ML20056A478: Difference between revisions

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{{Adams
#REDIRECT [[IR 05000327/1990022]]
| number = ML20056A478
| issue date = 07/26/1990
| title = Insp Repts 50-327/90-22 & 50-328/90-22 on 900606-0705. Violation Noted.Major Areas Inspected:Operational Safety Verification,Including Control Room Observations,Operations Performance,Sys Lineups & Radiation Protection
| author name = Harmon P, Little W
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
| addressee name =
| addressee affiliation =
| docket = 05000327, 05000328
| license number =
| contact person =
| document report number = 50-327-90-22, 50-328-90-22, NUDOCS 9008080003
| package number = ML20056A476
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 23
}}
See also: [[see also::IR 05000327/1990022]]
 
=Text=
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                      [ [, p nas                          o                      NUCLEAR REGULATORY COMMISSION                          ,
                                                                                                                                                      *
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              W -*                                              :
                                                                                        101 MARIETTA STREET.N.W.
      +
                    *
                              ' I* s                          a                        : ATLANTA,OEORGIA 30323                                )
                                    *                                                                                                          ~
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                                /y...../
                                      .
                                                        <
                ,
                                      x Report Nos.: _50-327/90-22 and 50-328/90-22-
  '                            '
                                            Licensee': Tennessee Valley Authority
                *U                                                6N.38A Lookout Place-                                                              3
    #'o.                                                          1101_ Market Street                                                                '
      , ,                                                        Chattanooga, TN. 37402-2801
g a
"
                                                                                                                                                  .
                  '
                                        lDocketNos.:                  50-327'and 50-3281            License Nos.: DPR-77 and DPRa79                  i
                      ,
                                  '
                                              Facility Name:            Sequoyah Units 1 and 2
                                                                                                                                                  l    '
          '
*
                                              Inspection Conducted: June-6, 1990 thru July-5,- 6 0                                                  _;
                                        ' Lead Inspector: LAAMA @                                                          7      70
                                                                        Pf/ Harmon~, pi@ Rbsident Inspector                Date Signed
                                          . Inspectors:                D. Loveless, Resident Inspector.                  6          4
  [g        '
                  c,                                                  J. Brady, Project Engineer
                        *
                                        . Approved by:                          ct                                              ,    D
                                                                        W. 5. Little, Chief, Projec't Section_1 -
                                                                                                        r                    Date Signed              1
                                                                        TVA Projects
      7                                                                                                                                            ]-!
),
                                                                                            . SUMMARY'                                                  i
                                                                                                                                            ,
6                                    1 Scope:-                                                                                                      -,
                    '
      '"3,                              -This> announced inspection involved inspection' effort .by, the Resident Inspectors                ,f
  A,
                                              in 1 the E area; of . operational safety verification? including _ control room
                                                                                                                    -
                                                                                                                                                        I
                                          , observations.Loperations performance, system lineups, radiation protection,                                  4
  sd            t                            safeguards, and conditions. adverse.to quality. LO.ther areas inspected included
                                              surveillance testing observations maintenance observations,, review of previous
                                              inspection findings, foilow-up of events, review of, licensee identified items,_
, '', ,                                  : and revi_ew .of inspector follow-up items. .Also included in this report is
                                          - documentation of special inspection findingc in'the area.of cable testing.
  3E
  '
                                                                                                                                        ,
                                                                                                                                                            ;
                                        < : Results:
                                *
                                            iWea'knesses were noted in the licensee's control of overtime for plant personnel
  y
  "'                                      ' as described in_ paragraph 7, 'and' use of ' administrative control programs in
            ,                                response to unanticipated changes in core differential temperature.'
                                              The areas-of Operations, Maintenance,.and Surveillance were a_dequate and fully
                                              capable ~ to supporti current plant operations. The observ_ed activities of the
                                              control room operators were professional and well executed.
                                                    9008080003 900726
                                                    PDR
                                                    O              ADOCK 05000327                                                                            i
                            >
                                                                                PDC                                                                            (
        ic
              fi                ,          ;
                                                  '
 
'y                      -,o
      i. ~ . :        .
                          . .,      ,
              ..                  .
                                -
          k    .
                  s-
                    -
                              .{.
                                                                                '2
w
    '
                            'One violation was identified which involved a failure to control overtime for
                                plant' personnel.
                                        -327,328/90-22-01, Exceeding of Overtime Limits Without Proper Approval
                                          - paragraph 7.
                                Two unresolved items * were identified pertaining to cable testing issues.
                                          327,328/90-22-04, Unissued Calculation for 1E Cable Testing - paragraph
                                          10.
  ,
                                      '
                                          327,328/90-22-05, Assumption Not Identified in Calculation - paragraph 10.
                                  One non-cited violation was identified involving a loss of a shutdown board
                                  during surveillance testing.
                                          327,328/90-22-03, Partial Blackout Caused By Energizing Circuit With
                                          Testing Rig Still Connected - paragraph 9.c.
                              :  One. licensee-identified violation was identified involving an inadvertent entry
                                  into TS 3.0.3' and placing the plant in an unanalyzed condition during
                                  containment purge operations.
                                          327,328/90-22-02, Unit 1 in Unanalyzed Condition During Containment Purge
                                          Operations - paragraph 9.b.
                                  No deviations or inspector follow-up items were identified.
                          '
                                  Three events that occurred during the inspection ; period are described in
            '
                                  paragraph 9. The events were a partial blackout and emergency diesel start due
                                  to improper testing of a shutdown board; inadvertent entry into TS 3.0.3 by
                                  placing Unit 1 in an unanalyzed condition during containment purge operations;.
                                  and. identification and resolution of anomalies in core delta T identified
                                                                                                                    -
                                  during the Unit 1 startup from the refueling outage.
                                    *      Unresolved items are matters for which more information is required to
                                            determine whether they are acceptable or may involve violations or
                                            deviations.
                                                        _ _ _ _ .    . . . . .
 
    .,    .
          ..      .
        -        .
  ,.  .  .. .
                                ..,
l
                                                  REPORT DETAILS
              1.    Persons Contacted
                      Licensee Employees
                    *J. Bynum Vice President, Nuclear Power Production
                      W. Byrd, Manager, Project Controls / Fir.ancial Officer
                      C. Vondra, Plant Manager
                    *R. Beecken, Maintenance Manager                                            .
                                                                                                  .
                                                                                                  '
                      L. Bryant, Work-Control Superintendent
                    *M. Burzynski, Site Licensing Manager
                    *M. Cooper, Compliance Licensing Manager
,
                      J. Gates, Technical Support Manager
                    *W.  Lagergren,-Jr., Operations Manager
                    *M. Lorek, Operations Superintendent
                      R. Lumpkin.-Site Quality Manager
                    *T. Flippo,, Quality Assurance / Quality Engineering Manager
                      R. Proffitt, Licensing Engineer
                    *H. Rogers, Technical Support Program Manager
                    *M.'Sullivan, Radiological Control Manager
                      P. Trudel, Project Engineer
                      C. Whittemore, Licensing Engineer
                    *G. Hipp, Licensing Engineer
                    *R. Thompson, Licensing Engineer
                      NRC Employees
                    *L. J. Watson,.Chitf, Project Section 1
                    * Attended exit interview.
                      Acronyms and initialisms used in this report are listed. in the last
                      paragraph.
              2. -  Operational Safety Verification (71707)
                      a.    Control Room Observations
                            The inspectors conducted discussions with control room operators,
L
                            verified that proper control room staffing was maintained, verified
                            that access to the control room was properly controlled, and that
                            operator attentiveness was commensurate with the plant cor#iguration
l                          and plant activities in progress, and with on-going contro' room.
                            operations. The operators were observed adhering to approp iate,.
                            approved. procedures, including Emergency Operating Procedure 3, for
                            the.on-going activities. The inspectors-observed upper management in ,
                                                                                                '
                            the control room on a number of occasions.
 
                    '
      o              ' "
,
        ' M.              ,  . , ,
  '                      ''
          ;+                ..
4
                  .
                    ..
  ,
                                            ,
  e                                      The inspector verified that the licensee was operating the plant in a
                                        normal plant configuration as required by T3 end. when abnormal
              ,
                                        conditions existed, that the operators were complying with the
                                        appropriate LC0 action statements. The inspector verified that RCS
                                        leak rate calculations were -performed and- that leakage rates were
                                        within the TS limits.
            *
                                        The inspectors observed instrumentation and recorder traces for
                                        abnormalities and verified the status of selected control room
                                        annunciators to ensure that control room. operators understood the
                                        -status of the plant. ' Panel indications were reviewed for the nuclear
                                        instruments, the emergency power. sources, the safety parameter .
                                        display system and the radiation monitors to ensure operability'and
      '
                                        operation within TS limits.
                                        No violations or deviations were observed,
                                    b.  Control Room Logs
b                                      The inspectors observed control room operations and reviewed          i
                                        applicable logs including the shift logs, operating orders, night
                                        order book, ' clearance hold order- book, and , configuration log to
                                        obtain information concerning operating trends and activities. The
                                        TACF -log was reviewed to verify.that the use of jumpers and lifted
                                        leads causing ' equipment to be inoperable was clearly. noted and
                                        understood. No issues were identified with these specific logs.
                                        Plant secondary chemistry reports were reviewed. The inspector
                                        verified that primary plant chemistry was within TS limits.
                                        The implementation of the licensee's sampling program was observed.
                                        Plant specific monitoring systems including seismic, meteorological
                                        and fire detection indications were reviewed for operability.      A
                                        review of surveillance records and tagout logs was performed to
                                        confirm the operability of the RPS.
                                        'No violations or deviations were observed.
                                    c.  ECCS System Alignment
                                        The' inspectors walked down accessible portions of the Unit 1 charging
  -
                                        and letdown portions of the CVCS System, to verify operability, flow
                                        path, heat sink, water supply, power supply, and proper' valve and
                                        breaker alignment.
                                        The inspectors verified that a selected portion of the containment
                                          isolation lineup was correct.
                                          No devietions or violations were identified.
    '          .
 
                                                                                                                        *
  NW
  m                              .
                                                                                                                              :
      W                      ;.9~    .
                                                                                                                          1
  g                    ~,          .                                                                                ,    o
  r                    .        e,-                                                                                .
                                                                                                                            j,
                            .
                '
                                                                              3
                                                                                                                            j
, . _
  %w ,
                                          ,    .
      [                                d. Plant Tours
                                            Tours of the diesel generator,- auxiliary, control, and turbine
                                            buildings, and exterior areas were conducted to observe plant
        m
                              '
                                            equipment conditions, potential fire hazards, control of ignition
                                            sources, fluid leaks, excessive vibrations, missile hazards and plant
          '
                                            housekeeping 'and cleanliness conditions.      The plant was observed to            l
  ,
      '
                                            be clean and in adequate condition. The inspectors verified thet                  ,
      9                                    maintenance work orders had been submitted as required and that
        &                                  followup activities and prioritization of work was accomplished ty
  g                                        the licensee,                                                                    f
  4
      d                                    The inspector visually inspected the major components for leakage,                q
    +                                      proper lubrication, cooling ' water supply, and any general condition          n
                                            that might prevent fulfilling their functional requirements.
              '
                                                                                                                            ,
            ,
                                            The inspector observed shift turnovers and determined that necessary.
              ,,                            information concerning the plant systems status was adoressed.                _;
                    '
                                            No violations or deviations.were observed.                                      l
  llq '                                  e. Radiation Protection                                                              ,
        ,            ,                                                                                                      ,
                i                          The inspectors observed HP practices and verified the implementation
                                            of radiation protection controls.      On a regular basis, RWPs were
  m                                        reviewed and specific work activities were monitored to ensure the
  '
                                            . activities were being conducted in accordance with the applicable              r
                                            RWPs.    Workers were observed for proper frisking upon exiting
                                            contaminated areas and the radiologically controlled area. Selected
                  '
                        e                    radiation protection instruments _were verified operable and
                          4                calibration frequencies were reviewed.                                        _i
                                            -No violations or deviations were identified,
                                        f.  Safeguards Inspection                                                          !
      o m
                                            In the course of the monthly activities,- the inspectors included a            '
                                            review of the licensee's physical security program. The performance          1
                                            of various shifts of the security force.was observed in the conduct
                                            of daily activities including: protected and vital ' area access
      ,
                                            controls; searching of personnel and packages; escorting of visitors;
g                                            badge issuance and retrieval; and patrols and compensatory posts,
m'
                                            The . inspectors observed protected area lighting, and protected and
H                                            vital areas barrier integrity.      The inspectors verified interfaces
        F                                    between the security organization and both operations and
                                            maintenance. Specifically, the Resident Inspectors:
                                                  -      witnessed firearms training and qualification
                                                          interviewed individuals with security concerns
,
'                                                  -
l
                                                  -      visited central and secondary alarm station
                                                  -
                                                          verified protection of Safeguards Information
                                                  -      verified onsite/offsite communication capabilities
  *
                                            No violations or deviations were identified.
i.
 
  **    '. . ,
        e
    :.      .
      .
        ,'.
                                                        4
                        .
                                                                                              t
                                                                                              5
                g.  Conditions Adverse to Quality-
                                                                                              t
                      The inspectors reviewed selected items to determine that the          o
                      licensee's problem identification system as defined in STD 3.1.1,    ,7
                      Corrective Actions, which supersedes AI-12, Corrective Action, was
                      functioning.    CAQR's1 were routinely reviewed for adequacy in
                      addressing a problem or event. A sample of the following documents
                      were reviewed for adequate handling:
                            -
                                  Work Requests
                            -    Potential Reportable Occurrences
                            -    Radiological Incident Reports
                            -    Problem Reporting Documents                                4
                            -    Correct-on-the-Spot Documents                              !
                                                                                              t
                            -    Licensee Event Reports
                      Of the items reviewed, each was found to have been identified by the
                      licensee with immediate corrective action in place. For those issues
                      that ~ required long term corrective action the licensee was making
                      adequate progress.      One exception to this was identified in .
                      paragraph 10 relative to.1E cables.
                      Corrective actions' teken in response to the unanticipated core        -
                      differential temperature were effective. Details-are described in
                      paragraph 9.a.
                      No violations or deviations were observed,
                ho trends were identified in the operational safety verification area.
                General con'ditions in the plant were adequate.
                                              .
  g              Radiation protection and security are adequate to continue two unit
                operations.-
                                                      '
            3.  Surveillance Observations and Review (61726)
-                Licensee activities were directly. observed / reviewed to ascertain that
                surveillance of safety-related systems and components was being conducted
                in accordance with TS requirements'.
                The inspectors verified that: testing    L., performed in accordance with
  -
                adequate procedures; test instrumentation was calibrated; LCOs were met;
                test results met acceptance criteria and were reviewed by personnel other
                .than the: individual directing the test; deficiencies were identified, as    4
                appropriate, and any deficiencies identified- during the testing were
                properly reviewed and resolved by management personnel; and system            1
                - restoration was adequate. For completed tests, the inspector verified        ;
                that testing frequencies were met and tests were performed by qualified        l
                individuals,
                                                                                                i
 
        93                                                                ,+
          f.h      ,
                        ' ' o
                                                                                                            ,
                                                                                                            '
        y          .,
                          '
          &l        .-.
  !} ; 1                                                          S
      '
            '
                '
.
$                            During surveillance testing on June 25, 1990 at 9:21 a.m., with both units
        ,                    at 100 percent power, an improper test sequence resulted in loss of the
                              6.9 kv shutdown board 1A-A and the start. of all emergency diesel
                              generators.    The IA-A shutdown board was reenergized by the 1A-A EDG and
n.                          loads were sequenced back onto the board. . Details of this event are        <
    *              *
                              described in paragraph 9.c.                                                    ;
  -        *
                              The following activity was observed / reviewed with no deficiencies            l
    -
                              identified:
  I"                                DPS0-SMI-4C, Indicating Voltmeter Calibrations for 6900 Volt Shutdown
              1
                                    Boards.                                                                  l
                              No trends were-identified in the area of surveillance performance during    .j
                              this inspection period.      The area of surveillance scheduling and            '
  ' '
                              management was observed to be adequate.                                        ;
          '
                              One non-cited violation in the area of surveillance activities -is
  4                          described in paragraph 9.c.
                          4.  Monthly Maintenance Observations and Review (62703)                            !
                              Station maintenance activities on safety-related systems and components      I
                              were observed / reviewed to ascertain that they were conducted in accordance  ,
                              with approved procedures, regulatory guides, industry codes and standards,
                                                                                                            '
                              and in conformance with T.S.
                              The following items were considered during this review: LCOs were met        ~!
                              while components or systems were removed from service; redundant              j
                              components were operable; approvals were obtained prior to initiating the        j
                              work;: activities were accomplished using approved procedures and were        1
                              inspected' as applicable; procedures used were adequate to control the          j
                              activity, troubleshooting - activities were controlled and the repair-          l
                              records accurately reflected the activities, functional testing and/or          3
                              calibrations were performed prior to returning. components or systems- to      !
                              service, QC records were maintained, activities were accomplished by
                              qualified personnel, parts and materials used were ' properly certified,
                              radiological controls were implemented, QC hold points were established      ]
                              where required and were observed, fire prevention controls were
                              implemented, outside contractor force activities were controlled .in
                              accordance with the approved QA program, and housekeeping was actively
                              pursued.
                              The following work request was reviewed:                                        ;
                                                                                                              i
                                    WR C001716 -    Following an ABI initiated by performance of an SI,        q
                                                    the licensee proceeded to search for the leak that        ;
                                                    had been causing auxiliary building airborne
                                                    radiation problems for several months. A leak was          ,
                                                    found on 2-FT-62-137, emergency boration flow              *
                                                    transmitter, causing 3800 MPC of noble gas in the
                                                    local area. This leak only existed during certain
                                                                                                                ,
 
                m
                                                                  ^
    ,
                                                                                                            ,
  /,            .
                ''
  ?[,
                          '
                  -
                        ;
.      ,
              j    .                                                                                        !
                                                                    .6                                      i
                              .    .
            3
      e
              i
  "
                                                    evolutions and the Auxiliary Building ventilation
                                                    system caused quick dispersion.    Rusted bolts were    j
                                                    removed from the fitnge and new bolts torqued to        j
                                                    normal specifications. This isolated the leak and the
                                                    airborne zone was cleared the next day.                  ;
                      5 .-  Management Activities in Support of Plant Operations                            1
                            TVA management activities were reviewed on a daily basis by the
                            inspectors. Resident Inspectors observed that planning, scheduling, work;
                            control and other management meetings were effective in controlling plant
                            activities.    First. line supervisors appear to be knowledgeable and
                            involved in the day to day activities of the plant. During the inspection
*                            period, management involvement in resolving the problems associated with
                            the observed core Differential -Temperature was conservative, and
                            effective. The event was well managed with the exception of the failure          ,
                            to promptly initiate the approved problem reporting procedure as described
                            in paragraph 9 a.        First line supervisor involvement. in the field has.
                            been observed and appeared adequate.
                      6.    Site Quality Assurance Activities in Support of Operations (71707)
                                                                                                                l
                            The inspectors discussed the overtime problems addressed below with site
                            QA.    Following monitoring activities QA issued a CAQR (SQQ900287) on
                            overtime.    The ' CAQR indicated that Technical Support, Maintenance,
                            Radiation Control,' and Modifications were not meeting the guidelines of
                            Al-30 Section 23.0. The inspector reviewed additional data and determined          j
                            that-Operations and Chemistry were also not meeting these guidelines. QA          1
                            was questioned'as to why their monitoring did not pick up these critical
                            organizations. . They stated that . only organizations that were not
                            implementing the' criteria at all were documented in the CAQR. In
                            this. specific instance, QA did not appear to be aggressive in pursuing the        I
                            issue.
                      7..    NRC Inspector Follow-up Items,-Unresolved Items, Violations (92701, 92702)
                              (Closei.) URI- 327,328/90-20-01, ' Overtime Rules Inadequate and not
                              Implemented.                                                                      .
                                                                                                                l
                                                                                                                '
                            The inspector reviewed Al-30, Nuclear Plant Conduct of Operations,
                            Section 23.0, Plant. Staff Overtime Limits.        These requirements were
        .                  reviewed in three areas from Generic Letter 82-12 as follows:
                              1)  Enough plant operating personnel should be employed to maintain-
                                  adequate shif t' coverage without heavy use of overtime. The objective
                                  is to have operating personnel work a normal 8-hour day, 40-hour week
                                  while the plant is operating.
                                  AI-30 states that rormal shifts will be 8 hours in length.        Every
                                  effort will be made to eliminate 16 hour shifts. No 16 hour shifts
                                  will be approved unless the circumstances are extreme.
          ,
                                          w
 
                                                                                                        ]
                                                                      '
                    '
*
        ,.
                  .
                j .
                      z, ;
        .:          ~, ,
F      . _ v.7
            ..;
y                                                                7'                                    3
        '
                                                                                                        l
                                Nuclear Power' Standard STD-2,1.7, Administration of- Overtime,          '
                                Revision 0, interim. change 2, restates the above NRC guideline in its~
,
                                entirety in Section 2.1.3.
                                                                                                          '
                            2)  In the event that unforeseen problems require substantial amounts of
                              . overtime to be used, or during extended periods of shutdown for
                                refueling, major maintenance l or major plant modifications. on a
                                temporary basis, the following guidelines shall be followed:            ;
  \
                                a)    An individual should not be permitted to work more than 16 hours
                                      straight (excluding shift turnover time).                            l
                                b)    An individual-should not be permitted to work more than 16 hours
                                      in any 24-hour period, nor more than 24 hours in any 48-hour
                                      period, nor more than 72 hours in any seven day period (all
                                      excluding shift turnover time),
                                c)    A break of at least eight hours should be allowed between work
                                      periods (including shift' turnover time).
,                              d)    Except during extended shutdown periods, the use of overtime
'
                                      should be considered on an individual basis and not for the
                                      -entire staff on shift,
                                Recognizing that very -unusual circumstances may arise requiring
                                deviation from the -above guidelines, such deviation shall be
                                authorized by the plant manager or his deputy, or higher levels of
                                management.
                                                                                                        ;
                                AI-30 embraces items a through c of the above guidelines, and among
    -e                        others states that the Maintenance, Operations, Technical Support,      ,
'
                                Division of Nuclear Engineering, Radiation Control and Chemistry
                                organizations are required to follow these guidelines. Deviations to
                              ~ these guidelines require prior plant manager (or designee) verbal
                                approval, with the exception that the SOS may approve shifts greater
                                                                -
                                than 12 hours with written notification to the Operations                !
                                Superintendent.-
                                STD-2.1.7 restates the above NRC- guidelines in their entirety in        :
                                Section 2.1.3.A for the same staff as listed in AI-30. Additionally,
                                for deviations' to these guidelines the standard requires that' the
                                following criteria be met:
                                      -
                                            Very unusual circumstances exist.
                                      -    Significant reduction in the effectiveness of operating
                                            personnel would be highly unlikely.
              4
                                      -    Document the reason for authorizing the deviation.
                                                                                              .
 
..  .
      ,
    ''
  .
        ,_
.
    ,.
                                                  8
                    .
            3)    Procedures-a're encouraged that would allow licensed operators at the  .
                  controls to be periodically relieved and assigned to other duties
                  away from the control board during their tours of duty.
                  AI-30 states that if an operator is required to work in excess of 12
                  continuous hours, their duties should be carefully selected. It is-
                  preferable that they not be assigned any task that affects core
                  reactivity or could possibly endanger the safe operation of the
                  plant.
            The inspector reviewed the requirements under item 1) above and fotermined
            that while the operations organization is staffed at levels conensurate
            with two unit operation, often the licensed operators are not utilized to
                                                                            ~
            the extent necessary to reduce excessive levels of overtime. During the
            recent Unit 1 outage operations personnel were pooled for overtime usage.
            Thus the usual high levels of overtime seen during an outage were also
            seen for operators running the operating unit. From March 14 through
            March 27,1990, while Unit 2 was running at 100% power the ASOS's were
            working six or seven 12 hour shifts per week as a result of planned
            overtime for the entire SR0 staff.
            Additionally, the inspector noted that overtime was used regularly during
            times of_ both units 'in operation. In a period from January 1 through
            March 4,1990 a random selection of 72 man-weeks were reviewed and found
            that 21.-included overtime in excess of normal shif t turnover. Staffing
            levels and normal usage of overtime were not reviewed for other plant
            sections.
            The _ inspector reviewed plant usage of the guidelines listed under item 2)-
            .above.- Following questions asked by the inspector, site QA issued CAQR
            SQQ900287 for " numerous ' cases of personnel exceed lng the overtime
            (requirements) of Al-30 with no, documented authorization" in Technical
            Support, Maintenance, Radiation Control,- and Modification. Therefore, the
            inspector did not review these sections further.
            Numerous examples were found in the Operations section .where' overtime
            requirements - were exceeded without documentation or meeting' other
            guidelines as required.      From March 12 through April 29, 1990, the      ;
              inspector noted- 14 examples of individuals working more than 72. hours
              in .a seven day period without an AI-30 Attachment E being prepared.
            A review of AI-30 Attachment E's dating from November 13, 1989 through
            March 23, 1990 showed fourteen examples of prior verbal approval not being  !
            obtained and no ex'planation as to why prior verbal approval was not
            obtained; six examples of sixteen hour shifts being authorized for reasons  I
              that appeared' to' be routine when the procedure requires extreme cases
            only; and four cases where overtime in excess of the guidelines was
            authorized without a documented reason per procedure.
 
                                                                                          ,
.
        ,
            ,
      '
  -
          .
    -
... ,.
                                                    9
                .    .
                                                                                                    .
              Additionally, a review :of payroll records for March 1990 showed a
              Chemistry first-line supervisor worked greater than 72 hours _ in a seven
              day period without proper documentation or approval.
              During tnis inspection period the inspector discussed a six hour-
              turn-around time scheduled for an AUO. The SOS stated that the individual
              had a normal- 8 hour turn-around scheduled. However, he had worked 4
              overtime hours that morning. The shift clerk _had tried to reach him to.
              tell him not to return until he had been off six hours, but had failed.
              The inspector asked the SOS the significance of six hours when AI-30'said            ,
              eight._ The SOS replied that this was a Union requirement, that the Al-30            ,
              requirements were broken all the time, and that six hour turn around times
              were a normal occurrence. When asked if management needed to be informed,            -
              the SOS stated that it was not necessary. After further discussion, the
              SOS called the Plant Manager who denied the six hour turn ~ around and
              informed the'S0S that these were considered unaccepteble.
              The inspector . reviewed the implementation of the guideline stated in
              item 3) above and did not find evidence of any formal implementation.
              Discussions with operators, senior operators and duty plant managers
              showed that consideration of overtime for job assignment was not being
              performed. Individuals interviewed stated that they would consider it in
              the future but .no formal method or program existed other than the
              suggestion in Al-30._ The inspector identified four cases on May 1, 14, 22
              and 24 respectively where Unit Operators on the operating unit'were
              required to work; greater than 12 hours without being transferred from lead
              operator to B0P or.to the non-operating unit.
              The exceeding of the Al-30 overtime limits without proper approval is
              considered a violation of T.S. 6.8.1 for failure to follow procedure
              and is designated as VIO. 327, 328/90-22-01. This violation'is    a repeat
              of Violation 327,328/87-78-01 issued March 14,.1988.
              Generic Letter 82-16 requested that plants submit a TS requiring
              compliance with the overtime guidelines as stated in GL 82-12. TVA did
              not submit a technical specification at that time.
                    -                                                  This position was
              accepted by the NRC .in the SER documenting review of GL 82-16.    The SER          '
              stated that Al-30 procedural -limits are considered sufficient for Sequoyah
              to meet the_ requirements of the GL. However, the SER stated that if the
              staff. should identify any significant problems in the implementction of
              this procedure, the staff would request TVA to propose that the madel TSs
              on overtime in the GL be - added to the Sequoyah TSs. Because of the
              broad ~ scope and apparent lack of management attention for the overtime -            '
              problems, this item is being forwarded to NRR with a recommendation .that -
                they request that TVA submit the standard TS as written in GL 82-16 for              *
              addition to the Sequoyah TSs.
                                                                                                    ,
              'The followup of plant overtime problems will be addressed in
              VIO 327,328/90-22-01 above; therefore, URI 327,328/90-20-01 is closed.
                                                                                            . _ _ .
 
                  -                                                                                _
                                                                                                        -
  . . .      .
            .,i
      .
,                .
    .
          ,7 .-
                                                          10
                      .      .
                    (0 pen)URI 327,328/90-06-05, Resolution of SSONI Issues,
                    a.  The following . observations were noted during Inspection Report.
        '
                          327,328/86-68- as referenced in the stated report: sections. These
                          items were addressed in Inspection Report 327,328/90-06 as needing        4
                          further NRC' review:
                                Section    Observation    Description
                                2.1.2.5:  0-2.1-1        Lack of controls .to properly              i
                                                            identify and. segregate material storage.    ;
                                                            areas.
                                2.5.1      0-2.5-1        Root cause of valve damage not
                                                            identified.
                                2.5.2      0-2.5-2        No evaluation of maintenance                ;
                                                            deficiency.                                .
                                  2.5.3    0-2.5-3        Vendor instructions not included
                                                            in site work procedures.
                                  2.5.6    0-2.5-4        Root causes of hardware failures
                                                            not identified.                          .;
                                  2.6'      0-2.6-1        QA/QC coverage of modification
                                                            work,
                                  2.6        0-2.6-2        Two-party verification of                  i
                                                            maintenance activities.
                          The above listed. items represent inspector observations at the time
                          of the inspection and are not regulatory requirements. These were
                            addressed in an internal NRC memorandum dated May 31,1990 from
                            Pierson to Watson. Given the breadth and depth of TVA's Sequoyah-
                            Nuclear Performance Plan, the causative' factors precipitating these
                            observations were determined to have been corrected. These items
                            are, therefore, administratively closed,
                    b.    Observation 0-2.1-1, Material Control-Issues, wcs resolved during the
                            procurement-inspection documented in IR 327,328/88-07 which provides
                          justification for. closure of this observation,
                    c.-    IR .327,328/86-68, Section 2.3.5, addressed deficiency 0-2.3-5,
                            Installation Different from Flow Diagram. This item was addressed in
                          'IR'327,328/87-40 and left open.      The inspector stated that, while
                            this deficiency was not used to support a violation, it was
                            identified in Section 2.3.4 of the Report (IR 327,328/86-68) as a
                            deficiency, as well as in Table I of the report.
                                                                                                      .
 
          '
    :                                                                                            .
        'v      3,                                                                                  '
          .
        .:
              [y  ..
                                                                                                      -
                                                          11
  '
                                                                                                  ..
                                                                                                      r
  a-
              '
                          On June 12, 1990 TVA provided the Resident Inspector with a copy of        '
                          Drawing Deviation #87 DD 2219 and the new drawing to correct this
'
                          situation. D-2.3-5 is closed.
                      d. The inspector reviewed licensee action. on VIO 327,328/86-68-05,
                          Safety Related Equipment Was Not Installed in Accordance with the
                          Instructions, Procedures or Drawings, associated- with Deficiency
                          D-2.4-7, Resistors Not Properly Mounted to Terminal Blocks. The            +
                          violation indicated that field inspection found that resistors added        c
                          to the control circuits in Auxiliary Control Panel 2L-11A on Terminal      ;
                          Block 3-30 were not mounted to the panel but were held in place only
                          by their jumper wires. TVA admitted the violation stating that
                          improperly mounted resistors discovered during the inspection were
                          the result of a lack of design-approved installation criteria for
                          terminal _ blocks, ' resistors, and diodes.      TVA stated that the
                          resistors were' remounted in a, ~ edance with a revision to WP 12188
                          and TVA drawing 47A348-284.        ... IR 327,328/87-40 the inspectors
                          reviewed the corrective actions for this deiiciency and identified no
                          problems. The item was left open for verification of corrective
                          action to be accomplished during a future inspection. The inspector-
                          observed the mounting of the resistors in question and determined
                          that they were now properly mounted to the panel. Deficiency D-2,4-7
                          of VIO 327,328/86-68-05 is closed.                                          '
                      e. The inspector reviewed Unresolved item 327,328/86-68-14, documented
                          and tracked in IR 327,328/86-68 as U-2.4-1, Unclear Status of Freshly
                          Painted Welds. The inspectors observed a number of welds on reworked
                          pipe supports which were freshly painted. Although this indicated'-
                          that the welds were new, there was no identification of these joints
                          on the WP weld maps, nor were there any-'new weld documentation
      e                  records. The inspection team was unable to determine from the
                          records whether these were new, undocumented welds, or old welds that.
      '
            ,
                          had been repainted by the craft.                                            '
                          The licensee has contacted personnel that were associated with the
                          SSOMI inspection. TVA personnel contacted had no knowledge of what
                          welds were involved and many. people that may have had knowledge of          I
                          this item have left Sequoyah and could not be contacted.          The
                          licensee stated that they were not aware of any new undocumented            ,
                          welds. Following this 1986 inspection TVA committed in Section III.8
                          of the Sequoyah Nuclear Performance Plan to evaluate the adequacy of
                          the TVA welding program for all of the TVA plants and the suitability
                          of welded structures and systems for service,
                                                                                                      l
                          This program was reviewed, inspected and approved by the NRC in              l
                          NUREG-1232, Volume 2, Safety Evaluation Report on Tennessee Valley          I
                                                                                                      '
                          Authority: Sequoyah N iclear Performance Plan, Section 3.5 dated
                          May 18, 1988, and its supplement dated February 3, 1989. ^. herefore,        i
                                                                                                      '
                          U-2.4-1(327,328/86-68-14) is considered closed.
                                                                                                      l
                                                                                                      l
 
                                                                                            .-
                                                      -,
                                                                                                    '
    m..    .
              .4
            '
      .        ,
      w  ;?
          '
                                                        12                                        l.
                                                                                                      '
-i ,
                  f.    The inspector reviewed the. item identified in IR 327,328/90-06 as
                        327,328/86-68-17,- referred to -in IR 327,328/86-68 as U-2.4-4,
                        Question on Concrete Expansion Anchor-Sizes.
                                                                                                    '
                                                                            WP 11400 (ECN L6322)
                        modified or added-16 supports to seismically qualify ERCW piping in        .
                                                                                                    '
                        the Diesel Generator Building. Eight supports were inspected and no
                        hardware discrepancies .were observed. The expension anchor test
                        report for support 17A586-2-15 shows 1/2 inch diameter anchors were
                        tested and specified the appropriate test load for 1/2 inch anchors      '
                        as required. However, 3/4 inch diameter anchors were installed and
                        the indicated actual test load was for 3/4 inch anchors.                  .,
                        The inspector reviewed procedure G-32. Bolt Anchors Set In Hardened
                        Concrete.    Section -3.8 states that, unless specifically prohibited,
r                        anchor substitution may be made if the load capacity of the
'
                        substitute anchor equals or exceeds the load capacity of the called
                        for anchor in both tension loading alone and shear loading alone.
                        TVA determined that these criteria were met in the above referenced
                        substitution, and that the testing met the requirements of M&Al-10,
                        Installation, Testing and Documentation of Anchors Set in Hardened
                        Concrete.    The ' inspector had no further questions.      U-2.4-4
                        (327,328/86-68-17) is considered closed.
                  The items remaining open and requiring closure before URI 327,328/90-06-05-
                  can be closed are as follows:
                    Item            86-68 #        Description and Resolution Required
                  86-68-05          D-2.4-2        Improperly documented work. This
                                                    item was denied by TVA, and the NRC denied      1
                                                    this denial in.IR 327,328/87-40. Resolution
                                                    is still pending.-
                    86-68-05        D-2.4-15      Missing anchor bolt pull test
                                                    records.    Item- 2 of this deficiency is
                                                    pending inspection of new installation
                                                    inside containment.    TVA denied item 1 of
                                                    this deficiency, and the NRC denied this
                                                    denial in IR 327,328/87-40.    Resolution is  E
                                                    still pending.                                      l
                    86-68-11        U-2.1-5        Missing snubber test and procurement
                                                    documentation.      TVA  indicated that
                                                    corrective actions are comple te.      The
                                                    inspector requested additional information
                                                    for closure during this inspection period.
                    86-68-13        U-2.3-1        Failure to include vendor instructions in
                                                    work procedures.    This item remains open
                                                    pending further TVA corrective action'.
 
                              -.
                                                                                        -                -
    .-
            .
                  a,;    ,
                                                                                                              ,
                  *
              -
                      .
                                                                                                              ,
                  m,                                                                                          4
                                                                13-                                          .
                            .
  :n
.                                                                                                    .
                                                                                                              .
                          86-68-18        U-2.5-1          Effect of loose controi room electrical          .
                                                            panel. doors on seismic ' qualification.  TVA
                                                            is currently performing a calculation
                ,
C'
                                                            related to this condition.      This item-    -;
                                                            remains open 'pending NRC review of the-
                                                            calculation.-
w
"
                          The additional items addressed above from item 327,328/90-06-05 are still
                          under NRC review. Therefore, this item remains open.
                                                                                                              ,
                          (Closed) NCV 327,328/89-18-05, Failure to Supply Licensed Operators With
                          Updated TS Within the Appropriate Time Period
                          The licensee revised Al-30, Nuclear Plant Conduct of Operation, to address        t
                          the implementation of TS changes including emergency and exigent changes..
                                -
                          These corrective actions are adequate. This NOV is closed.
                      8.  Licensee Event Report Followup (92700)                                            i
                          The following LERs wereLreviewed and closed. The inspector verified that:
                          reporting requirements had been met, causes had been identified,
                          corrective actions appeared appropriate, generic applicability had been            .
                          considered, the LER forms were completed, no unreviewed safety questions
    ,
                          were involved, and violations of regulations or Technical Specification          '
                          conditions had been identified.
                          UNIT 1
                          327/89-006      Auxiliary Building Fire Door Was Breached Without
                                          Appropriate Compensatory Measures in Place Due To
          i
                                          Inadequate Training                                                .
                          327/89-020      Failure To Perform A Proper Monthly Check For the
                                          Reactor Vessel Level Instrumentation ' System. Upper-range
                                            Indicators
                          327/89-035      Sequoyah Unit 1 Reactor Trip Because of ~a High-high.
        .
                                          Steam Generator Level
                          327/90-002      Main Control Room (MCR) Isolation and LC0 3.0.3
                                          Entry Because Both MCR Air Intake Radiation Monitors Were
                                            Inoperable As a Result of Accidental Bumping of Circuit
                                          Breaker.
                          327/90-003      Two Inadvertent Containment Ventilation Isolations
                                          Caused From Power Supply Failures of the Upper Containment
                                          and Containment Purge Radiation Monitors
 
                                                                            .
      -    +  ,
                                        '
                                                                                                  ;
            *
      .-        ,                                                                                ,
    .-  ,'-                                                                                    ,
e                                                        14-
                            *
                                                                                                  ,
                9. ;EventFollow-up-(93702)
  ,
                    a.  On June 12 during power escalation following the refueling outage on    '
                                                                                                  .
                        Sequoyah Unit 1 an anomaly was not:J in the data taken for the-
                                                                          -
                        primary / secondary calorimetric.    The. observed- core differential
                        temperature, delta T, was highcr than expected. This increased delta    ,
                        .T resulted in a calculated value of less than the TS allowable for      J
                        RCS flow when the delta T Ws used to calculate flow. The licensee
                        observed this anomalous high delta T at each increased power level..    .
                        At 90% - power (based on toe- Hghest indicating delta T) the power
                        ascension was stopped.    Calorimetric data indicated that true core
                        power was substantially less, approximately 84%.-~Further analysis of
                        the apparent' discrepancy resulted in a conclusion that the RCS flow
                        had not changed from previous cycles, was in fact well above the TS
                        value, and the errors introduced in the flow / power calculations were
                        caused by the hot leg RTDs reading higher than actual core outlet          ,
                        temperatures. The hot leg RTDs were checked using bridge networks to
                        verify that the 'new signal processing system, Eagle 21, was not
                        causing - the error. . The RTDs were determined to be reading
                        accurately, and the Eagle 21 system was translating that information
                        correctly.
                        The . licensee concluded that the RTDs were not sampling a truly
                        representative version of the RCS flow in the hot legs. Although the
                        RTDs are arranged to sample the RCS water temperature at three points
                        around the' circumference of the pipe, TVA has concluded that the flow
                        in: the hot legs at the RTD locations' is experiencing a streaming
                        phenomenon. . The three RTDs, located at the 0,120, and- 240 degree
                        -locations around.the hot leg pipe, are sampling the water in a manner
                        that results in higher than actual temperatures.      This is not
                        occurring at the cold leg RTDs because of the mixing effect of the
                        RCS pumps prior to the cold leg RTDs. _The licensee removed the RTD      ;
                                                                                                  '
                        bypass manifolds for Unit 1 during the refueling outage. .This
                        replaced the flow nozzles in the - hot leg- pipe with RTDs in
                          thermowells.    The flow nozzles _had multiple inlet holes which
                        probably resulted in a better bulk temperature reading than the
                        single point RTDs.
                        After discussing _the issue with NRC on June 13 and June 14, power was-
                        escalated to 98% (U1118), with the highest delta T indication
                                    -
                        calculated to be 104%.    Further increases in power would challenge
,                        the turbine runback circuit, which occurs at approximately 106%
                        delta T on two of four (2/4) delta T channels. Correction of the
                        delta T channels was accomplished by rescaling the delta T channels
                          to reflect the 100% delta T value for the 63 degree delta T condition
                          versus the 60 degree condition. This was accomplished using TI-2.1,
                          Calorimetric Calculation for Unit 1, on June 14, 1990.
                          The technical resolution of this issue was reviewed and followed
                          closely by the NRC and problems with that resolution were not
                          identified. However, the administrative processes designed for such
                          issue resolution were not taken by TVA until NRC requested the formal
 
r                                                                                                ,
;    w.'
,'
    *
            ; - *
      =.-      .,
            . .                                                                                ,
      .
          ,
1                                                    15
                          .
                      documentation of the evaluations, corrective actions, and problem
                                                                                                '
                      identification forms supporting this issue's resolution. A CAQR was
                      written detailing'the anomalous delta T condition after the condition
                      had already been corrected by rescaling of the delta T circuit for-
  >
                      the ' new delta T. TVA managers stated that a CAQR does - not
                      necessarily have to be initiated prior to resolution of_ the problem,    ;
                      but did concede that the issue should have been reported using .the
                      form for reporting problems, i.e., Problem Evaluation Report (PER),  9
                      as specified in Site Standard Practice SSP 3.2. TVA standard
                                                                                            ~
                      STD 3.1.1, Corrective Actions, stipulates that corrective actions are
                      to ~ be determined and documented in the CAQR within 30 days of        ,
                      discovery. Resolution of this issue is an example of the use of
                                                                                                ~
                      administrative programs merely to document the identification and        ;
                      resolution of problems, and not as .the process used to initiate and
                      control corrective actions after discovery.    Although no problems
                      were-identified' concerning the technical decisions in this instance,
                      identification of the problem and its resolution were not
                      accomplished using the corrective action program. Rather, the
                      CAQR/PRD process was employed as an afterthought, primarily to
                      document the corrective actions arrived at outside the corrective
                      action program.. Similar instances of a reluctance to initiate the
                      administrative programs have occurred previously and includc such
                      issues as the failure to document the RHR pump-to-pump interaction
                      problem on a CAQR, and repeated instances of failure to initiate
                      Safety Evaluations-for identified deficiencies. Although violations
                      of NRC requirements are not identified in this instance, lack of a
                      commitment to aggressively use the proper administrative programs is.
                      considered a weakness,
                    b. On June 28, 1990, at 10:35 p.m., a Unit i upper containment purge was
                        initiated in accordance with S01-30.2, Containment Purge System
                      Operation. At 7:20 on June 29, the oncoming operating shift crew        ,
                      discovered that the purge was being ' performed without the
                                                                '
                      Environmental- ' Allowance Monitor -(EAM) disabled. as required by
                        Instruction Change Form (ICF) 90-0326 to the procedure. The ICF
                      requires that the EAM is to be disabled during. purge operations to
                      change the SG level trip setpoint from the normal 13 percent to a
                      more conservative'19 percent setpoint. The EAM automatically resets
                        this setpoint during accident conditions due to harsh environment
                      conditions inside containment that could cause the . indicated SG
                        levels to read non-conservatively.      Analysis by Westinghouse
                      determined that the automatic reset function could be delayed for
                      certain accidents including a feedline break while the purge system
                        is operating. Therefore, the more conservative-setting of 19 percent
                        should-be in effect while the purge system is operating. The change
                        to the purge system procedure was in the form of a temporary ICF
                      which had not been interfiled in the procedure.
                      When the operators determined that the EAM had not been disabled, all
                        SG level trip channels were declared inoperable until the EAM was
                        disabled and TS 3.0.3 was entered. After disebling the EAM, TS 3.0.3
                      was exited at 8:16 a.m. on June 29. A DNE evaluation was initiated
                                                                                            1
 
                                                                                                        '
  p
                                                                                                        ;
      "-
  ;                .              ,
        ...
                -
                    ;,.
              . . . .
        ,
            ,
                                                                                                        ;
                                                            16-
  "                      .    .
                                                                                                        -
                                                                                                        i
    '
                            and the assessment concluded that during the time the level trip          i
                            channels .were at the non-conservative setpoint of 13 percent cith
                            purge in effect Unit I was in an unanalyzed condition.          This
                            determination was made at 1:10 p.m. on June 29, 1990, and 1-hour
                            notification made to the NRC at 1:40 p.m.
                            The : licensee's investigation determined that the operator did not
                            check the front of the controlled copy of the procedure to ascertain        ,
      g                    whether a temporary ICF was attached.- The normal method of -              i
                            . conducting the purge requires the operators to obtain a controlled    .;
                            copy of the procedure from the Technical Information Center (TIC).
                            which has copies of all ICFs interfiled in the procedure. In this      7
                            instance, a controlled copy from the control room was used which does  ]"
                            not routinely interfile temporary ICFs, but simply attaches the 'ICF
                            to the front of the procedure.
                            Corrective actions included making the ICF permanent and interfiling
                            the ICF in the procedure.      The licensee is investigating the
                            possibility of a design change to disable the EAM at. all power
                            conditions above the point at which 'the reset function is needed,.
                            that is,'during plant startup while the SG level _ control system is in    .
                            manual' control and level transients routinely occur. Once the level      !
                            control system is in automatic, there is no real need to lower the
                            trip setpoints to-the 13 percent'value.
                            Response to this event and the corrective actions taken were timely
                            and appropriate. .- This licensee identified violation is not being
                            cited because criteria specified in Section V.G.1 of the NRC
                            Enforcement Policy were satisfied.    .This item is designated as NCV    !
                            327',328/90-22-02.
                        c.  During surveillance testing of control board voltmeters on June 25,
                            1990, an improper test sequence resulted in blown fuses in the .          '
                            shutdown board logic relays which deenergized the IA-A 6.9 kv board
                            and startup of_the emergency diesel generators.
                            The event occurred during performance of procedure DPS0-SMI-4C,
                            Indicating Voltmeter Calibrations for 6900' Volt Shutdown Boards. As
                            required by the' procedure, 'the bus voltmeter. selector switch was-
                            placed in the OFF position to electrically isolate the voltmeter for
                            testing and calibration.    A voltage source with a test meter- was
                            applied to check the installed voltmeter. - After completing the.
,
                            calibration, the test requires a check of the voltmeter for upscale
L                            readings: on all three phases by placing the installed voltmeter back
p                            in service and selecting each of the three phases in sequence. The
'
                            test director did not remove the test equipment from the circuit, but
                            instead only turned off the test meter used with the test equipment.
                            The installed voltmeter was returned to service with the test
I
                            equipment still connected and applying a test signal. This tied the
                            test voltage to the installed voltmeter circuit and overloaded the
                            entire bus potential transformer circuit because a short circuit was
l :.
                                                                                                        1
 
,
                                                                                                      ;
              ,                                                                                      ,
      .:. :      _,
  ,
      .>    .,3-
                                                            17
                        .      .
                                                                                                      >
                                                        .
                              induced around the meter. This blew two fuses which protect the
                                                                                                      '
                              voltmeter and' the IA-A logic panel relays', 'deenergizing the relays.
                            ; Deenergizing the -logic relays causes. the bus to- trip free from the
                              normal supply and actuate the blackout relays and start the F.DGs.
                                                                                                      '
                              All required loads were then ' sequenced on after the 1A-A EDG          :
                              reenergized the board.
                                                                                                      '
                              After finding the cause of the-event, the operators replaced the
                              blown fuses, paralleled the board with the normal supply, then          ,
    ,                      ' stopped the EDGs'.    The event was reported and investigated as.
                              required. The cause of the event was determined to be a deficient
                              procedure used in the surveillance, DPS0-SMI-4C. The procedure did
                              not specify when the test equipment was to be removed, and lift the-
                              removal up to the performer. There was also no precaution statement
                              to alert the user to the possibility of causing a shutdown board
                              blackout.
                              Corrective actions included revising the procedure, reviewing similar  i
                              procedures for deficiencies of this type, briefing of test personnel    ,
                              on.the event, and submitting a Design Change Request for considering  ,
                              a .eparate fuse indication circuit.
                              This ' violation is not' being cited because criteria . specified in
                              Section V.A of the NRC Enforcement Policy were satisfied. This item    a
                              will be tracked as NCV 327,328/90-22-_03.
                10.    Electrical Cable Issues (37700);
                        During- a review of the electrical calculations which support the
                        licensee's March 28, 1990 submittal pertaining to their cable test
                    -program, it was revealed by the licensee that calculation SQN-CSS-009 for
                        selecting the worst case 1E conduits relative to pullbys, jaming, and '
                        lack of support of vertical cable runs .was never issued by TVA. The
                    _ inspector verified that'the review and approval-blocks on the calculation'
                      _
                        were not signed. The licensee stated in Enclosure 2 of the July 31, 1987
                        submittal titled, " Revised Cable Test Program," that the implementation of
                        the cable selection criteria would be in a calculation. The licensee            1
                        stated to the inspector that the unapproved and unissued calculation            l
                        SQN-CSS-009 was the document that implemented that statement.
                        The _ licensee told the inspector that they were aware that calculation
                        SQN-CSS-009 was not approved as early as March of 1988. This calculation        1
  .                    implemented the conduit selection criteria for 1E cables to select the          l
                        worst case conduits for high-potential testing. The test results were the        i
                        basis,for NRC and TVA concluding prior to restart in 1988 that 1E cables
                        were properly installed in the plant and that_ the associated                    l
                        safety-related equipment would perform its intended function. Nuclear
                        Power Standard 3.1.1, Appendix B, states that one of the criteria for
                        preparing a CAQR is failure to comply with licensing commitments to NRC
                        where it is apparent the discrepancy directly affects quality. Since the
                        adequacy of this calcuiation was part of the basis for justifying
                                                                                                        l
                                                                                                        l
 
                                                                                                i
t'
        1.  .g
  e:    y ,_
,      A
      ,
                                                      18
                                                                                                ,
                  equipment operability, and since tne calculation was not reviewed and
                  approved by the licensee, the basis for concluding that the licensee
                . tested the '" worst case" conduits is in question. The inspector concluded    ,
                  that this issue affected quality.      The licensee stated that they do not    i
                  know why the calculation was not approved-and. issued.    The inspector was
                  told by the licensee that a CAQR had not been issued ~ the failure to
                  approve and issue this calculation. This issue ~ is designated as URI-
                  327,328/90-22-04, Unisssued Calculation for IE Cable Testing, pending
                  further review of why this calculation was not issued.
                  During the review of Calculation SQN-CSS-031,- Electrical Cable
                  Installation Pull-by Calculations Comparison, the inspector-noted that the    ,
                  conclusion was depeMent on an assumption which was not stated in the          i
                  Calculation. The assumption was that the cable reviewed using the
                  criteria developed from Watts Bar plus the cables tested from the original
                  criteria were representative of the population of IE caWas from
                  SQN-CSS-005.    The inspector did not find this assumption 'tated in the
                  calculation as required by NEP 3.1, Calculations. Attachment 4 to_NEP.3.1
                  states that assumptions shall be adequately identified and documented.
                  This . item is designated as URI- 327,328/90-22-05, Assumption Not Identified
                  in Calculation, pending further review of why this assumption was not
                  stated.                                                                        l
            11. - Exit Interview (30703)                                                        l
                -- The inspection scope and findings were summarized on July 3,1990, with
                  those persons indicated in paragraph 1. The Senior Resident Inspector.        i
                -described the areas inspected and discussed in detail the inspection
                  findings listed below. A weakness was also identified in the use of            l
                  administrative-control programs to correct conditions adverse to quality.
                  The licensee acknowledged the inspection findings and did not-identify as
                  proprietary any of the material reviewed by the inspectors during the
                  inspection.
                  Inspection Findings:
                  One violation was identified which involved a failure to control overtime
                  for plant personnel.
                        327,328/90-22-01, Exceeding of Overtime Limits Without Proper
                        Approval - paragraph 7.
                                                                                                1
                  Two unresolved items were identified pertaining to cable testing issues.
                        327,328/90-22-04, Unissued Calculation for 1E Cable Testing -
                        paragraph 10.
                        327,328/90-22-05, Assumption Not Identified in Calculation      -
                        paragraph 10.
 
c
                                                                                                      '
          ; :; c                              i
                  .
          : -:
                    .#
    . :,:          g.
          ..
              m . --                                                                                    ]
                                                                19
                            .      .
                            One non-cited violation was identified , involving a loss of a shutdown        '
                            board'during surveillance testing.
                                327,328/90-22-03, Partial Blackout Caused By Energizing Circuit With
                                Testing Rig Still Connected - paragraph 9.c.
                            One licensee-identified violation was identified involving an inadvertent-
)                          entry into TS 3.0.3 and placing the plant in an unanalyzed condition          .
                          'during containment purge operations.
                                                                                                            1
                                  327,328/90-22-02, Unit 1 'in Unanalyzed Condition During Containment
                                  Purge Operations - paragraph 9.b.
                            No deviations or inspector follow-up items were identified.
                            During the reporting period, frequent discussions were held with the
                            Interim Site Director, Plant Manager and other managers concerning
                            inspection findings.
                      12. List of Acronyms and Initialisms
                            ABGTS-    Auxiliary Building Gas Treatment System
                            ABI' -    Auxiliary Building Isolation
                            ABSCE-    Auxiliary Building Secondary Containment Enclosure
                            AFW -      Auxiliary Feedwater
                            AI    -
                                      Administrative Instruction                                        '
  .
                            A01  -    Abnormal Operating Instruction
                            AVO  -    Auxiliary Unit Operator
                          'ASOS -    Assistant Shift Operating Supervisor
                            ASTM -    American Society of Testing and Materials
                            BIT  -    Boron Injection Tank                                                .
                            BFN  -
                                      Browns Ferry Nuclear Plant
                            C&A  --  Control and Auxiliary Buildings
                            CAQR -    Conditions Adverse to Quality Report
                            CCS -      Component Cooling Water System
                            CCP -      Centrifugal Charging Pump
                            CCTS -    Corporate Commitment Tracking System
                            CFR  -    Code of Federal Regulations-
                            COPS -    Cold Overpressure Protection System
                            CS    -
                                      Containment Spray                                                  4
                            CSSC -    Critical Structures, Systems and Components
                            CVCS -    Chemical and Volume Control System
                            CVI- -    Containment Ventilation Isolation
                            DC-  -
                                      Direct Current
                            DCN  -
                                      Design Change Notice
                            DG    -
                                      Diesel Generator                                                    i
                            DNE  -    Division of Nuclear Engineering
                            ECN  -    Engineering Change Notice
                            ECCS -    Emergency Core Ccoling System
                            EDG    -
                                      Emergency Diesel Generator
 
r,
L-            ; ,
q
              "<
                      y    ,7
g                ,; m ,. ,
                    .;n
          .
                                                                      20                1
  <'
                                  .      ..
                                                                                          I
                                El    -      Emergency Instructions
                              ENS    -      Emergency Notification System
                                E0P' -        Emergency Operating Procedure                i
                                EO    -      Emergency Operating Instruction              '
                                ERCW -        Essential Raw Cooling Water                  '
                                ESF    -      Engineered-Safety Feature
                                FCV    -
                                              Flow Control Valve
        ''
            .
                                FSAR -        Final Safety Analysis Report
                                GDC .-        General Design Criteria
                                G01    -    General Operating Instruction
                                GL      -
                                              Generic Letter
        U
                                HVAC -        Heating Ventilation and Air Conditioning-
              '
                                Hic -        Hand-operated Indicating Controller
      ,                        H0      -    Hold Order
                                HP      -    Health Physics
'
                                ICF    -    Instruction Change Form
                                IDI    -    Independent Design Inspection
                              'IN      -    NRC Information Notic.e
                                IFI    -    Inspector Followup Item                        j
                '
                                IM      -
                                              Instrument Maintenance                          !
                                IMI    -
                                              Instrument Maintenance Instruction              I
                                1R      -
                                              Inspection Report
  ..
                                KVA    -    Kilovolt-Amp                                .!
f;                              KW      -
                                              Kilowatt                                        ,
                                KV      -
                                              Kilovolt                                        i
                                LER    -    Licensee Event Report                            i
                                LC0 -        Limiting Condition for Operation            j
                                LIV      -
                                              Licensee Identified Violation-                  =
                                LOCA -        Loss of Coolant Accident
                                MCR= -        Main Control Room                                I
                              : MI      -
                                              Maintenance Instruction                          i
                                MR    --      Maintenance Report                              I
                              -MSIV -        Main. Steam Isolation Valve
                                  NB-    -
                                              NRC Bulletin
                                  NOV -        Notice of Violation
                                  NQAM --      Nuclear Quality Assurance Manual                  i
                                  NRC -        Nuclear Regulatory Commission
        '
                                  OSLA -      Operations-Section Letter - Administrative
                                  OSLT -      Operations Section Letter - Training
                                                                                            4
                                :0SP -        Office of Special Projects
                                  PLS          Precautions, Limitations, and Setpoints
                                  PM      -
                                              Preventive Maintenance                              l
                                  PPM' -      Parts Per Million                                  l
                                  PMT-    -
                                              Post Modification Test
                                  PORC -      Plant Operations Review Conmittee                  4
                                  P0RS -      Plant Operation Review Staff
                                  PRD      -
                                              Problem Reporting Document
                                  PRO      -
                                              Patentially Reportable Occurrence
                                  QA      -  41ality Assurance
                                  QC        -
                                              Quality Control
                                                                                                    I
                                                                                                    l
                                                                                                    l
 
Q/o ,
                                                                                  _
y t p o;
'              *
        *  '!.
      .
          ,
            o.
                                                        21
                    ..        .
  t
                  RCA    -
                                Radiation Control Area
                  RCDT -        Reactor Coolant Drain Tank
                  PCP    -      Reactor Coolant Pump
                  RCS    -      Reactor Coolant System
                  RG    -      Regulatory Guide
                  RHR    -      Residual Heat Removal
                  RM    -      Radiation Monitor
                  R0    -      Reactor Operator
.'
                  RPI    -
                                Rod Position Indication
                  RPM    -    Revolutions Per Minute
                  RTD    -    Resistivity Temperature Device Detector
                  RWP -        Radiation Work Permit
                  RWST -        Refueling Water Storage Tank
                  SER    -    Safety Evaluation Report
                  SG      -    Steam Generator
                  SI      -    Surveillance Instruction
    ,
                  SMI    -    Special Maintenance Instruction
                  501    -    System Operating Instructions
                  SOS    -    Shift 0)erating Supervisor
                  SQM    -    Sequoyal Standard Practice Maintenance
                  SQRT -      Seismic Qualification Review Team
                  SR      -    Nrvallance Requirements
                  590    -    Senior Reactor Operator
                  S!0MI-      Safety Systems Outage Modification Inspection
                  SSIE -      Safety System Quality Evaluation
                  SSIS -      Solid State Protection System
                  STA      -    Shift Technical Advisor
                    ST1 -        Special Test Instruction
                    TACF -      Temporary Alteration Control Form
                    TAVE -      Average Reactor Coolant Temperature
                    TDAFW- --    Turbine Driven Auxiliary Feedwater
                    T1      -    Technical Instruction
                    TREF -      Reference Temperature
                    TROI -      Tracking Open items
                    TS      -  Technical Specifications
                    TVA      -  Tennessee Valley Authority
                    UHI      -  Upper Head Injection
                    UO      -  Unit Operator                                        !
                    URI      -  Unresolved Item                                      ,
                    USQD -      Unreviewed Safety Question Determination            i
                    VDC      -
                                Volts Direct Current                                i
                    VAC      -  Volts Alternating Current                            '
                    WCG      -  Work Control Group
                    WP        -  Work Plan                                            .L
                  .WR        -  Work Request
                                                                                    -  .
                                                                              \
                                                                              ~
                                                                                g
                      .
}}

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