - ___ _ _ - _ _ - _-- .. . 42 w ' , 9' SI-2, Operations Shift' Log. SI-2, Operations Shift Log SI-3,. Operations. Daily Log SI-3, Operations Daily Log. .' SI-5, : AFW Valve Position Verification SI-37.1, Containment Spray Pump 1B-B " Quarterly Operating-Test SI-166;1, UHI Isolation Valve Full St'oking- r SI-744, Monitoring of UHI Isolation Valve Accumulator Pressure During the conduct of the surveillance. tests, ' operations ~ personnel responsible' for conducting the tests, whether from the SI . team or: not, appeared- to have an excellent knowledge of the sis and the associated plant systems. The instructions reviewed contained an . adequate level: of information and technical detail' to allow proper performance. All' personnel involved in the ' performance of the testing attended a' . pre-evolution briefing and all SI prerequisites were satisfied prior to starting each test. ' The testing was properly coordinated and applicable steps were followed verbatim during actual performance of the testing. The NRC inspector reviewed numerous completed QM reports covering- operations activities that had been performed since the creation of the site QM organization. QM reports . reviewed' included: configuration control, temporary alterations, TS surveillance, . operator logs, shift - relief and turnover, shift compliment, and plant staff overtime. Additionally, the.NRC inspectors reviewed completed quality audit reports SQA 88-808 and SQA 88-815 which covered a variety of operations-related activities. The review of QM reports and quality audit reports identified no significant problem areas. However, past control of temporary alterations and the accuracy and detail contained in past operator logs were identified repeatedly by QM and audit reports as weaknesses that warranted greater management attention. The NRC inspector noted that operator logs have since improved and that the control of temporary alterations has recently received improved management attention. An operator journal, or log, should contain a narrative of the plant's status and all events required to provide an accurate history of plant operations as stated in AI-6, Log Entries and Review. Although improvement has been noted in the detail included in the various logs maintained by the operating shifts, a reviewer has not always been able to i~ determine the full history associated with a given event from the 00 log or SOS log. Often it has been necessary to refer to several other logs and charts to reconstruct the actual conditions that led up to a particular event. The NRC inspector believes that an increased emphasis by management and the QM Group would result in continued improvement. Although log entries ' were not the subject of inspection findings identified in previous NRC inspection reports 327, 328/88-17, 88-35, and 88-39, the lack of good _ ' ' ______.m_.-- ____._----- . _ _ . _ _ - _ _ . _ _ _ * _ _ _ . _ .- .__._..2.2,...m . m.

_ - - - - - _ - - _ _ _ _ u ' (. ' i l . 10 ' I detailed. log entries contributed to the failure to perform an adequate post trip review in previous Violation 88-35-01 ~and problems associated j with operators unknowingly ' entering. LCOs in previous URI 88-17-02 and < Violation 88-39-02. i The licensee has had a long term problem with the control of temporary alterations. Abuse of the TACF system in the past had resulted in ] many modifications being controlled under AI-9, Control of Temporary . Alterations, long after the short time period intended for conversion had elapsed. This resulted in a large backlog of TACFs and had the potential to -strain administrative control effectiveness. This problem had been identified' in NRC Inspection Reports 85-46, 86-20, 86-27,.87-08, 88-47, and 88-51. The. problem was also discussed in two recent QM reports and the Febuary, 1988, ISEG Monthly Report, 88-02-SQN-I. As .of' December 1986, approximately 200 TACFs still remained active for both units. - Although a management action tracking system required routine review for the purpose of justifying continued need, as of May 1988, 138 active TACFs'still existed, and many TACFs issued prior to 1984 continued " to remain in place. The licensee had committed to INP0 to clear all temporary alterations that were in place on January 1, 1984, before Unit I startup following Cycle 4 completion. In spite of this, the problem appeared to receive inadequate ~ management attention until the later half of 1988. During July - August 1988, two monitoring reports were issued which identified various administrative errors such as incorrect entries in the TACF index and missing TACF forms. Additionally, one of the monitoring reports identified that as of July 18, 1988, 90 TACF reviews were delinquent. To ensure TACFs are properly reviewed and closed in a timely manner, AI-9, Section 7.2.2, requires that a review by the responsible section supervisor be documented on Attachment 6A. . Forty-four of the 90 deliquent reviews were on CSSC Systems. This deficiency was documented in CAQR SQQ 880420. The NRC inspector reviewed the corrective action for this CAQR and determined that it was adequate. The recent revision to AI-9, if followed, should preclude recurrance of this problem. Recent improvement in licensee management attention in this area resulted in further reduction of the backlog. AI-9 recently enderwent a .inajor revision. Changes included a decision to reassign responsibility as implementing organization from Planning and Scheduling to the Technical Support Group Superintendent - who will perform periodic reviews of all installed temporary alterations. Additionally, all proposed TACFs and extensions for existing TACFs must be approved by the Plant Manager. PORC review will be required prior to installation of any TACF on CSSC o. equipment or for extension of a TACF installed on CSSC equipment. As of L January '18, 1989, 94 active TACFs exist and the licensee has developed a work-off schedule for the remaining TACFs to meet the INPO commitment. This- schedule includes budgeting, for FY89 and FY90, for DNE and the onsite modifications group to close all but six of the remaining TACFs by - - - _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - - _ _ _ _ _ _ - _ _ _ _ - _ -

= ,__ _ _ _ _ L-

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I . .. . ' ' -11. ym. . I f [ September 1990. . These six TACFs are associated' with' future UHI removal and temporary thermocouple in Unit 2 steam valve' vaults. Alth'ough:the licensee appears to be on .the path toward' correcting this' l j deficiency, a large amount' of. time .was allowed to pass without adequate - management' attention in this' area. On January 11, 1989, the licensee attempted to' transfer spent resin from " - the .2ACVCS mixed-bed demineralized to the' SRST -using SOI-77.3, Rev 15', l Section'B'- Waste Processing - Spent Resin Storage Tank. The. resin' failed , to transfer.. completely. Radiation dose rates in: the' vacinity of the ' transfer lines. varied but ranged as high as 90 Rem pe'r hour near one valve. It was believed'.that a commonly-used diaphragm. valve in the , transfer line might have failed with a ruptured diaphragm, blocking the line. Because. ~ of 'the unusual result ,of the transfer attempt, NRC inspectors reviewed controlling' literature and licensee recovery activities to determine the extent of quality attainment and quality. verification Literature reviewed included:

SOI-77.3,LRev.15, Waste Processing,. Sections B and C - . MI-11.7.1, Rev.6, Hand Operated Grinnell or Saunders Type Diaphragm- - Valve Rebuilding.for All Systems ' Aproved vendor manual SQN-VTM-I207-0010, Vendor Technical Manual for - . In'sta11 ati on and Maintenance Instructions for Nuclear Diaphragm Valves. - Numerous PM requirement sheets for diaphram valves SOI-77.3B was found to be unuseable as a category."A" verbatim compliance procedure ICF 89-0035 was written under detailed management direction, field verified, and approved prior to use' in the recovery. S0I-77.3C was also found to be unuseable and not followed - resulting in a personnel contamination, which is discussed in more detail below. CAQR SQN 890016 was issued for the inedequate procedure. The vcendor manual had two conflicting sections addrest.ing the same valves. Items in confl?ct included body-to-bonnet bolt tcrque valts and whether or not to use torque values or just tighten the bolts l finger-ttght. CAQR SQN 8900026 was issued to resolve this issue. t . Although the PMs and MI-11.7.1 appeared to be sound, having been written i with informal vendor consultation, t. hey did not implement all of the requirements of the above VTM. DNE had not documented evaluations for these deviations. CAQR SQN 890026 also addressed this condition. At the end of the inspection, DNE representatives indicated that it may be acceptable to issue conflicting information and let the maintenance staff pick the version to use. This is an indication of an interface problem between DNE and the plant staff in relation to vendor technical manuals. Interface problems between DNE and the plant had been previously identified in IR 327,328/87-52 pertaining to the use of compensatory ! measures. I l -,

- - _ _ - _ -_ . . - . H , - l- 12 l - Based on the above-described findings, the status of vendor manuals as the l basis for plant technical instructions is unresolved pending licensee, and subsequent NRC, review. This is URI 327, 328/88-50-07. l On - January 16, 1989 an AVO was contaminated during the performance of l SOI-77.3C, Waste Processing - Transfer of Spent Resin from SRST to l Shipping Liner. Numerous procedural errors, failures to follow procedures, and operations outsic'e of procedures were common in the S01-77.3C performance. The review of the incident highlighted the following problems: - Instruction step "B" required the performer to initiate a WR to install a jumper on contacts SCG-3 and SCG-2 in junction box 3017 to allow 0-FCV-77-225 to remain open during resin sluicing. This step was- performed without using a WR. The design of 0-FCV-77-225 was to automatically close upon a high level indication signal from 0-LE-77-225. This level element was not in current use at the plant. A temporary level indicator was used in the shipping liner instead. Previously this indicator was jumpered out via a TACF. This TACF was terminated and step "B" added to control the subject jumpers. Step "G" later re-verified that this was accomplished by requiring the performer to review the procedure prior to loading the shipping liner and to verify that the jumper had been installed by the WR initiated in procedure step "B". Lack of a WR or design change to control these jumpers is an example of inadequate configuration control. Step "E" required the performer to obtain a SRST level reading on - 0-LI-77-48. The SRST water level indication was poorly designed such that it was unreliable at higher tank pressures. Therefore, the operators would open the tank vent, 0-FCV-77-51, to reduce the pressure prior to taking true tank water level readings. This knowledge was not discussed in the procedure nor did the procedure allow tha opening of the vent valve for this purpose. - Step "F" required the performer to open 0-FCV-77-45, fill the SRST to

approximately 10-15's above the resin level, then close 0-FCV-77-45.

2 Step "F" was not performed because there was no indication of SRST resin level. Additionally, as addressed above, the water level indication for the SRST was unreliable. The operators explained that the step was accomplished Dy first draining the tank watar through 0-FCV-77-698, then slowly filling the SRST while alternately opening nitrogen valva 0-FCV-77-46 and venting via 0-FCV-77-51. When the point was reached that the tant quickly pressurized, they knew that the void space was steall and, therefore, that the tank was almost

full. This, they claimed, provided the basis for the assumption that the resin was covered with water. The licensee later determined that the SRST was completely full and overflowing with the resin. Appropriate performance of this step would have determined this problem. This is an example of inadequate instrumentation to accomplish the procedure requirements. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . - - - - -)

_- .. . . 13 Step "H" required the performer to open valves 0-77-953 and 0-77-845. - Valve 0-77-845 did not exist in the plant. Operators simply crossed out the valve number in SOI-77.3C and-wrote in those valves they wanted to open. This was accomplished without approval or knowledge of the control room operators. This is an example of inadequate i configuration control, inadequate procedures, and failure to follow procedures. Step "J" required the performer to test the shipping liner level - transmitter after setting up to load the liner. This evolution involved multiple valve manipulations not addressed in the procedure. This is an example of an inadequate procedure. - Steps "N" and "0" required the performer to perform the following manipulations: ! N. Pressurize SRST to 80 psig, then open in order. (1) 0-FCV-77-300 PW to flush (2) 0-FCV-77-225 Spent resin tank to liner (3) 0-FCV-77-49 Spent resin tank outlet O. When back flow is noted: Open 0-FVC-77-226 Close 0-FCV-77-300 (Liner is now filling) NOTE: (1) Maintain Nitrogen pressure at 80 psig during transfer Procedure steps "N" and "0" wvre adequate to perform the evolution by pressurizing the tank, back flushing the lines, and then reversing flow in the lines to sluice the resin to the liner. However, the operators in the radioactive waste crganizatior, did not understand the procecure's approach. '1 hey, therefore, took steps to pressurize the tank after aligning for the sluice. This it an example of fnilure to follow procedures. - Step "P" required the performer to watch the liner level transmitter. When all four lights are on, the liner is full Then perform steps Q, R, and S in rapid succession. ! During this evolutien, the operator only transferred a small quantity of resin to the liner and did not attempt to fill tne liner. The operator knew from his training that the resin had to be sampled prior to filling the liner. The movement of a small quantity of resin, the sampling process and the analysis of that sample were not proceduralized. This is an example of an inadequate procedure. . _ _ - _ _ _ _ - _ _ _ _

_

..

- . 1 ... ' 14 - ' u Step "S"'. required the performer to close 0-FCV-77-226, Liner Fill - Valve, when water flow to the liner was observed. As'an alternative, j the operators 'and HP technician determined when the dose rates in the line dropped, indicating that the : resin had' passed, because. the temporary'. liner level indication was not sensitive enough to. indicate the onset of' flow by a changing level. This is an example of an . inadequate' procedure. Step'"T" required the performer. to close 0-FCV-77-225, Spent Resin '1 - Tank -Isolation Valve, when flow to the SRST was observed. The operator stated that 0-FCV-77-225 would actually be closed shortly after 0-FCV-77-226 was closed because flow indication to the SRST was . unreliable. .This is an example of an inadequate design. Following .the perfo'rmance.of step "T", S01-77.3C assumed the lines to be clear of resin.- -Actually, following the flush, the HP technician detected a hot spot of approximately 15 Rem per hour on the backside of valve 0-FCV-77-400. The operator told the inspector that this usually occurred during resin transfers from the SRST to the shipping liner. The operator .then disconnected the dewatering pump discharge line from the ~ fitting at 0-FCV-77-401 (to the tritiated drain tank) and~ connected it to the fitting upstream of 0-FCV-77-400 (resin dispensing header-to auxiliary contract [ equipment]). He then opened valves 0-FCV-77-400, 0-FCV-77-225, and 0-FCV-77-226; started the dewatering pump; and began to recirculate the water in the liner. This' was accomplished by taking a suction on the dewatering vanes inside the liner, discharging into the auxiliary contract header through valves 400, 225 and 226 and then back into the liner. This evolution flushed the resins from the hot spot and into the liner. This entire evolution was performed outside of an approved procedure and is an example of an inadequate procedure and a design deficiency. Following this flushing evolution, the operator, again without a procedure, isolated the valves acd disconnected the discharge hose from the fitting. This breaching of the system was not first discussed with the continuous-coverage HP technicians. Upon disconnecting the hose, residual pressure in the lire caused water to spray on the walls and floor, and c' contaminated the: operator's face. The operator was subsequently decontaminated. This is an example of equipment and personnel hazards generated from ' performing evolutions without approved procedures. This contemina- tion would likely have besn prevented had appropriate procedures been in place. The inspector noted that during the performance of 501-77.3C. the operator had not signed several steps as being complete prior to being contaminated. The operator's replacement, noting that he had accomplished these steps, signed them off. This is a violation of i L 1 b. . _ _ . _ _ . _ . . _ . _ _ . _ . _ _ . . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ ____._______________________________._____________w

, 'i . . 15 , , AI-4, Preparation, Review, Approval and Use of Site Procedures / Instructions. This is an example of failure to follow procedure. The licensee management replaced the supervisor of the radioactive waste organization and temporarily revised 501-77.3 to enable them to move the spent resin and flush the hot spots. The inspector observed licensee activities during this recovery. Several times, the operators from the waste organization suggested that evolutions be performed that would violate the procedure. These evolutions were not performed because the new group management determined them to be unacceptable. This suggested to the inspector that, prior to the management changes, the entire group lacked adequate training, supervision, and self discipline to safely perform the operations. Violation 327, 328/87-30-01 was written in July 1987, to address a lack of control over plant evolutions and the status of systems and equipme'nt. The NRC issued the Notice of Violation to obtain the management attention necessary to resolve the underlying problem. This previous citation noted specifically that: a. The unit operator did not use the formal change process of AI-4 to revise SI-166.3. Instead, a system realignment was improvised without written or formally-approved instructions. This example is similar to many of the current examples noted above, including steps E, F, H, J, N, 0, P, S, and T. b. Five valves were shut in an attempt to isolate the SG maintenance area from the RWST without entry in the configuration log as required, and vent valve 1-HCV-68-594 was opened without this deviation from the normal valve alignment being entered in the configuration log. In the current examples, valves were re-aligned during these previous evolutions without procedure or configuration control, as noted in steps E, F, H, J, N, 0, S, and T. ,c. On December 2, 1966, December 14, 1986, and May 22, 1987, SI 4S.1 was performed without appropriate instrumentation to verify the j flow rate. In tne current examples, level indications in the solid radioactive waste (spent rer.in) system were inadequate and flow 1 indications did not exi st to perform the steps of S01-77.3C as written, at noted in steps E, F, S, and T. The licensee's response to the previcus Notice of Violation committed, in part, tc the following actions: ' i Perform procedural adherence training that emphasized the require- a. ments and reasons for procedural compliance to over 900 employees. _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ -

__ . . _ _ s. . ., o.. 16 , b .' Discuss'the.importance of configuration control and failure to comply with procedures regarding these [ previous] events during the " Operations Lessons' Learned Sessions. l' NRC Inspection Report 327,328/87-50 re-opened NRC review of ' violations. !' resulting from a 1984. thimble tube ejection event,- including violation 84-24-01. Violation 84-24-01 addressed failureLto establish and implement- procedures for ~ the conduct of equipment control, procedure review and .approva , per orman e of maintenance, radiation -work permit access con- l f c trol, and access to containment. The NRC was concerned that perhaps the corrective actions taken for violation -84-24-01 had been inadequate to i preclude subsequent violation 87-30-01 discussed above. In a letter dated November 8, 1987, the licensee submitted a revised response to Violation 84-24-01. In this response, the' licensee stated othat procedurals adherence was now required by AI-4 and that additional - training of plant personnel had been- performed on procedural adherence. Additionally the licensee stated that this training corrected problems with individuals not obtaining procedure changes prior to performing work not specifically addressed ~in the procedure. 10CFR 50, Appendix B, Criterion XVI, Corrective Action, requires that, for significant conditions. adverse to quality, corrective measures determine the cause and preclude repetition. Prior to January 16, 1989, the ' licensee failed to preclude repetition of Violation 327, 328/87-30-01 . af ter having ' completed corrective action. Lack of control over plant- evolutions, system status and equipment status was still evident in the radioactive waste area. As a result: , a. Multiple evolutions were performed outside of approved plant procedures during spent resin transfers from January 11-19, 1989. b. Known inadequate drawings were utilized affecting procedural controls of the temporary resin interface valve. ' c. Known design deficiencies were not corrected causing recurring radiation hot spots in excess of one Rem per hour to be created. The conditions existing, cembined with evolutions beir.g performed outside of approved procedures, allowed an unwarranted personnel contamination. .' This is identified as Violation 327, 328/88-50-01. Eh', 6. Surveillance Control / System Outage Control / Work Control Processes I To understand how work at SQN was controlled and quality verification in the line organizations was accomplished, the inspectors interviewed SQN 1 J personnel who were involved in the preparation, scheduling, and control of I plant work. In addition to the interviews, the inspectors observed samples of their in progress work. l ' 1 1 - _ _ _ _ _ _ _ _ _ _ _ . ___ _ -___-_-_-____ _ _ _ __

_ a; - < .. 17 At SQN, surveillance. control, system outage control, and work control were three: interrelated ~ functions performed by- the Work Control / Outage . organization, the Outage Scheduling organization, and the Maintenance V

Planning and Technical organization. These. organizations coordinated - their; respective activities to ensure that required work within the plant,. i.e. , sis, WRs, and _ PMs were identified, scheduled, and performed as req'ui red. , ' Work Control /0utage' Control Organization'n a. The: Work ' Control / Outage Control organization . was responsible for scheduling when the work was- to be performed in . the plant and assuring that the work was accomplished. These functions were accomplished. utilizing System Evaluators (SE), Work Schedulers, Shift Outage Managers, and Impact Evaluators. , , .SEs were' assigned to specific systems in a specific Unit (Unit 1 or l 2) and were . responsible for maintaining cognizance of . work being performed on tneir assigned systems. Duties of the SEs included: review of WRs,- processing of WRs to the required organizations. .for planning, determination of when work required removal of equipment , ' from plant -service, . and when equipment . tag outs were required, coordination of work on assigned systems with work schedulers and the craft organizations performing the work, and ' ensuring the work was completed. -Work schedulers were responsible for the actual scheduling of the work start and duration. The schedulers utilized a computer listing of work available to track the work required to be completed. Using this information, the schedulers, through coordination with craft organizations and the SEs, established schedules for work performance. For periodic work having pre-established procedures for performance, e.g., sis, scheduling was controlled by the " Periodic Test Coordinator" organization. This organization was responsible for ensuring that periodic tests were. includec in the PDWL, which was differem f or each shift of each day and was irsued three times each day. The PDWL also included other significant information relative to plant activities for each shift including identification of major work activities of the previous shift and major work activities for the onccming shift. In addition to ensuring that periodie work 1 activities were scheduled, the periodic test coordinator organization was responsible for ensuring that the periodic work was indeed ] completed as required.

Shift outage managers kept track of the work scheduled to be performed each shift and the work actually accomplished.. The shift , ) outage managers utilized shift turnover meetings to update oncoming shifts of the work previously accomplished and the work scheduled to be accomplished during the oncoming shift. During these turnover ) ' - _ - - 1 .

. _ - _ _ _ _ _ _ _ - m .7 m Hg . 18 m meetings, the shift managers provided copies of the PDWL. . As discussed above,'these lists included. scheduled sis as well.as major work-' activities -completed during the: previous shift and those scheduled for the oncoming shift. Impact evaluators were responsible for reviewing-work which was' to be performed in the plant to evaluate the work's effect on the. plant status. While- the SEs were concerned with the effects of work on their assigned systems, the itpact evaluators, who were SR0s, were concerned with whether work on one . system would adversely affect another system, or systems, .resulting in unplanned changes in the plant status, e'.g., a reactor trip. . The Work: Control / Outage. organization and other groups also utilized several data bases in addition to the PDWL to track work and ensure ~ , , it was. completed. The" data bases included the " Prime" and "P/2" data bases. for. tracking 'and following work to completion. Some of the lists utilized were the Preliminary DWL, the DWL, the Site POD, the Site Weekly WR Summary, and the STORM, which was used to track outage work, b. Outage Scheduling Organization The Outage Scheduling organization was responsible for the tracking 'of sis, PMs, and WRs; providing advance notice to the Work Control / Outage organization when work-should be scheduled; and either providing work packages or ensuring that approved work packages had been provided to the organizations performing the work. In addition, this organization was- responsible for outage schedule development. Another function of the organization was to maintain a history of maintenance activities that had been performed on plant equipment. The NRC inspectors' interviews of personnel and in progress work observation focused on work in the support of sis, PMs, and WRs. The scheduling organization utilized PM coordinators, WR coordinators, and SI schedulers to keep data bases up-to-date regardir.g the plant work to be performed in their respective areas of re spon sibil ity. In addition to keeping the data bases current, these personnel either put together the work packages, e.g., PMs, for the craft people to follow in perforr.41ng the work or ensured that the packages, e.g. , WRs, had -received the required processirg by others before being provided to the craft people for use. , ' c. Maintenance Planning and Technical Organization This organization was responsible for writing working procedures for plant work when specific procedures had not been previously written. The planners were required to be knowledgeable of how to perform work - in their area of expertise, e.g., mechanical or electrical equipment, the equipment configuration in the plant, and the plant design and technical specification requirements. If they were uncertain of the ,> m . - - _ . _ _ _ _ _ . - _ _ . _ . _ _ - - _ _ - . - _ - . _ . . - _ - -

r-

p l. a j: 4- I t 3 3 . 19 .. required work, they had access.to the latest approved plant drawings ' .and other approved plant requirements. In 'some cases, the planners may have lto go to the worksite to. see the required work. in order to write adequate procedures'. In addition. to writing- work procedures, the planners worked with . the maintenance craft persons reviewing the work procedures to ensure that they understood-the work and apeed that it- could be - accomplished as: written in the - procedures. . Aod'tionally, if ? the , craft people were ' to . run -into .significant proble.Ls during the ' performance of a' procedure, the craft may have to stop work and have the planner revise - the work procedures. - Any revised procedures of - i -this type ' would still be required to undergo strict. approval processing' before' . implementation of the work. ' If minor procedure. problems were to. be identified during the performance ofa' work . package, the. craft may modify the activities, document the change's on .an ICF, and proceed with the work. These ICFs would be then- routed back to. the planners so that the problem can be. corrected in future activities, d. LDocument Closure In additional- to the organizations discussed above, the inspectors g reviewed the operation of the Maintenance ' Department's " Document Closure Group", where completed SI packages are reviewed for adequacy and completion prior to being transferred to the QA Department for review and . storage as QA records. This group's function was to monitor the interface between the Maintenance Department and QA Department to ensure that quality products are transferred. This review identified that approximately 50% of the packages were being l' ' rejected by the Document Closure Group as not being completed properly; Most of the rejections were considered minor or of an administrative nature. Although the packages were being corrected prior to becoming QA recoros, the inspectors were concerned with this high rate of rejection, aspecially since PMs and WRs were also scheduled to begin sienilar review by the: Document Closure Grcup in early' 1989. This group was considered a.c an example of a positive use of interfaces .for quelity verification. L Tne inspectors discussed this problem with mainter.ance managers who assured the inspectors that the Maintenance Department was also L concerned with the high rejection rate and was working to resche the h problem. It developed that one of the root causes was that the group p responsible for the administrative instruction had changed it to L require speciff e data on the coversheet turned in with sis, but had ' not changed . over 1200 SIS - each with their own coversheet - to conform. 'The maintenance - shops were then be'i ng blamed for the inability to perform procedures. This was a management control l ' problem and indicated that interfaces between maintenance groups within the Maintenance Department were not being effectively used as a quality verification tool . SQN appeared to be taking action to resolve it and no specific NRC follow-up is planned. i J ! {" ._ _ _ __

- - . - - - - - _ - _ _ - - _ - - - - - _ - _ - - _ _ '[

,, f ' ' '

.: ., z. :. . 20 .. e. Site Procedures Staff ' During the inspection, the inspectors learned that.SQN had. initiated a new. program to improve site procedures, including sis and PMs. The l inspectors ' reviewed the program as it existed during the inspection L and the plans for the future. Th'e program included standardization

and computerization 'of procedures as well as an extensive database L development ' effort based on existing database information. This

program was part of an .overall TVA effort to standardize procedures-

throughout TVA and to make them more user friendly. The program at SQN was scheduled to be completed in 1991'. f. Summary The NRC inspectors. performed the .following tasks in evaluating- the control.of work within the SQN plant. ' Reviewed documents and procedures associated with the work - Li control activities. discussed above (including SI-1, Surveillance Program . the controlling document for the execution of the -plant surveillance program). Interviewed personnel in the organizations discussed above - concerning their job activities and observed the performance of -job activities by some of the personnel interviewed. - Observed demonstrations ' of SQN data bases related to work control and the information contained in the data bases. - ' Witnessed the performance of SI-37.2, " Containment Spray Pump 1B-B Quarterly Operability Test," following installation of a new pump rotor. The inspectors found SQN's' process of plant surveillance control / system outage control / work control to be well established, contro*tled by procedures, and working. The process provided for the identification of work needin;; to be performed, establishment of procedurer to do the work, scheduling of the work, performance of the work, and tracking of the work to completion. . . The inspector found that, within the organizations discussed above, H each organization faad estabibbed internal controls to verify the quality of it's own vork. Tho SQN work control process also required s > which provided extensive coordinat' ion between organizations - additional checks to verify the quality of the work being performed. T The SQN process provided for corrections to problems as the work was being performed. In fact, while the inspectors witnessed the performance of SI-37.2, workers noticed a packing leak on a valve used during the SI and documented the condition on a WR for correction. o . - - _ - - . _ _ _ - _ . - _ _ . _ _ _ - - _ _ - - - - . _ _ _ - _ - . _ _ - . _ _ _ . _ . _ - _ - _ . - - . _ _ - _ _ - _ _ - _ _ _ - - _ - - - - - . - _ . _ . - - _ _ _ _ _ _ _ . - _ - _ _ . - - - - _ _ _ _ _ _ _ _ _ . _ . - _ _ _ - - _ - _ _ . - _ _ _ . _ _ _ . _ . - _ . - _ - - _ .-._ _ _ . - . - - . _ _ _

__ , , - a . 21 In addition to the quality verification activities of the organi- -Iations discussed above, the quality assurance organization was involved in the approval of certain specific sis, WRs, and PMs, e.g., work being performed on TS equipment or systems. As further verifi- cation of the quality of the work, QA performed audits of the various organizations to verify the quality: of their work. At the time of the inspection, QA had performed audits of the Outage Scheduling and Maintenance Planning and Technical organizations and had an audit scheduled for the Work Control / Outage organization in 1989. The inspectors did not review the audits which had been performed. With the exception of the observation relative to document closure (paragraph 6.d), the inspectors did not identify any additional concerns regarding the control of work at SQN. 7. Corrective Action Program The inspectors reviewed the corrective action program as revised in September,1988, and documented in the corporate NQAM Part I, Section 2.16, Corrective Action, and SQN procedure AI-12, Part III, Corrective Action. This program was designed to comply with 10 CFR 50, Appendix B, Criterion XVI. It consisted of the CAQR program and a number of administrative control programs. TVA management was responsible for evaluating adverse conditions ar<d documenting the condition in - the appropriate corrective action program. This review generally emphasized the following attributes: Management Review Process - DNE Support - Involvement of QA/QV in the detection and resolution of problems. - - Operability Determination Root Cause Evaluation -- - Recurrence Control Distribution of CAQRs/PRDs - Evaluation of ACP items and trends to detect CAQRs/PRDs - a. Conditions Adverse To Quality Report (CAQR) Program As part of the inspectors' evaluation of the CAQR program at SQN, they discussed the history of the program with TVA personnel. Prior to February 1987, TVA had a program for controlling - conditions adverse to quality (CAQs) which focused more on design problems than non-design CAQs. This program identified CAQ processing documents under various names such as problem _ - _ _ _ _ _ _ _ _ _ -

mm .. - . -

  • . y

y;. t i . 22 i identification reports or significant condition reports. At the time of this inspection, SQN was still dealing with approx- c ' imately-180 items from this earlier program. LIn- February .1987, TVA/SQN ' implemented the revised corrective - -- action program .to identify all .CAQs as CAQRs. .This program superseded and combined the various earlier programs. While this program was an improvement to. the old system, problems- istill existed relative to the identification of CAQs and obtaining adequate and timely corrective action. In March 1988,

TVA/SQN set up a. management review. committee to help process CAQRs and ensure adequate and timely resolution of identified problems. At. the time of -this inspection, there were ' approximately .400 CAQRs (since February,1987) being tracked at SQN.~ Since this program was being overloaded with quality concerns - which were considered of minor. significance compared to. other ' concerns, TVA/SQN, in September, 1988, revised the procedures to . take credit for administrative control procedures which could control the less significant concerns. These ACPs addressed, but were .not limited- to, such things as MRs, drawing discrepancies, LERs, and responses to external audit reports. -This latest ~ revision provided for the documentation of a problem - (CAQ) on a CAQR-PRD form. Once.the problem was documented on the CAQR-PRD form, it was evaluated by a management reviewer from the organization of the individual who documented the problem. If the management reviewer could not determine whether the problem .was a CAQR or PRD, the reviewer could take the problem 'to the MRC for classification. At the time of this inspection, SQN was taking all submitted CAQR-PRD forms to the MRC to ensure that an appropriate call was being made on identified problems. Even though PRDs were considered to be less safety significant than CAQRs, SQN also tracked the PRDs to resolution to ensure adequate resolution. C ' The inspectors' review determined that both TVA corporate and SQN procedures provided for escalation of CAQs to the highest levels of management, under certain conditions defined by procedures, if CAQ classification and resolution were not considered acceptable by the initiator. The SQN MRC, which consisted of the management reviewers from onsite organizations, met every morning (Monday-Friday) to review and .. discuss CAQR-PRDs. The inspectors attended one MRC meeting on December 13, 1988. The inspectors observed the discussions relative to about a dozen different CAQR-PRDs and found the actions taken by the MRC to be conservative and appropriate to the safety significance of the CAQR-PRDs. SQN personnel pointed out to the inspectors that l l all CAQR-PRDs resolved as " accept as is" or " repair" would be sent to the PORC and plant manager for approval. - - - _- _ - - . _ - - _ _ _ _ _ _ _ - _ - _ _ - - _ _ _

gym ,' , - .. y f.gg,E,^ ', . w b l ' J y2 ' I23 ' , , , , i p As - part off the CAQR program: review, ' the inspectors selected- the i below-listed CAQRs from a computer 11 sting of- CAQRs. These CAQRs 'I

' were reviewed in detail for . appropriate classification and adequacy ~ l .offresolution. The.CAQRs reviewed were: p ?SQP 880532 SQP 880533 S0P 880575 SQP 880592 SQP 880570 SQP 880578 SQE 880557901 SQE 880560901 SQA'880576902 SQA 880577902 SQP 880496 Of note, CAQRs SQA- 880576902 and SQA 880577902 identified,. respec- tively, TVA. audit concerns with the disposition of five CAQRs as > " accept as is" without proper- PORC and plant manager approval . and enine:CAQRs which were classified as PRDs when the TVA auditors.f'elt theyl should have been CAQRs. The' inspectors reviewed.the resolution- or pboposed resolution of these two CAQRs, and the other CAQRs listed 'above, and. found the resolutions to be conservative and appropriate ' to the identified concern. .The inspectors reviewed the following two periodic reports which SQN used to maintain an overview of the CAQR-PRD process: ' Sequoyah Nuclear Plant - Condition. Adverse to Quality Weekly - Report

.

Sequoyah Nuclear Plant - Quality Assurance Monthly Trend Report - This review.found these reports-to provide the recipients (including corporate,.SQN, and other site personnel) a good overview of how the . CAQR-PRD process was working. . The reports provided trending 'information on old ' program CAQs, new program. CAQRs, and PRDs; information concerning areas, e.g. , design, where the problems were occurring; as well as other information deemed significant to controlling the CAQR-PRD process. The inspectors also reviewed ECP Investigation Report - . ECP-88-CH-185-01, dated December 6,1988, from the Employee Concern ' Program. This report raised significant questions concerning "the ability of employees to process conditions adverse to quality in the corrective action program." This investigation, which was conducted from September 14 to November 28, 1988, was initiated because of employee concerns received at ECP of fices at Knoxville, SQN, and BFN. Since these locations were the only locations with specific concerns, S the investigation concentrated on implementation of the CAQR-PRD f

program at these locations. As a result of the investigation, the concerns with the corrective M action process were partially substantiated. One CAQR (CHS880070) was written citing several examples of a CAQR-PRD not being issued " g l , l ..

. _ _ , _ - _ _ _ _ _ u,

s ,

> e . ..- g [ ' ' ' s, t 4 n 24 ' .because~the:effect on quality;could not be determined at the time of . b~ ^ the proposed initiation of the CAQR-PRD. Additionally, the report made 12 recommendations for the TVA corrective action program. The report ' documented that a majority of the concerns in the investigation occurred before- the latest revision of the corrective . action process in September 1988. The'~ inspectors held several discussions with ECP personnel, SQN site R QA' personnel, and corporate NQA personnel regarding the meaning of the investigation findings and 'the ' corrective actions planned or already taken to resolve the : concerns. .The inspectors reviewed a- ' draft of the response .to the ' report from the Vice President, NQA to the Manager, ECP, Based on the discussions and review- of. documents associated with ' resolving the investigation report concerns, the inspectors found that TVA and - SQN. were taking the concerns and recommendations of the report seriously and were actively-working to resolve' theiissues to everyone's satisfaction. The inspectors.noted that some of the concerns identif.ied in the report had been corrected by the September .1988 revision to - the . corrective action program summa ry. Based on reviews and discussions with SQN personnel concerning the ~ CAQR-PRD program, the inspectors determined that, while the TVA/SQN . program was experiencing some implementation problems, the program was working. The inspectors found that TVA corporate and SQN staffs i were taking actions to identify problems-- and to address and resolve

the' problems as they were identified.

That SQN had identified '

problems in the program and was working to correct them was evidence

that-SQN had QV organizations and methods in place and functioning. Methods of quality verification included, . but were not limited to, . trending reports, engineering assurance audits, quality assurance c audits, and ECP investigations. The inspectors had no specific new findings regarding the CAQR-PRD process. In view of the TVA-identified problems being addressed at the time of the inspection, future NRC inspections will continue to monitor the TVA and SQN corrective action program. The inspectors dia ' identi fy problems with the ACPs. These are discussed in the following paragraphs. b. Administrative Control Programs The ACP review consisted of two phases: Review of documents to determine that appropriate licensee - document reviews had been made to ensure that CAQs were not present. The review found no single items which required CAQRs to be written. It was concluded that the individual items in the ACPs had received adequate reviews. = __

_ _ _ _ _ _ _ _ _ _ _ _ Q v .. ..; y "25 ,. A Review .'of. tracking and trending programs to determine ' that - .. appropriate: licensee trend reviews had been made to ensure that- ' CAQs were not present.

The ~ inspector 'revie'wed the ACP documents listed in Appendix' A,

paragraph 2, to determine that appropriate licensee document reviews .had been made to~ ensure that CAQs were not present. The. review found no. items whic.h required- CAQRs to be written. It wts concluded that: the individual items in the ACPs had received adequate. reviews. The inspector reviewed the tracking and trending programs within the. ACPs.' - Each tracking program was documented and appeared to be adequate to follow corrective actions to completion. - No examples of inadequate tracking were found. Regarding the ACP trending programs, the ACPs were. determined to - be part of the CAQR process in September 1988. Therefore, . very little was: available to trend in some programs. Several ACP . trending programs were not established until - January 1,1989. The inspector noted QA involvement in the development and- standardization of the trending programs. These trending programs will .be reviewed further during future -inspections, as part of URI 327,328/88-50-02 discussed below, when more material ~ is available-for review. The inspector reviewed, in depth, the trending program for the WR process. TVA utilized the NPRDS computer program and an in-house program called EQIS to trend WRs. Equipment data and failure analysis from the WR were entered into. the programs' data files. ~ Periodically the programs were asked to look for previously- established trigger points intended to indicate potential trends, e.g., any 8 failures of items from a single vendor in a given year. These trigger point results were then reviewed by appropriate engineers to determine if an actual trend existed or if the trigger represented some other condition, perhaps a procedural inadequacy. Once a trend was noted, the engineer would determine appropriate corrective action and assure that it was implemented. At this point, the maintenance organization was not entering these trends into the CAQR process. AI-12, Part III, Section 2.1.2.E stated that CAQRs would be written for " confirmed adverse trends in activities identified by trend analysis." The maintenance management stated that the above requirement only dictated that they review these trends for meeting other criteria of the CAQR process. At the end of this -inspection, corporate QA and the maintenance department were working together to resolve this possible procedure violation and to determine an adequate threshold for entering the CAQR process. i 4 , . ' ' .

<

. . . a- 26 This item is unresolved pending the results of the corporate QA _ review and will be reviewed, along with recently-started trending programs discussed.above, during future inspections. This is URI 327, 328/88-50-02. The June 30, 1988, Semiannual Component Failure Trending Report was reviewed. It identified 36 components- listed in Appendix A, paragraph 2, as having trends and documented the corrective actions taken. Also reviewed were approximately 150 WRs on the above-discussed and similar components written from July through December 1988. No examples of continuing trends were noted and the corrective actions taken were seen as a positive influence on the system. 8. Maintenance Activities During this inspection, an in-depth review of the maintenance program was conducted. This review included corrective maintenance, the predictive analysis program, and preventive maintenance. The inspection included observation of work in progress and review of the associated work documentation. The inspection also included a detailed review of completed work order packages - including applicable maintenance and calibration procedures, the vendor manual for each component, and associated documentation for the completed work. Completed work order packages were selected based on the importance of the component to plant safety and to provide a cross-sectional overview of the various types of maintenance activities. All work reviewed had been completed within the past two years. The primary focus of this review was to determine the technical adequacy of the work performed. Specific areas reviewed are addressed in the following paragraphs: a. Review of Maintenance in Progress and Completed Corrective Maintenance. As previously discussed, inspection sampling was designed to provide a cross section of maintenance practices. The review of completed work requests focused on the Unit 2 AFW System. Work reviewed ranged from a simple gauge replacement to the overhaul of pumps and motor-operated valves. In general, the following attributes were reviewed: - The work instructions, including the applicable maintenance procedures, for each WR were reviewed for technical adequacy, clarity and inclusion of appropriate acceptance criteria. Adherence to the work instructions and proper sign-offs were also verified. - The vendor manual for each component was compared to the WR and maintenance instruction to assure that vendor recommendations had been properly included. This review also included a

____ a .- . . 27 I comparison of vendor manual recommendations for preventive maintenance to the site PM for the equipment. Torquing requirements for system closure fasteners were - verified. The control of calibrated equipment esed to perform the work was - reviewed. This review included verification that the range of the equipment was appropriate, the equipment was within the current calibration cycle when used, and that the individual calibration records showed that the accuracy of the instrument, , when calibrated, was within the stated acceptance criteria. l - The purchasing documentation for installed materials was reviewed for the inclusion of required regulatory requirements. The receipt inspection records for installed materials were - reviewed to assure that the materials were certified as ready-for-installation. - Traceability of installed materials from the WR to the purchase order and receipt inspection records was verified. - Performance of the proper PMT was verified. In many cases this check also included a detailed review of the results of the testing. The component history was reviewed for component failure trends. - The following specific WRs were included in this portion of the inspection. A brief description of each work item is included and any specific problem areas are discussed so that appropriate corrective actions can be initiated: WR B295059,- AFW pump 2A-A. This work request replaced the i - ' inboard pump bearing due to a degraded condition found during vibration analysis. l No documentation was present to verify that PMT had been accom- i plished following the bearing replacement. Work instruction MI-10.4.1 required performance of SI-130.2 as the PMT but para-

graph 6.5 of MI-10.4.1, for SI performance verification, had been marked "N/A". Licensee investigation determined that SI-130.2 had not been performed, however, a vibration analysis l in accordance with TI-96 had been performed - but no records of this test had been included in the WR. Further investigation determined that the personnel who had marked the PMT requirements "N/A" had failed to follow paragraph 11.3.6 of SQM 2, Maintenance Management System, Rev. 33, which required: "If the section responsible for the PMT makes a determination that the test is not required; a responsible person. . . shall "N/A" o-- - _ - - - - - - - - - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _-.---.--------------_______----.____________________a

_

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. . 28 L l 1 ' the PMT, provide - explanatory remarks, . sign, and date in the appropriate block." Note: Site CAQR SQP890013 was issued to correct this

deficiency.

This deficiency is an example of non-willful failure to follow maintenance procedures. It did not result in degradation of equipment. It occurred in early 1988 along with other .similar problems which- tha licensee recognized as requiring generic corrective action. NRC Violation 327, 328/88-28-01 was later issued concerning a similar situation in May 1988. LER 2-88022 reported that situation. Widespread training was conducted under CAQR SQP 880349. Recent NRC reviews have not found this situation to be common. This violation meets the criteria specified- in Section V of the NRC Enforcement Policy for not issuing a Notice of Violation and is not cited. This issue -is identified as LIV 50-327,328/88-50-03, example 1 of 2, for tracking purposes and is considered closed. - WR B262462, Steam Generator #4 Level Indicator 2-LT-003-107F. This WR'was issued to correct the condition where this level indicator read greater than 6% higher than the other steam generator level indicators. The work involved tightening the fittings on the low pressure sensing line, back filling the transmitter, and performing an operability check. During the review of this WR, it was noted that paragraph 5.4.6 of MI-19.1.5 required the maintenance technician to record the positions of bistables LS-548A, LS-548B, and LS548C as either " normal" or " tripped". The bistables were recorded as being.in the " tripped" condition. Later, paragraph 5.4.13 required the bistables to be returned to the original position found in paragraph 5.4.6, in this case, " tripped". In lieu of returning the bistables to the " tripped" position or obtaining a pro- . cedure change to allow the bistables to be put in the " normal" I position, the maintenance technician placed the bistables in the " normal" position in violation of the procedure - then continued with the remainder of the procedure. Note: Site CAQR SQP890017 was issued to correct this deficiency. This deficiency is an example of non-willful failure to follow I maintenance procedures. It did not result in degradation of equipment. It occurred in early 1983 along with other similar problems which the licensee recognized as requiring generic corrective action. NRC Violation 327, 328/88-28-01 was later issued concerning a similar situation in May 1988. LER 2-88022 reported that situation. Widespread training was conducted _ _ _ _ -

7- o lN] I_k \\ s , ! 29 ' + under CAQR SQP 880349. Recent NRC reviews have not found this ' situation to be' common. This violation meets the criteria ~ specified in LSection V of; the NRC Enforcement- Policy ' for not issuing'a Notice of~ Violation and is not cited. ' This issue is identified 'as: LIV 50-327,328/88-50-03, example 2. .of 2, for tracking purposes and is considered to be closed. _ WR B237011, Steam Generator #2 Flow Transmitter 2-FT-003-155. - This WR 'was ~ issued ' to correct an indication problem.'under 3 no-flow conditions. - WR B267211, Turbine Driven AFW Pump 2A-S. This WR replaced the . Unit 2 turbine' driven AFW pump rotating element.' WR B234108, Feedwater Isolation Valve 2-FCV-003-87. This 'WR - corrected this valve's out-of-specification stroke time. -" WR B239447, AFW Pump 2B-8. This WR replaced the pump thrust ' bearing'due to a degraded condition being indicated by vibration analysis. - WR B267403, AFW Level Control Valve 2-LCV-003-164A. . This WR L corrected a problem with leak-through when the valve was closed. 'WR AS48867, Temperature Indicator on the Turbine Driven AFW Pump - Outboard' Bearing Housing. This WR replaced the subject temperature indicator due to the temperature readings being high when the. pump was not running. WR B104839, AFW Piping Welds AFD F25AA, AFD F25BB, and AFD F38. - This.WR was issued to correct unsatisfactory weld defects found by nondestructive testing of the welds. WR B261181, AFW Level Control Valve 2-LCV-003-172A. This WR ! - corrected a problem with leak-through when the valve was closed. WR B789967, AFW Pump 2B-B Inlet Pressure Indicator. This WR ) - recalibrates the indicator to correct inaccurate readings when 4 I the pump was not . running, i.e., the indicator indicated 29.5 psig vice 16 psig for the condensate storage tank head pressure. WR A298541, Steam Generator #3 AFW Flow Indicator 2-FI-003-147C. - This WR was issued to replace and calibrate the subject flow indicator. 1 j WR B267237, AFW Steam Trap Drain Valve 2-VLV-003-C18S. This WR l - lapped the valve seats. l- l l. i [ ! _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ - _ _ _ - _ - _ _ _ _ _ _ _ _ _ - _ - _ - _ - _ _ _ _ _ __ a

hW' _ i & %- K yo- , n.

. ~

O' ' . y ' 30 . h [ - WR' B218701. Thiss WRf was . issued to correct -aEleak at the- - - downstream' flange of the cavitating venturi in the' discharge ' U line from AFW Pump 2A-A. ' WR B117221, Turbine Driven AFW Pump 2A-S. This WR realigned the m - H subject pump. p 'WR B128783, AFW Check Valve. 2-VLV-003-873B. This ~ - WR - L. disassembled and. inspected the check 1 valve to sati sfy : the requirements of IE Notice 86-09. Note: .The vendor manual for this valve was. not avaliable on site. . WR B234138, AFW Level Control Valve 2-LCV-003-156A. This WR - corrected this valve's failure to meet its required stroke time and involved disassembly, cleaning, inspection and lubrication of the valve operator. WR B218703, Current Meter 2-EI-003-1198 for AFW Pump 2A-A. This - WR calibrated the meter. WR B219827, Motor Driven Suction Valve 2-FCV-003-116A from ERCW - to AFW Pump 2A-A. This WR disassembled the valve for internal pipe-weld inspection. Note: The vendor manual for this valve was not available on site. WR B267624, AFW Level Control Valve' 2-LCV-003-1738. This WR - replaced the valve positioner. WR B784804,- Containment Spray Pump IB-B. This WR replaced the - pump's rotating assembly due to a degrading condition identified by vibration analysis. Note: A portion of this work was observed by the inspectors. Two particular concerns were identified during a review of the pump coupling vendor manual, i.e., K0P-Flex Inc. Service Manual ' 1900-01 of June,1987, as follows: a. The coupling vendor manual, in the section entitled " Dynamically Balanced Couplings", states that each bolt, J nut and lockwasher must be maintained as a set to assure proper coupling balance. Additionally, this same section j requires that the coupling be assembled observing the i coupling match marks. These requirements were not included ! in site procedure MI-6.8, Containment Spray Pump

Inspection, Revision 6, the MI for disassembly and reassembly of the containment spray pump. Investigation 3 . _ _ _ __ ___ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

m . ,,( a.: ' . -31 determined that the requirements were covered in craft training,. however, it is believed that including these requirements ,in the' applicable site procedures would- enhance the procedures and. possibly prevent rework'in the ' future. , Note: Several .Sequoyah pumps utilize this type of '- coupling. b. TheLvendor coupling manual coupling assembly section, sheet 2, required torquing of the ' coupling fasteners and the lubrication plugs to specified torque values. These torque values had not been specified in MI-6.8, Rev. 6, as being applicable to the work. on WR B784804. Further investigation. determined that. no engineering evaluation of - ' these torque' values had been performed by the site. The lack of inclusion of these torque values in site procedures: and the lack- of an engineering evaluation ., ' justi fying the_ exclusion of these torque values is unresolved. pending engineering- evaluation -(URI 327,328/88-50-04). 'During a tour of the Auxiliary Building, the inspector. observed'that a limit. swit'ch was ' misaligned with the actuator arm and 'would not operate l as required on the containment spray pump room A cooler supply control- valve 1-FCV-67-184. -A .second limit switch 2-FCV-67-176 was also misaligned. The licensee provided the following information relative to this issue: The function of the limit switch was to indicate valve status by - operating lights on panels OM-27A. The consequences of the limit switch not activating would be l - ' that the valve position would have to be visually verified. - The valves are checked every 92 days by performing SI-166, which full -strokes certain valves during operation. Valves FCV-67-176 and FCV-67-184 are required to be stroked on a 92 day period. The SI would identify misalignment problems because the lights - would not operate properly. On December 14, 1988, WRs B283176 and B283172 were issued to - realign the limit switches. Numerous other limit switches were observed during various tours of the facility and no other discrepancies were observed. The inspector concluded that the misalignment probably occurred as a result of the - _ - _ _ _ _ _ _ _ _ _ _ _ 1

- - n g,... , < d .. . - 32 i switch being inadvertently bumped:during' work activities in the area.- ,'l The surveillance program appeared.to be-adequate and this. appeared to be an~ isolated occurrence. b. Predictive Analysis , .The predictive' analysis program was also reviewed during -this inspection. ' The" licensee. primarily. used three Ltypes' of- predictive . analysis to anticipate component failures. . Lubricating oil analysis and vibration analysis were_ used to predict failures in : rotating. equipment. M0 VATS testing was used to predict motor-operated valve ' failures. - The vibration ' analysis program has played a. very active role in predicting component failures at Sequoyah. This was evidenced by three of the- corrective maintenance WRs on .the AFW pumps and the- containment ' spray. pumo, reviewed at ~ random by the inspector, being - 1 the result of degrading pump performance.being found during vibration. analysis. ' Vibration analysis was' an integral part of the site's ASME Code Section XI pump testing program and was also an integral element of.the PMT~following any maintenance which would. change-the vibration . signature -of the equipment. The program was started in 1979 and computerized data trending became fully developed by-about 1983. I MOVATS motor operated valve testing also played a very active part-in the' Sequoyah maintenance program. The program:was started about 3 1/2 years ago -and has expanded' to coverEnearly 375 safety-related valves. Site procedures required MOVATS testing any time maintenance

affected MOV operability.

Additionally, MOVATS testing was accomplished by the PM program. A sampling of Unit 2 AFW system valve data determined that 13 valves were in the program:and most of these valves had, in fact, 'been tested twice. ' Seventy' five valves- ,4 per unit require mandatory- outage testing due. to NRC commitments made because the valve motor thermal overloads were by passed. < 1 At the time of the inspection, enh one M0 VATS engineer was on' site. Site management representatives indicated that they intended to provide additional personnel to accomplish the required testing during the impending Unit 2 outage. The lubricating oil analysis program, on the other hand, had not yet j : ' been developed to the extent that the vibration analysis and MOVATS i l programs had been. Though lubricating oil analysis was being accomplished as required by site PMs, the site had not yet accomplished the close integration of the vibration analysis program, the lubricating oil analysis program, and' the periodic oil change outs specified by various PMs. The PM program manager indicated , plans to take future actions to strengthen this area. j l 4 l ) _ - _ _ _ _ - _ _____ ____-_ __ - - _ .

, , _ _ _ . _ _ _ _ _ _ _ _ _ _ g m x

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L- .., ; . .

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3I; 33 , ' . c. - Preventive Maintenance During this inspection, ' the. licensee's PM program was reviewed to .- o _ determine the extent of PM actually being performed. This portion of - the inspection also focused primarily on AFW system components, with r additional emphasis on personnel airlocks. The inspection was . accomplished- by comparing vendor manual PM - requirements / , recommendations to - site PM instructions, reviewing adherence to . licensee ' and vendor-established PM intervals, reviewing the PM upgrade program, and discussion of the program with licensee , [ - personnel. The .SQN' PM ~ program was in a :significant period of change. In October, 1987,' the -site had developed approximately 2400 PMs. In January,1989, there were over 3900 PMs, i.e. , a jump of over 1500 - PMs in just 15. months. This significant increase was attributed' to the upgrade - program efforts by both TVA and contractor personnel . The upgrade program was being accomplished through a very detailed review of various data bases; e.g., NPRDS data, Sequoyah Failure- History,'IE Notices and Bulletins, vendor manuals, etc.; to determine PM.l requirements for generic plant equipment items - followed by . l . development of PM instructions to accomplish the requirements. The - licensee's effort in this area is noteworthy since, once completely L implemented, Sequoyah will have a very strong PM program. However, two areas of concern were noted: The ' by-weekly . listing of delinquent PM's, dated January 11, - 1989, (A delinquent PM is one which has exceeded its 25 percent grace period) listed 330 delinquent 'PMs - 146 of the 330 involved safety-related equipment. Of 55 required PMs on the Unit 2 AFW system, 41 had never been - accomplished on the equipment. Also, of the 55 PMs, about 30, or over half -of the PMs, had been developed in the last 15 months. If these statistics hold true for the other plant systems, then the PM workload indicated by the total number of delinquencies may be grossly understated since many of the new PMs have not been scheduled. Additionally, although the licensee has been very aggressive in developing PMs, the equipment in the plant has not been subjected to the PM, thus the benefit to plant equipment has not yet been realized. In December 1988 the licensee recognized the delinquent PM problem

- and initiated the following actions to reverse the previous year's increasing trend. These corrective actions have not been implemented long enough to' determine their effectiveness: A PM analyst staff of 3 engineers was established. - Maintenance discipline managers were made accountable for - performance of PMs by tying PM performance to the personal performance evaluation. . hm__au_m______-m_.__m_m.-m___________..___m_-___.___ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ - _ _ _ _ . _ . _ - _ _ . _ _ _ _ _ _ . ___. _ - _ _ _ _

. I r ... .. , , ..,

34-

. ' Electrical maintenance was provided an act' ion. plan'to reduce PM - delinquencies.

The PM administrative program was revised toistrengthen the

- control of.PM performance. y - A feedback' system was established to improve PM tasks and . - correct problems with' proper identification of PMs as outage or .non-outage. An. integrated -schedule we.s being developed - which -should - . minimize delay. . Efforts'were ongoing to level the PM workload throughout the - . year. .The PM work flow path had been improved to minimize - administrative delays. Specific PM activities reviewed also included those discussed below: PM' Instruction 1897, Upper Head Injection Level Switches. This - PM accomplished a set point check of the UHI valve closure switches. Note: .A portion of this PM was observed by the inspectors. .This PM ;was accomplished- under a TS one-hour LCO. Personnel . performing the PM were very knowledgeable and the PM was performed in accordance.with the procedural. instructions. Three problems did occur during this work and these problems were properly dispositioned by PR0s, as discussed below: PRO 2-89-005 reported, as required by.the PM instructions, - that switch 2-LS-87-21 was out of the range specified by TS. - PRO 2-89-006 reported that the one-hour LCO time limit for performance of this PM had been exceeded by six minutes. PRO 2-89-007 reported an actuation of valve 2-FCV-87-23, - l which was caused by valving-in switch 2-LS-87-21 too rapidly during the return of the switch to service. ' L The PMs for the personnel airlocks were reviewed. Vendors Manu- - l al SQN-VTM-C310-0030 " Personnel Airlock, Contract 92120 Fabric- l- ator, Chicago Bridge and Iron Company" contained maintenance l requirements in the following areas: 1 Shaft Seals - The shaft seals are teflon packing and - require no maintenance. i Q_--_-- __- . _ - _ - _ _ _ _ _ - - _ _ _ _ - _ _ _ _ - - _ _ _ _ - _ _ - _ - _ _ _ _ _ . __ ___- ____ . _ _ _ _ _ - - _ _ _ _ - . _ _-_ _ _

_- -__. __ i:p

,,

1 n ~ , , un ;w 35'- b , m + Bearings and Flange Block Lubrication - 1 . General - lubrication of gears,. sprockets', and chains and - bushings The -inspector reviewed the PM performance data for both Unit 1- and 2 reactor building airlock doors, as listed below: ,. Unit 1 Lower- Airlock: 2-19-87, 8-15-87, 2-07-88, 8-17-88 - _ - Unit 1. Upper Airlock: 2-16-87, 8-16-87, 2-08-88, 8-17-88: - Unit 2 Lower Ai_rlock: 2-15-87,,8-17-87, 2-01-88, 8-29-88- - Unit 2 Upper Airlock: 2-17-87, 8-17-87, 2-01-88, 8-29-88

-

The PMs' referenced above were performed as required and complied with the vendor manual. L. 9. Safety. Information Management System -/ Corporate Commitment Tracking System a. . Safety'Information Management-System During this . inspection period, the inspectors reviewed the accuracy of licensee input. to the SIMS - system and 'the licensee's use of the - _ CCTS system to. manage commitments made'to the NRC, SIMS is an NRC - - database system which provides NRC. management with a single source of reliable information- on NRC's management of generic and afety issues. It .contains, in part, the plant-specific s ' licensee-supplied data -that shows the status of - the licensee's implementation' of safety ' issues which are ' resolved by safety evaluations issued by the NRC staff. For Sequoyah, the SIMS database- contains data on the following issues: - Licensee employee concern element reports. L NRC Multiplant Actions - including the TMI action plan in - NUREG-0737. License amendments, which are primarily amendments to the Units - 1 and 2 TS. The remaining plant-specific issues. - SIMS was not a restart issue for the Unit 1 and 2 restarts from their recent extended outages. The last update of the SIMS for SQN on the licensee's implementation of safety issues was with data in the I licensee's letter dated February 17, 1988. The licensee will be providing the NRC another SIMS update by a letter scheduled to be , submitted by February 15, 1989. j . u - -. ._ _ _ . - - _ . - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ - _ _ _ _ _ _ -

- ___ , l tr; _e t !' 1 g.. l. 36 During this inspection ~ period, the inspectors reviewed the SIMS database update data planned to be provided by the licensee in the above letter. The inspectors reviewed files relating to the employee concern element reports for Sequoyah, the amendments to the TS, and the TMI -~ action plan for Sequoyah. The files relating to the employee concern ' element reports for Sequoyah were reviewed at the TVA offices in Chattanooga. The other files were reviewed at the Sequoyah site. The data to be submitted to the NRC in February, 1989, is to be compiled from the files that were reviewed. The inspector concluded from this review that the licensee is providing accurate input data for the SIMS database. b. Corporate Commitment Tracking System The CCTS is the licensee's system for tracking commitments to the NRC. This inspection reviewed the CCTS system for the following attributes: - The process to add commitments. Tracking of commitment completion status. - Late responses. - Verification of commitment completion. - Commitment closure documentation. - Closure document errors. - QA or QM surveillance and audits. - CCTS discussions were held at the corporate headquarters with individuals who control CCTS and at the Sequoyah site with site licensing coordinators and site CCTS coordinators who use the system. Procedures reviewed included: - ONP Standard ONP-STD-6.1.1, Rev. 0-C, Managing and Tracking NRC commitments, dated September 21, 1988. Procedure 0605.01, Commitment Management and Tracking, Rev. O, - dated January 13, 1987. - Sequoyah Standard Practice SQA-135, Commitment Management, Tracking and Closure, Rev. 8, dated December 22, 1988. i, _ _ _ _ - _ .

- -- - _ _ _ - . _ _ _ ___ - . _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ - - - _ _ _ _ - y . i lp 37 l. L Sequoyah Site Licensing Staff Section Instruction Letter - SLS-SIL-03, Rev.1, Status Management, Escalation, Trending 'and Distribution of NRC Commitments, dated January 20, 1989. Thel 'following ' CCTS closure packages, completed by TVAL during the period from July I to September 30, 1988, were reviewed: .NCO-85-0102-001 NCO-85-0522-001 NCO-85-0527-001- NCO-86-0156-159 NCO-86-0282-006 NCO-86-0471-007 NC0-87-0226-001 NCO-87-0232-001 NC0-87-0260-002 NC0-87-0356-005 NCO-88-0102-001 NCO-88-0119-005 NCO-88-0127-005 NCO-88-0136-001 ~ Commitments are made through letters, through LERs from the.licen'see to the NRC, and during licensee /NRC meetings or telephone conference:

calls.

The licensee had procedures to formalize these commitments. and input. them into CCTS. . The licensee's processito formalize a commitment to the NRC included developing an action plan with the - estimated cost and manpower to complete the commitment. Licensee personnel stated that, because the commitment action plan put the cost of completing the commitment into the budget, there was an incentive for 'the licensee to identify all commitments made to the ~ NRC. The inspectors concluded that the procedures were acceptable to ensure that commitments made by the licensee to the NRC would be tracked by CCTS. 'Sequoyah site licensing personnel explained the format of the CCTS computer printout and the different reports made at Sequoyah to track the status of the CCTS items, to assure that the commitments would be met on the agreed-upon schedule, and to identify late commitments. The inspectors concluded that the licensee has been acceptably tracking the status of CCTS items to assure timely completion. The selected completed CCTS commitment packages listed above were reviewed, and site licensing and CCTS coordination individuals were interviewed concerning CCTS commitment closure methods. Examples of

QA audits of CCTS closure packages were also reviewed. Ten percent ' of the CCTS closure packages were routinely sampled by QA, in addition to other closure packages specifically identified by Site Licensing for QA verification. QA trending did identi fy some problems with the commitment closure process. However, recently-developed TVA Section Instruction Letter SLS-SIL-03, Rev.1, l ' of January 20, 1989, provides more explicit guidance to appropriate staff personnel to further streamline the CCTS review and verification process. Continued training using these newly revised guidelines, plus the QA trending audits, should provide the licensee with continuing confidence that proper CCTS closure is occurring. I _ _ - _ - - _ _ _ _ _ . - _ . _ _ _ _ _ _ _ _ _ _ __ _Y

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. { ' ' 9 38: u 1.- Within the scope of this inspection, the inspectors concluded that the-licensee has been acceptably addressing commitments made .to the " NRC. 10. Plant Modification Process This portion of the inspection focused on modification workplans, observation of field work, and QM activities in the. modification area. a. Modification Workplans Discussions with Modification personnel revealed there were approximately 1300 workplans remaining . open, some dating back to 1980. All required physical work was completed on these workplans . prior to plant startup, however, the work plans were left open for the various reasons indicated below: 255 Workplans were status WI "Workplan Implementation". The - workplan was being implemented, and the cognizant engineer had .the workplan. 106 Workplans were status HPC " Hold - Partial Complete". Work -- had been started but not completed. -The document coordinator had the workplan in the hold file. ~ 15 Workplans were status HM " Hold for Material". The document - coordinator had the workplan in the hold file. 14 Workplans were status HPM - " Hold for Manpower". The. - document coordinator had the workplan in the. hold file. . 390 Workplans were status- FC "Workplan in Field - Complete". - The cognizant engineer had-39 days to complete the documentation and turn in the workplan. - 489 Workplans were status DC "Workplan Sent to Drawing Control Center for Drawing Update". - 3 Workplans were status DCP "Workplan Sent to Drawing Control for a Partial Update". - 33 Workplans were status ANI " Authorized Nuclear Inspector". The ANII had the completed workplan for review and sign-off. "Workplan Documentation 206 Workplans were status WDR - - Resolution". The completed workplan had been returned to the cognizant engineer for documentation corrections. The inspector selected six workplans for review: - Workplan 7346-01 - Issued in April 1988 . _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ - _ _ _ _ _ - _ - _ - _ - _ _ - _ _ _ - _ _ _ _ _ - - _ _ _ - _ _ _ _ _ - _ _ _ - - - _ - _ _ _ _

. - - - _ _ _ ., x: . j. l. 1 39 Workplans 9557, 9560, 8652 - Issued in 1981 - Workplan 11788 - Issued in 1985 - - Workplan 12618 - Issued in August 1987 l - Workplans 9557 and 9560 required the changing of carbon. steel lines to stainless steel on the ERCW 2-inch diameter and smaller lines. They were issued in 1981 and some work was accomplished through 1984. Other piping change-outs described on these workplans had never been scheduled for work. The inspector's review did not identify any safety issues caused by these workplans being left open. However, some items that required licensee action were incomplete. For example, of 28 weld data sheets reviewed, 26 were not confirmed by QC, as required, to assure that the welders were properly qualified. , l 1 Workplan 7346-01 was issued to replace existing normal feeder j - cables-for the turbine driven AFW pump and to replace cables for - the turbine driven AFW pump room DC vent motor and starter. The field work was completed on September 27, 1988. At that time, the cognizant engineer sent the secondar.y drawings to drawing control for revision. The workplan will remain open until the drawing revisions are completed. Procedures SQEP-30 and SQEP-42 required that secondary drawings be updated (revised) within 90 days. The drawing ' revisions should have been completed by December 27, 1988. On January 24, 1989, the completion notice had not been returned to modifications. The licensee advised that CAQR No. SQP880126, Rev. 1, issued February 16, 1988, identified a problem in the drawing update program. The corrective action, issued January 4,1989, stated that: , 1) Primary Drawings: The Sequoyah Engineering Project will develop an overall revision plan to encompass resolution of all [the] primary drawing backlog in order to prevent recurrence of redlines on the primary drawings. 2) The Sequoyah Engineering Project shall meet the 90 day timeframe for secondary drawing revisions as designated in site procedures SQEP-30 and SQEP-42 for all work received on or later than December 1, 1988. Scheduled completion date: March 1, 1989. 3) The Sequoyah Engineering Project will develop an overall secondary drawing revision backlog plan (work received prior to December 1, 1988). Currently this backlog is approximately 5,100 revisions. This plan shall be established based on budget and schedule for ECN/DCN I l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ l

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A ' a. 40 h . closure. A scheduled completion date cannot be determined at> this ' time due to pending budget / sche'dule' approvals . by 'the. senior vice president of nuclear power in April 1989. A schedule. shall be prepared within 60 days of the'.. ' established budget,- noted above, .to encompass all of the . secondary drawing backlog. , f. Based on the above, it could not be determined when this workplan will.be completed and closed. Some line items remained open. ' The engineer advised that they would be addressed when < - the workplan was reviewed for closure. The inspector's review of the workplan did not identify any safety' issues.- Workplan 12618 was issued August 14, 1987 to ' replace the. - .non qualified discharge pressure gauges for AFW pump A-A and B-B with qualified seismic -1(L) pressure gauges. The work was completed and the workplan closed on October 14', 1988. The 1nspector's-workplan ~ review included seismic qualification report SQEP-C2-L7156. " Seismic Qualification of Auxiliary Feedwater Pressure Gauges." This document justified the seismic adequacy'of the new gauges. All items reviewed on this workplan were found acceptable. , , -- Workplan 8652 was issued May 22, 1980 to replace the ice condenser air handling unit three-way glycol valves with two position solenoid valves. This work was completed shortly -after the work plan was issued. However, the workplan required a .PMT be performed per procedure MTI-11 " Ice Condenser Air Handling Unit Performance". This PMT was . never. performed. Further,' MTI-11 has since become an obsolete procedure. The workplan remains open with no planned completion date. Workplan 11788 was issued to move two pressure transmitters and - one level transmitter from inside the crane wall to inside of the incore instrumentation room. The pressure transmitters were identified as 1-PT-68-322 and 1-PT-68-323. The level transmitter was identified as 1-LT-68-320. Approximately 91 welds were completed on this worKplan in 1985. In each case, the workp'lan required a QC post work review to verify that the welders were properly qualified to perform the applicable weld procedure. This verification review had yet to be performed in I " 1989. After QC completes the welder qualification verification, the cognizant engineer is to review and sign the weld data sheets. This had not been done. -_-___=___-_:____- . _ - - _ _ _ . _ _. _ . - _ _ - _ _ _ _ _ _ _ _. _ _ _ _ _ _ !

_- - _- _ _ - - - _ _-_ - - w .; -.: 41' , - As : indicated in ' th'e workplan on August 29, 1985, the ANII ' " required a review of the complete package. This - had not been . done. - On September 21, 1988, a Partial Modification Completion Form was' completed for the subject' workplan. It identified the work- remainingL to be a leak check of new piping 'and tubing and

calibration of the instruments.

.The partial modification ! package was approved on September 25, 1988. The inspector met with the onsite ANII to discuss the requirements for completing the reviews he had specified on the

workpl ar.s .

1 AS!!E Code Section XI contains the ANII's authority and - -jurisdictional boundaries. The code does not specify that the ANII review the workplan prior to placing the unit in -service. ASME Code Section XI does specify the following: - Paragraph IWA 6250 requires that records and reports of repairs and replacements shall -be prepared . . _ and filed with the . . . regulatory authorities having- jurisdiction at the plant site. Paragraph IWA-6220 requires that the inservice inspection reports shall be filed with the regulatory authorities ... within 90 days after completion of the inservice' inspection. Discussions with -the ANII and mechanical test engineers show a large . number of outstanding workplans which could involve ASME Code Section _XI, Class 1 or 2 activities, and require ANII review before submittal to the NRC. These reviews had to occur within 90 days after completion of the inservice inspection (end. of the refueling outage) so that the report could be submitted as required. The licensee advised that a nonconforming condition, discussed below, was being issued in this area. This concern was identified by the licensee. CAQR SQP880607 was prepared January 13, 1989 for Unit 1 and was in the review cycle prior to issuance at the end of the inspection. It identifies the following deficiencies regarding compliance with ASME Section XI: - All summary reports of repairs and replacements to ASME Code Class 1 & 2 pressure-retaining components and their supports and pressure retaining parts of components have not been submitted within 30 days after Unit i refueling i outage. Note: The 30 day requirement is an internal l requirement so that the 90 day reporting requirements are met. 1 i " _ L l l 1

- L c ,- t . 42 i - Several ASME work requests / maintenance requests have been - microfilmed as completed QA records without obtaining ANII final review of the package. The CAQR identifies that approximately 49 percent of the ' ASME k Code Section X7 work packages have not been reviewed or approved by the ANII or QA, as requided for package closure. The CAQR indicates that February 5,1989, is the required date for the Unit I refueling outage summary reports to ba submitted to the. NRC, but only about 51 percent of the reports will be completed. The remaining 49 percent of the reports will be submitted 90 days later on May 6, 1989. The CAQR apparent cause analysis identified that, due to the large number of workplans and WRs that were worked to support Unit 1 and Unit 2 restarts and the priorities of field wo'rk completion that were set, not all workplans and WRs were administratively and procedurally closed out. Therefore, the reviews, corrections, approvals and summary reports were not accomplished as the field work was completed - which caused the delays and an extensive backlog. Additionally, a memorandum dated December 19, 1988 (RIMS S08 881219 805) identified the following conditions relative to this issue: $ - No tracking system existed for ASME Code Section XI WPs,

WRs, or MRs that required ANIl review and involvement. Lack of defined responsibilities in the preparation of - documentation for repairs and replacement. - No compiled list of ASME Code Section XI work during the

Unit 1 Cycle 3 (April 1984 - November 10,1988) timeframe. Different [various] implementing procedures. I - Though this inspector's review has not identified any issues which would directly affect safety, the inspector did determine, based on this review and extensive discussions with engineers regarding the large backlog of uncompleted workplans and uncompleted ASME Code Section XI reviews, that the fcilowing management control oversights exist concerning workplans, WRs, " and MRs: i The large backlog of workplans exists primarily due to the -l - large volume of modifications at the site. However, better 6 l administrative controls should have been in place to systematically complete the verifications and paperwork after the field work was completed. For example, welds that were completed in 1985 should have received a i ! _ - _ - _ _ _ _ - . _ _ _ _ _ _ _ _ __

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e, . 43 confirmation that the' welders were qualified shortly'after completion of the welding. Completion of the-items at thes time they occurred would have prevented some of - the backlog. Failure to implement status contro.ls for inservice -- inspection of repairs and replacement, along with the failure to complete workplan processing thru the ANII and QA review cycle, could result in a failure to comply with 10 CFR 50.55(a.)- requirements if prompt action. were - not taken. This is identified as URI 327,328/88-50-05 " Completion of Workplan Review and Reporting of. Section XI- Repair and Replacement" pending the licensee's resolution of these issues. b .- Observation of Field Work The inspector conducted several tours of the Auxiliary Building and ' reviewed work being performed by the licensee's modification group. The. modification activity reviewed involved welding of- the -16-inch diameter stainless steel- piping serving two new - CCW HXs. The ~ . licensee:was replacing one' existing CCW HX with .two smaller units. The ' welding activity' in progress at the ; time of the inspection was the fitup-and tack welding of the 16-inch diameter, 0.375. inch wall- thickness, stainless' steel pipe. Tack welding was physica'lly . in progress. The inspector interviewed the welder and welder foreman.regarding the welding. paramaters required- for welding. stainless steel . The welder advised that'the following parameters-were applicable for the weld in progress: preheat was +60 F,. interpass temperature was controlled at 500 F maximum, weld rod was 1/8" diameter, and weld material was type E316. The welder also advised that -the welding procedure was GT ' 88-0-3. Review of the welding parameters established in the qualification of GT 88-0-3 showed that the maximum interpass temperature was established at' 350 F and not at 500 F as stated by the welder. All of-the other stated parameters were correct. -Interviews of the welder, welder foreman, modification engineer, QC inspector, and QA surveillance inspector regarding interpass temperature controls established the following: The welder stated he did not use Tempilstiks to check interpass - L temperatures. The digital thermometer found at the weld station had a range of - 0-300 F and was only effective for measuring preheat temperature. ) ! ~ . - ' _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

__ e m r 'l . 44 . ' QC inspectors did not verify interpass temperatures. - - QM monitors did not perform interpass temperature checks, i ' Early QM monitoring records did not indicate that stainless - steel interpass temperatures were verified in the past. Recent reports for monitoring performed after the issue was identified showed interpass -temperature controls to be in effect and so noted in the report. The modification group contained a weld engineering section - which had been performing in process monitoring of welding using a draft procedure. The inspector reviewed numerous reports titled " Welding - Activity Verification Report" which had been conducted by the weld enginee' ring section in accordance with draft procedure DNC-GCP 8.1.4-02 " Verification of Procedural and Specification Requirements for DNC Welding Activities." At the time of the inspector's review, this document had not received final approval and was not issued. Prior to the exit meeting, the licensee advised that the procedure would not be issued for 2-3 months and, during the interim period, an SQA procedure would be issued - but the SQA procedure planned issue date was also unknown at the time. A review of past reports found that the weld. engineers had not specifically indicated on the report that interpass temperatures had been checked. However, numerous reports issued after the concern was identified by the inspector specifically addressed interpass temperature controls. The licenses immediately implemented a training program on January 11-12, 1989 for all boilermakers, steam fitters, and foremen involved with welding activities on site. The training was titled " Welding - Interpass Temperature". A total of 51 employees received the training. Additionally, the licensee was conducting further training of welders and foreman for understanding welding symbols and the details of welding procedures. This training was scheduled for completion in Jan ua ry , 1989. On January 11, 1989, the QC Supervisor directed that effective January 17, 1989, the Welding QC Section would implement an informal weld surveillance program. The welding surveillance program included, among other things, an ettribute for verifying interpass temperature controls during welding.

The inspectors concern regarding the need to control interpass temperatures was that elevated temperature tends to increase the sensitization of austenitic stainless steel, and if combined with causative conditions, intergranular stress corrosion cracking might occur. The licensee provided the inspector with a report from the - - _ _ _ _ _ _ . - _ _ . _ - .

- -__- -- _ < , 7.s ..- ,f . .' 45' ' ' licensee's metallurgical engineering ~ group that had conducted metallurgical analysis of _ weld samples that were welded to the parameters of welding procedure GT 88-0-3 except that- the interpass temperatures were intentionally raised- to greater' than 800 F. The tests .had been performed -to' address an NRC Notice of Violation at Watts Bar involving similar issues. The welds had been subjected to a ' corrosion test in accordance with ASTM A262, Practice A. The samples were then sectioned, polished, and etched using the ASTM A262' . procedure. The - results for butt, welds were reported ' as followsi "Neither weldment shows complete sensitization as evidenced by grains " completely surrounded by ditches. ;Both specimens show partial ditching of 'the grain boundary, which 'is an acceptable-microstructure under the conditions of A262, Practice A." The inspector's review of these tests concluded that some sensitization- occurred at increased elevated temperatures, which was undesirable, however, failure would not. occur as a result of the increased temperatures. The inspector concluded, based on the interviews and . records reviewed,'that although~interpass temperature controls were not being assured in the past, adequate controls were being implemented by the ~ end of the inspection. Further, the ASTM A262 test determined that -the welds would not' fail when ' welded at elevated interpass temperatures. Failure to follow weld procedure GT 88-0-3 for control of interpass temperature is a severity level V violation. This review has found that it was not willful, not similar to prior violations for which . corrective actions have not been sufficient to prevent recurrence, and that the licensee has taken extensive corrective action prior to -the end of the inspection. This . violation meets the criteria specified in Section V of'the NRC Enforcement Policy for not issuing a Notice of Violation and is not cited. This issue will be. tracked as LIV 327, 328/88-50-06 and is considered to be closed. The inspector audited the activities associated with a welder perforaing qualification test welding in accordance with ASME Code Section IX. The weld procedure and procedure qualification test requirements were at the test station. The weld engineering section personnel administering the test were very knowledgeable of the requirements for qualifying welders. Additionally the inspector witnessed the sample preparation and the performance of the required side bend test. .The areas reviewed by the inspector were found acceptable. < I c. Quality Monitoring Activities - Modifications The inspector reviewed the licensee's QM program in the area of modifications, including interviews with the quality manager and l quality supervisor. Additionally, the inspector accompanied a QM ' monitor on a field monitoring activity of in process welder i - - _ _ _ - _ - - - - _ - - - - _ - _ _ - -

'. - 4 ~ 46 qualification. The QM monitor was knowledgeable of the area of welding and ' welder qualification requirements. The items specified on the checklist were followed and appropriately marked. Areas not applicable were also marked on the checklist as required. The inspector also reviewed 10 completed QM reports on mechanical activities. All areas reviewed were found acceptable. The review found the QM' program to be a very effective means for the licensee to self monitor and implement corrective actions, when needed, on areas previewed as potential problem areas. Additionally, it was found to be a very effective tool for the licensee to verify that acceptable areas' continue that way. The monitoring activities reviewed were well planned, and were being adequately implemented. They appeared.to be an excellent management tool. d. RHR Sump Valve Room The inspector reviewed selected listed items relative to the licensee's basis for determining that an adequate margin of safety exists in the RHR sump valve rooms as presently constructed. The rooms were originally intended to meet containment building post-LOCA conditions. As presently constructed, the rooms, adjacent to the containment, are not fully able to withstand containment post-LOCA conditions because electrical and mechanical penetrations into the room from the auxiliary building are not leaktight. The licensee had changed the FSAR in 1988. The licensee reviewed the issue and established the position that catastrophic sump line or sump valve failure, which could result in large leakage into this room, were not credible based on the following: - Remote possibility of line failure Weld integrity adequacy - - Valve body failure probability being very remote - Postulated leakage from minor stem packing leakage being < determined to be virtually no leakage 1 The following documents relative to this issue were reviewed: l 1 - Ultrasonic examination reports for welds RHR-1-5 R1, RHR-1-6. ! (Weld RHR-1-6 was rejected for lack of fusion, repaired and reinspected, and found to be acceptable.) - Detail Weld Procedure GT18-0-1, Rev. 6 I - Weld History Records for Welds RHR-2-4, RHR-1-3, RHR-1-4, i RHR-2-3 l l

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ii 47 , _' Process Specification '3.M.7.1(2) " Specification For. Ultrasonic - , h Examination of Weld Joints" All areas reviewed by the-inspector were found acceptable. .' 11. QAlRcutineAudits/QCActivities/andSpecialSurveillances(orMonitoring Activities) in Support of Operations a. ' Quality Assurance Audit Organization The- NRC inspectors reviewed the following site . Quality Assurance audits which were either complete orlin progress: SQA.87-0003,. Mechanical Maintenance SQA 87-0015, Electrical Maintenance SQA:87-0018, Plant Modifications and Design Control SQA 88-804, Independent Qualified Review Process SQA 88-808, Compliance'with Technical Specifications SQA 88-811, Instrument Maintenance SQA 88-815, Ongoing audit on various programmatic areas associated with the CVCS system. CMAP Audit, performed in June 1988, covered the Quality Audit and Monitoring Programs The NRC-inspector noted that the Quality Audit reports appeared-to be thorough, containing an adequate level of detail'to stand alone as QA records, and clearly stated any findings - whether recommendations, concerns, or conditions ~ adverse to quality. The inspector. selected the names of two of the licensee's lead auditors for personnel qualification review. They were lead auditors

on two of five audits then in progress at Sequoyah~ Their audit . areas were: ) Inspection Audit of Quality Control - Experience Review Audit - This review compared the individual's certification to the 'following documents: QM1 317 Rev. 2 dated June 14, 1988 " Auditor Training and - Certification For NQA & EB Personnel" s - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ . _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ . _ _ . - _ _ _ j

- _ _ , . . 48 QMI . 329 Rev. O, dated June 12, 1987 (Addendum 1, dated - January 19,'1988; Addendum. 2, dated February 12, 1988) titled- " Periodic Evaluation of Auditor Performance." The review of the documents completed in accordance with QMI 317, Attachments 1 and 2, determined that both individuals met the , required elements in the areas of education, experience, professional accomplishment, and prior management experience. They had attended the- required training courses and had participated in an adequate number of previous audits. The records reviewed indicated they were well qualified as QA lead auditors. The CMAP audit discussed below also concluded that auditors were properly trained and certified. In a CAQR that identified five deficiencies associated with the NDE Level III qualification for one of the lead auditors, the deficiencies were administrative in nat'ure and did not affect the actual qualification of the lead auditor. During June .13 - 24,1988, a CMAP audit was conducted of the DNQA. This special audit of Segouyah quality-assurance-related activities was conducted by a team of experienced personnel from three outside nuclear utilities to satisfy commitments made in TVA Topical Report TVA-TR75-1A, Sections 17.1.2.5 and 17.2.2.5. The CMAP audit team concluded that both the quality audit and quality monitoring programs were effective, that both programs were conducted to the appropriate depth to identify program and hardware problems in a timely manner and that problems were clearly and accurately described on CAQRs. This audit resulted in several recommendations associated with more efficient utilization of personnel and planning / scheduling of quality audit and monitoring activities. The NRC inspector noted that these recommendations had been adequately addressed by the licensee with some minor program changes. Two CAQRs were identified during the CMAP audit: , CAQR CHS 880047 identified five deficiencies associated with the - NDE Level III qualification for one of the lead auditors. The deficiencies were administrative in nature and did not affect the actual qualification of the lead auditor. l l - CAQR SQA 880432 identified the failure to adhere to procedures l ' to adequately document the COTS and provide the required , l information for trending. b. Quality Monitoring Organization The licensee's onsite Quality Monitoring Organization was recently i created to replace the older Quality Surveillance Organization. ' 1 . 1 _ _ _ _ _ _ - _ _ _ - _ _ _ _ _

r -- _ _ _ _ _ _ r.: 4- , - p 49 k Quality Surveillance and Monitoring activities were verification R techniques intended by the' licensee to assist the site management in- L meeting quality ' objectives by providing continuing evaluation of performance and identifying conditions adverse to quality before they impact nuclear. safety . reliability or component' operability. Unlike . Quality Audit activities which .are generally. programmatic .in. nature, y are generally ; part of a structured program designed to satisfy l regulatory requirements, and usually consist.primarily of the review of documents or records, the QM group featured inspections on a real m . time basis - including observation of activities'in progress. The Quality Monitoring Program was designed to provide for a " quick. u l look" by an experienced QE within each of the disipline areas. This - was accomplished by a group of 12.QEs and 3 group supervisors. Each QE' was expected to complete an average of 8 quality monitoring reports ~per month for a group total of approximately 1,000 monitoring- reports per year. The group was still able to perform surveillance on . a - case-by-case basi s , however, these were, more detailed and required a' greater. amount of. time than monitoring reports. L An NRC inspector reviewed the. organizational structure of the QM ' - Group along with selected resumes for QEs and group supervisors. Interviews'were conducted with several QEs and:two group supervisors. Additionally, the licensee provided a matrix which outlined each individual's background. Although the group was a relatively new organization, it contained a well qualified blend of personnel with a good mix of education, training, and work related experience within 3 -the various technical. discipline areas. i One of the recommendations of the CMAP audit discussed above was for improvement of the monitoring schedule. It was viewed by the CMAp team as too structured and not - allowing sufficient time for independent inspection effort by the QEs. In response, the licensee modified their program to allow approximately 50% of a QE's time to remain free which provided sufficient flexibility to perform special and immediate monitoring while assuring specific areas were being monitored. To accomplish this, QEs and group supervisors were required to attend routine plant meetings such as the operations nift turnover briefings and daily plant status meetings, and to make frequent plant tours to maintain cognizance of ongoing plant status. They were then expected to schedule meaningful "real time" activities for monitoring. NRC inspectors accompanied QEs during portions of selected ongoing monitoring activities. In each case the individual attributes that were scheduled to be checked were verified and denoted as such on the checklist used by the QE. The NRC inspector then reviewed several of the completed monitoring reports and discussed the results with the respective QE. The results section of each QM report was found to be complete, including subjective comments concerning the area - _ _ _ - - _ - _ - _ _ _ _ _ - _ . ___ __. _- -- -_- - _ _ _ - _ - _ - _ _ _ - _ _ _ __ -____- _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - - _ - _ - _ _ -

--_ - _ - _ c.3 . l m l w 50 , monitored. The QE also orally communicated the results to a licensee management representative for the respective discipline. The NRC inspector determined that the QM Group was an experienced and motivated group of capable individuals that should contribute.to the management effectiveness at SQN. 12. Independent. Qualified Reviewer Process The IQR process was a new technical review and control process implemented prior to-Unit 2 startup that supported the review and approval of changes to procedures and changes or modifications to safety-related structures, systems', or components, as -allowed by TS 6.5.1A. The intent of the new process was to reduce the administrative burden on the PORC and allow more available time for PORC _ review of significant issues essential to the - ' operation of the plant,'thereby improving PORC effectiveness. IQRs were required to perform a detailed technical review of procedures, changes to procedures and' safety evaluations for the following: - Procedures required'by TS 6.8.1 0ther procedures which affect nuclear safety - P1' ant modifications to safety-related structures, systems, or - components These ' reviews must be performed by reviewers trained within their respective area of responsibility or expertise. Reviews were to be performed in accordance. with AI-43, Independent Qualified Review, by personnel- designated on Appendix A of AI-43. The reviewer could not be the preparer of the procedure change or the plant modification. Reviews must include a USQD screening review, i.e., a determination of whether or not an unreviewed safety -question per 10 CFR 50.59 was involved. Screening reviews would be performed in accordance with SQEP-128, 10 CFR 50.59 Qualified Safety Evaluations. Additionally, each review must include a determination of whether or not a cross-disciplinary review is necessary. The NRC inspector reviewed completed Quality Audit Report SQA 88-804, February 18, 1988, concerning SQN which was performed January 19 - programs for control of the USQC and procedure change processes. The audit concluded that both programs, including the IQR process, were effective in achieving the desired results. Although, no CAQRs directly relating to the IQR process were identified, several concerns and recommendations related to the other processes were identified. The licensee had generally addressed each of these concerns by redsion of an associated procedure or' otherwi se implementing the recommended improvement. Subsequent NRC Notice of Violation 327, 328/88-43-01 (September 1988) addressed the 10 CFR 50.59 review process, both from the TVA generic and SQN-specific aspects. i i a________________._________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ __ _ _ _ _ _ _ _ _ . _ _j

- _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ , e .' , 51 Various administrative, surveillance and operating procedures were selected for review by the NRC inspector to verify the adequacy of the licensee's determination of cross-disciphnary-review necessity. In all cases where the NRC inspector believed that a review by another section might be required, the associated procedure routing page reflected that a cross-disciplinary review had been performed by the appropriate section. In Audit Report SQA 88-804, a concern was identified about the failure of TS 6.5.1A to require TACFs to have an IQR review. In reply to the audit concern, the licensee stated that all TACFs on CSSC components receive a safety evaluation, screening review for USQ, and Plant Manager /PORC review and approval prior to implementation. Since PORC review and approval in this manner satisfies the intent of TS 6.5.1A, the licensee has adequately addressed the audit concern. Per AI-43, to qualify as reviewers, individuals must meet the following experience requirements: - Possess an academic degree in engineering or related field and have 3 years nuclear related experience or Hold an active Reactor Operator (RO) or Senior Reactor Operator - (SRO) License Individuals who do not possess the formal education or license specified above are evaluated for consideration on a case-by-case basis. Additionally, each reviewer must undergo four hours of formal training covering the duties of Independent Qualified Reviewers and eight hours of formal training covering USQD evaluations. The NRC inspector selected for review eleven Independent Qualified Reviewers from the approved list. The qualifications for those reviewers were then checked to determine the quality of the selection process. The licensee documents the management review of the individual's education, experience and training on Attachment 3 to AI-43. For those Attachment 3s reviewed, the level and type of experience appeared to be adequate and consistent with the area of discipline that the reviewer was assigned. Additionally, any personnel that required special case-by-case approval due to lack of a degree or license had a reasonable level of other experience to justify their use as a reviewer. The licensee's IQR process, as implemented, appeared to provide indepen- dent technical and screening reviews for those procedures, changes or plant modifications specified above. Additionally, this program also reduces the administrative burden on the PORC. 13. Exit Interview (30703) l The inspection scope and findings were summarized on January 26, 1989, with those persons indicated in paragraph 1. During the inspection period, frequent discussions were held with the Site Director, Plant Manager and other managers concerning inspection findings. The inspectors i ' 1 i -- _ _ _ . _ _ _ _ _ _ _ _ _ _

-- 3 n eii (b ' ) 52 1 , described thel 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 l- inspectors during the inspection. . Item Number Description and Reference [ 327, 328/88-50-01- Violation - Failure to_Take' Adequate Corrective-Action'en Previous Violation 327, ' 328/87-30-01, paragraph 5. 327, 328/88-50-02 URI - Trending Within ACPs and the Appropriate Thresholds for Entering the CAQR , Process, paragraph 6. 327, 328/88-50-03 ' LIV - Uncited level V Violation for fail'ure to Follow Maintenance Procedures, Two- Examples, paragraph 7.

327, 328/88-50 04

URI'- Inclusion of Vendor Manual Torque ' Requirements in Maintenance Instructions, paragraph 7. ~ 327, 328/88-50-05 URI - Completion of Workplan Review and Reporting of ASME Code Section XI Repair and Replacement, paragraph 10. 327, 328/88-50-06 LIV - Uncited Level V Violation for , Failure to Follow Welding Procedures, paragraph 10. 327, 328/88-50-07 URI - Engineering Evaluations of Vendur Manuals, paragraph 5. ' 14. List ~of Abbreviations ABGTS- Auxiliary Building Gas Treatment System ABSCE- Auxiliary Building Secondary Containment Enclosure ACP - Administrative Control Program Auxiliary Feedwater (system) AFW - Administrative Instruction AI - ANII - Authorized Nuclear Inservice Inspector Abnormal Operating Instruction ' A01 - ASME Code-American Society of Mechanical Engineers Boiler and Pressure Vessel Code Auxiliary Unit Operator AVO - ASOS - Assistant Shift Operating Supervisor BFN - Browns Ferry Nuclear Plant BIT' - Boron Injection Tank Control and Auxiliary Buildings C&A - ___ _ __-___--_- _

_ - _ _ _ _ - _ _ _ . \\ p3 a $ . 53 c. l[ CAQR - Condition Adverse to Quality. Report

CCP - ' Centrifugal Charging Pump , b. CCTS - Corporate Coinmitment Tracking System CCW HX- Component Cooling _ Water Heat Exchanger CMAP -- Cooperative Management Audit Program COPS - Cold Overpressure Protection System COTS - ' Correct on the Spot 'CSSC . Critical Structures', Systems and Components CVI - Containment Ventilation Isolation DC . -' Design Change DCN - Design Change Notice- DCR ' Design Change. Request DNE - Division. of Nuclear Engineering DNQA - ' Division of Nuclear Quality Assurance DWL ' - . Daily Work List- EA . - ~ Engineering Assurance .ECCS -

Emergency Core Cooling System-

ECP - Employee Concern ~ Program -Emergency Diesel Generator ~EDG - Emergency Instructions- EI - ENS -: Emergency Notification System EQIS - Equipment Qualification Information System ERCW - Essentia1' Raw Cooling Water Engineered Safety Feature ESF. - FCV - . Flow Control Valve Flow Indicator FI - FSAR - Final Safety Analysis Report GDC - General Design Criteria Generic Letter GL '- Hydraulic Control Valve HCV - Hand-operated Indicating Controller HIC - Hold Order H0 ' - HP '- Health Physics i to p- Current to. Pneumatic (converter) ICF - Instruction Change Form IFI - Inspector. Followup Item IM - Instrument Maintenance Instrument Maintenance Instruction IMI - NRC Information Notice IN - INPO - Institute of Nuclear Power Operations Independent Qualified Reviewer IQR - IR - Inspection Report ' ISEG - Independent Safety Engineering Group Kilovolt-Amp KVA - Kilowatt j KW - Kilovolt ] KV - LE - Level-Element Licensee Event Report LER -

Limiting Condition for Operation LC0 - LOCA - Loss of Coolant Accident i j . . - . _ . - _ - - _ - - _ _ _ _ - _ _ _ _ . _ - - _ _ _ _ _ _ - . _ _ _ _ _ _ .

} , A .- 1 m 54 LT Level Transmitter - ' Maintenance Instruction MI - -M&TE - Measuring:and Test Equipment.- ~MOVATS- . Motor-Operated Valve Actuator' Test System . Main Control Room MCR - - MRC - Management Review Co'mmittee MTI - Mechanical Test Instruction -Nitrogen Gas .N2 - Not Applicable .N/A: - ' NDE - -Non-Destructive Examination 'NOV Notice of Violation - > NPRDS - Nuclear. Plant _ Reliability Data System NQA - - Nuclear. Quality' Assurance NQAM'- Nuclear Quality Assurance Manual NRC. - Nuclear Regulatory. Commission -ONP - Office of Nuclear Power.(TVA) '0SLA-- -Operations.Section Letter Administrative OSLT - Operations Section Letter..- Training , i OSP_ - Office _of Special Projects' PDWL - Priority Daily Work List Problem Identification Report . _ PIR - ,PM ' Preventive' Maintenance (action, instruction, etc.) - PMT. - Post Modification Test PNL. - Panel- POD '_- Plan of the Day PORC - Plant Operations Review. Committee .PORS - Plant Operation Review Staff POTC - Power. Operations Training Center -Problem Reporting Document PRD - Potentially Reportable Occurrence PRO - Pressure Transmitter PT -- Quality Assurance 'QA - -Quality. Control QC - Quality Evaluator .QE - QM - Quality Monitor (ing) Quality Verification QV- - .QVFI - -Quality Verification Function Inspection Reactor Coolant System RCS - , Rem - Roentgen-Equivalent-Man ' Rev. - Revision Regulatory Guide RG . ? RM - Radiation Monitor l' RHR - Residual Heat Removal Resistance Temperature Detector -RTD - Radiation Work Permit RWP - ^ RWST - Reactor Water Storage Tank SCR - Significant Condition Report System Evaluator. SE - Safety Evaluation Report SER - Steam Generator .SG - SI - Surveillance Instruction i .. - - I

.- _ - -- Op ~ [. '. f . t ,.. $ 6 I 5 E-- o - T 55 ' , - ;. ESIMS - Safety,Information Management System

. , 1

SIS

1 Safety Injection-System - - System'0perating Instructions )

SOI

- -SOS: = . Shift Operating Supervisor I SQM -- Sequoyah Standard Practice Maintenance SQN'

Sequoyah Nuclear Plant.- -Surveillance Requiremen;s SR~ - - SRO' - Senior Reactor Operator 'SRST - Spent. Resin Storage Tank: .Special Test Instruction STI -- Shift Turnover Restart Meetingn(report) ' STORM- 1 ~ TACF - ' Temporary Alteration Control Form . -TMI - Three Mile. Island. Nuclear = Plant y TROI - . Tracking .of Open > Items (system) ! Technical: Specifications 'TS: - ,. L .TSCR -- -Technical ~ Specification. Condition Report ' TVA . ' Tennessee Valley Authority , ' , Upper Head Injection (system). UHI - Unresolved' Item URI - Unit,0perator_'_ _ . 3 ' 00; - USQD - 'Unreviewed Safety Question Determination .VIO ~- . Violation'(of NRC requirements) i ' VLV - Valve -WCG: - --Work' Control Group ' WP -

Work Plan
WR

' ' Work Request -

- n

j _. - . . - - _ _ _ _ . _ _ _ _ - _ . - .--_--____________-_-.-_-----____.--_-_.-______--Q .

- -- ) i , I . . . f 56 APPENDIX A Administrative Control Program Items Reviewed 1. The inspector reviewed the following documents from the below-listed ACPs to determine that appropriate licensee document reviews had been made to ensure that CAQs were not present: - Work Request (WR) - SQM-2, Maintenance Management System B 753029 - Main Steam Safety Valve Leaks Thru. " - 0 283836 - SIS Accumulator Tank 2 Pressure Indicator - Failed Low. - B 769983 --RHR Pump 2A-A Discharge Line Flow Transmitter Low Side Fitting Is Gaulded. - B 769234 - Negative Rate Trip Light Will Not Clear. l ' B 283669 - Lower Radiation Monitor Is Not Indicating That - It Has Power. B 283663 - Auxiliary Feedwater Pump 2A-A, Add Oil To - Inboard Bearing. Potential Reportable Occurrence (PRO) - SQA-84, Potential - Reportable Occurrences. - 1-88-263 - Waste Gas Analyzer declared inoperable. - 1-88-274 - Seismic Recorder 0-XR-52-7SB was found inoperable when investigating WR B753263. j - 1-88-293 - Entered LC0 for Unit I when condenser vacuum exhaust flow rate was declared inoperable. - 1-88-307 - Snubber clamp for 1-FLV-70-87 was not in place when mode 4 was entered. - 1-88-320 - Fire Alarm was initiated due to fire on the Unit 1 turbine head end. Drawing Discrepancies - AI25 Part II, Drawing Deviations. - 88DD4114 - Revise drawing number 47W 0810-1 - 88DD4128 - Revise drawing number 47W 0805-2 -

- s,< . , ,' ' N z,;. . 57: - ' 88DD4164;-_ Revise drawing number 47W'0865-1 Revise drawing ' umber 47W.0834-2 - - 88004171.- ~ n ?. 88DD4173 -LRevise drawing number.47W 0865-2- -

Radiological Incident Report - RCI_1, Radiological Control - . Program. ' . - - RIR-88-30 - Individua1Eentered containment in violation of . 'RWP. sc - RIR-88-31 - Employee failed to sign-in on RWP. - , RIR-88-32 - Worker failed to process previous RIR's within t. 1 hour, violation of RCI-1. - - RIR-88-39 Worker performed personal decontamination- ' without notifying Radiation Control. ~ - - Housekeeping Deficiencies - SOA-66, Plant Housekeeping. - Routine monthly report November 1988. Nine items in waste packaging railroad bay. Routine' monthly report September 1988. Two items in- - additional. equipment building. > - Routine monthly report November 1988. Seven items in the Auxiliary Building in elevation 653. Test Deficiencies - AI47, Conduct of Testing. - ' SI-3, Oly, DN-1 - RM-90-119 failed to pass SI-302, WR - B769333 written. SI-167, DN-4 - Valve 0-26-1755 Tagged Closed, H0 - 1-88-1221 lifted. SI-531, DN-4 - 0-VLV-26-895, replace stem per WR - B296912 - SI-540, DN-1 - Extinguisher removed when building taken down. , SI-2, DN-2 - XA-55-60-2 will not clear, WR B283717 - written - - Problem Reporting Document - A-12 Part III, Corrective Action, Appendix I, Processing of Problem Reporting Document. -.. -_n--,--------n-x---- - , - - - - - .


_--------_---.-_--_-_----------.--___--------------_------.ww-----_.---------------------------,---------------------_-.---___--------_-----_---,-----,------__------a

- _ - . _ . _ - _ _ _

3.;.

E d, L 58 SQP 880007P - 56 of.72 WR's already closed in last 6 - months did not include a failure cause. I SQP 880027P - 6 uses of AI-47 deficiency log in - WP5591-01-1 outside interpretation of AI-47. - SQP 880034P - Worst. case vertical drop conduit walkdown sketch was found in error. SQP 880041P - The 1987 full scale-general fire drill has - not been entered into plant QA record system. - SQP 880077P - The Discriminator voltage for channel N31 was charged on a WR and the WR closed prior to resetting the channel. . NQA Audit Report (correct-on-the-spot) - QMI-328, - SQA 88816, COTS 1 - Employee's confidentiality was not - tagged. (on the employee concern report file) SQA 88816, COTS 2 - ECP only closed 1 of 2 concerns - - expressed in exit interview. - SQA 88804, COTS 1 - SDR 88-036-02 was not responded to. - SQA 88804, COTS 2 - No EA review of SQP 871677.

- SQA 88812, COTS 1 - An M&TE investigation did not include all activities for which the equipment was used. - SQA 88812, COTS 3 - Procedure revision request QSB-384 was i issued to correct incorrect reference ) in NQAM Part II section 2.4. j i - QA Surveillance Reports (Correct-on-the-Spot) -AI-32, Quality J Assurance Surveillance, QMI-702.6, QSQ-M-89-015, COTS-1 - A copy of the workplan was not at - the site. ] 1 QSQ-M-87-048, COTS-1 - A copy of the implementing - procedure was not at the worksite. - QSQ-M-88-986, COTS-1 - The M&TE utilized in SI-93 was ' logged, however, an improper SI-90.3 was listed. l ______m__ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ ___________._____.______________________o

.. .,- - , g.(;; y p n'.. , j 59 c , ' QSQ-M-87-174, COTS-1 - The MI-6.20 utilized in-WR'B275782 - was not verified to be the latesti revision. QSQ-M-88-128; COTS-1 - Penetration 774 ABA 12018000 was' - missing.a metal tag. .QC-Inspection Rejections-'(Inspection Reports'and - . Correct-on-the-Sp>t) - AI-20,-QC InspectionLProgram, AI-11, , Receipt Inspection and Non conforming Items. IR WBB-2489 . magnetic particle inspection - COTS on --~ cleanliness. . -IRW88-2505 - Fixup' inspection COTS on' documentation.

- - ' IRM88-0326 - Quality control inspection COTS on documentation. IRE 88-0977 - Quality control of gasket COTS to replace. - Licensee Event Reports - AI-18, Plant. Reporting Requirements. - 327/88-039 - Unauthorized personnel.using a portable: - radio inside Unit 1 containment generated a reactor trip signal. "327/88-022 - Reactor trip signals generated from -- electromagnetic interference caused by welding machine operated near source range nuclear instrument' cabling. 327/88-043 - Inadequate firewatch patrol resulted in a - noncompliance with TS 3.7.12. 328/88-030 - The failure to identify that the effects of - excessive post-trip reactor coolant system cooldowns could have caused noncompliance with shutdown margin requirements. 328/88-006 - Engineered safety feature main steam line - isolation and reactor trip due to maintenance activities and inherent instrument response during steam plant heatup. Security Degradation / Incident Report - PHYSI-29, Security - Degradation / Incident Reporting. - 87-088-10 - Deficient barrier from the protected area into a vital area as a result of a HVAC duct penetration. , L_-___-__-_-______-______ b ,

_ _ _ . - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ -4o . .o 60

87-087-10 - Deficient Vital Area Barrier. - - 87-063-04 - Individual Piggybacked Through Vital Area Door. 87-087-05 - Failed to Return Badge Within Allowable Time - After Leaving ERCW. - 88-094-05 - Individual Accepted Wrong Badge At Access - Portal. - 88-118-06 - Failed to Return Badge Within Allowable Time After. Leaving ERCW. 2. The June 30, 1988, Semiannual Component- Failure Trending Report was reviewed. It identified the following components as having trends and documented the corrective actions taken: , 1-FCV-087-0021 UHI ISOLATION VALVE 1-VLV-002-0521 AUTOMATIC MAKEUP LCV ISOLATION VALVE ! 2-FSV-043-0287-A POST ACCIDENT SAMPLING VALVE 2-FSV-082-0231 DIESEL GENERATOR AIR START SOLEN 0ID 0-FSV-082-0160 DIESEL GENERATOR AIR START SOLEN 0ID l 1-VLV-062-0901 MIXED BED DEMIN INLET VALVE 1-PMP-002-0020 HOTWELL PUMP 1-FCV-043-0022 REACTOR COOLANT HOT LEG SAMPLE HEADER CONTAINMENT ISOLATION VALVE 0-VLV-082-0501-1A1A DIESEL AIR START COMPRESSOR RELIEF VALVE 0-VLV-082-0534-1A2A DIESEL AIR START AIR TANK RELIEF VALVE 2-FT-068-0006A RCS LOOP COOLANT FLOW TRANSMITTER 1-TS-030-0194 714 PENETRATION ROOM COOLING FAN 1-PDIS-067-0491E/F SWITCH 1-FT-001-0010B-E STEAM GENERATOR MAIN STEAM HEADER FLOW TRANSMITTER 1-FS-090-0106A LOWER COMPT. AIR MON. PART. LO FLOW SWITCH 2-FM-003-35 i to p CONVERTERS 2-PX-003-170 ISOLATED DC POWER SUPPLY 2-FR-002-0035 CONDENSER HOTWELL PUMP FLOW RECORDER 0-TR-082-5036/1 DIESEL GENERATOR WINDING RTD. ]-H2AN-043-0200 CONTAINMENT HYDROGEN ANALYZERS 1-RM-080-0106A RADIATION MONITORS 1-VLV-067-0585C-A ERCW RETURN UCV COOLER CHECK VALVE 1-PMP-002-0033 HOTWELL PUMP f 0-PMP-067-452 ERCW PUMP 2-FCV-063-0064 SIS ACCUMULATOR TANK N2 HEADER INLET VALVE 1-FCV-001-0182 STEAM GENERATOR BLOWDOWN CONTAINMENT ISOLATION VALVES i -__-___________A

. _ _ Ng y .: ' e g:.,

L l 61 ' .j ' .;e I ' - -1-LCV-003-0174- STEAM GENERATORzLEVEL CONTROL VALVE 2-FSV-043-0287-A POST ACCIDENT SAMPLING CONTAINMENT . AIR PNL ISOLATION VALVE " - 1-FCV-003-0191 FEEDWATER LONG CYCLE VALVES AND HEATER < , . . ISOLATION VALVES- 2-ENG-082-0002B1 DIESEL GENERATOR ENGINES 0-VLV-082'0501-1A1A ' COMPRESSOR-RELIEF VALVE 1-FCV-062-0086 CHARGING FLOW REACTOR COOLANT SYSTEM-

..

. COLD LEG LOOP FLOW CONTROL VALVE - 0-VLV-082-05161A1A' - TANK CONNECTION SHUT 0FF VALVE - 1-PMP-082-0001B/2-B' ~ DIESEL GENERATOR LUBE-0IL CIRCULATING PUMP 1-PCV-001-0012-B STEAM GENERATOR MAIN STEAM HEADERL PRESSURE RELIEF CONTROL VALVE 0-VLV-082-0528-1828L COMPRESSOR RELIEF VALVE , , s !- . __ ______-_.__m____. _____.___m__ _ _ _ . _ . _ _ . _ _ _. _ _ _ - }}