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{{Adams
#REDIRECT [[IR 05000482/1989005]]
| number = ML20245J425
| issue date = 04/21/1989
| title = Insp Rept 50-482/89-05 on 890201-28.Violations Noted.Major Areas Inspected:Plant Status,Followup on Previously Identified NRC Items,Operational Safety Verification,Monthly Surveillance Observation & Monthly Maint Observation
| author name = Bartlett B, Holler E, Skow M
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
| addressee name =
| addressee affiliation =
| docket = 05000482
| license number =
| contact person =
| document report number = 50-482-89-05, 50-482-89-5, NUDOCS 8905040126
| package number = ML20245J388
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| page count = 18
}}
See also: [[see also::IR 05000482/1989005]]
 
=Text=
{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
  ,                                                                                                                                            .
                .
                    <    .
                                                                                                                                                      ,
                                                                                                                                                      !
                            -
                      .                                                                                                                                i
                                                              APPENDIX B                                                                            f
                                                U.S. NUCLEAR REGULATORY COMMISSION
                                                                REGION IV
                                                                                                                                                    i
                        NRC Inspection Report:    50-482/89-05'        Operating License:  NPF-42                                                f
                        Docket:  50-482
                                                                                                                                                    :
                        Licensee: Wolf Creek Nuclear Operating Corporation (WCN00)                                                                  .
                                      P.O. Box 411                                                                                              ')
                                      Burlington, Kansas 66839                                                                                    ''
                        Facility Name:    Wolf Creek Generating Station (WCGS)
                        Inspection At: WCGS, Coffey County, Burlington, Kansas
,
                        Inspection Conducted:    February 1-28, 1989
r
          s            . Inspectors:                        !h
                                        B. L. Bartlett, Sen %r Resident Inspector
                                                                                              #bt / 99
                                                                                            Dat'e
                                                                                                    '
                                                                                                                                            -,
                                        Project Section D, Division of Reactor
                                                                                                                                        '
                                                                                                                                        .
            s                            Projects
                                                                                              Mf91ft9
                                        M. E. Skow, Residght Inspector, Project              Date
                                                                                                    '
                *i.
              *
                                          Section D, Division of Reactor Projects
                                                                                                                                          ,
                        Approved:                                -
                                                                                              N/ CPI!ff
                                        E. J.UHoller, Chief, Project Section D              Date
                                        Division of Reactor Projects
                                      8905040126 890501
                                      PDR    ADOCK 05000482
                                      Q                    PDC
. _ _ - _ -
_
 
      - _ _ _ _            _ _ _ _ - _ - _ - - _ _ _ _ _ - - _ _ -    -_ _ _ _ _ _ _ . _
                        -
!    .
                          -
!      ,
                      ,
            '
                    ,
                                                                                          -2-
    ,
r .
                ' Inspection Summary
                  Inspection Conducted February 1-28, 1989 (Report 50-482/89-05)
g
                Areas Inspected: Routine, unannounced inspection including plant status,
                  followup on previously identified NRC items, operational safety verification,
                monthly surveillance observation, monthly maintenance observation, review of
                  licensee event reports, onsite followup of events at operating power reactors,
                  and installation and testirg of modifications.
                  Results: Within the arcas' inspected, three apparent violations, one unresolved.
                  Item, and one open item were identified. The-violations dealt directly or
                  indirectly with the auxiliary feedwater system. One violation resulted from
                  not lockwiring a valve in the neutral position (paragraph 4.b), another
                  violation resulted from an inoperable fire barrier through one of the walls
                  enclosing Auxiliary Feedwater Pump "A" (paragraph 4.a), and the third violation
                  resulted from a failure to have control room drawings reflect as-built plant.
                  equipment (paragraph 9.a).                        The unresolved item concerned engineering followup
                  to an air bubble.in the auxiliary feedwater pump suction (para                                    The
                open item concerned updating the Safety Analysis Report                                  (USAR)
                                                                                                      to reflect  sitegraph 5.b).
                  conditions (paragraph 9.b).
                This . inspection identified an < example.of a: 2-year old minor modification made -
                  to a safety-system which was not reflected in permanent plant drawings.- In
                October 1988, the quality assurance (QA) organization identified similar
                . problems with " red 11ning" (color coding) control room drawings. _ The licensee-                                i
                  appears to be implementing corrective actions to the findings identified in the
                QA audit; however the licensee does not appear to be giving adequate resources
                  to this corrective action.                        In addition, the inspector understands that it
                                                                                            .
                                                                                                                                  i
                  could be 31/2 years before the drawings are permanently changed. This practice
                  is discussed in paragraph 9.
                                                                                                                                  i
 
    __-                          . _ -                  -
                    -
        ,. ,
          ,Y      ,
      ,
                    *
          x
                                                                    _
                                                              . DETAILS
                                                          <
                .
,                                        . .
              '1.. ' Persons Contacted
                        ' Principal Licensee Personnel
                      1*R.,M. Grant,'Vice' President QA-
                        *J. A. BaileybVice President, Operations
                                    .
                        *G. D. Boyer, Plant Manager
                                              .
                        *R. W. Holloway, Manager, Maintenance and Modifications
                        *0. L. Maynard. . Manager, Licensing
                        *B. McKinney, Manager,~0perations
                        *M. G. Williams, Manager, Plant-Support
                        *C. E. Parry, Manager, .QA, WCGS    .
                        *A. A. Freitag,' Manager,. Nuclear Plant Engineering-(NPE), WCGS
                                        .
                                                                                                                !
                        *W. M. Lindsay, Supervisor, QA                                                          !
                        *C. J. Hoch, QA Technologist.-                                                          .
                                                                                                                i
        M              *JJ Pippin, Manager, NPE
                        *S. Wideman, Licensing Specialist III
                        *C. W. r fler,LManager, I&C                      _
                        *T. L. . , ester, Manager, Facilities & Modification                                    1
                                                                                                                '
                        *R. S. Benedict,-Manager, Quality Control (QC)
                        *R. Flannigan, Manager,LNuclear Safety Engineering                                    q -
                      . *J.  L' Houghton,
                                .-              Operations Supervisor
                      .*J. A. Weeks,-Shift Supervisor
                        *C. Sprout, Section Manager, NPE Systems
                              .
                        *S. F. Hatch, Supervisor, Quality Systeins
                                                                                                                ;
                                                                                                                '
                          The NRC inspectors also contacted other members of the licensee's staff
  ,
                          during the. inspection period to discuss identified issues.
                                                                                                                !
                        * Denotes those personnel in attendance at the exit meeting held on
                          March 7, 1989.
              2.          Plant-Status                                                                        ,
                          The plant operated in Mode 1 (100 percent power) during this inspection
                          period. On February 2, 1989, the plant tripped from 100 percent power.
                          The cause'of the trip and the licensee's followup activity is discussed in
                          paragraph 8.        The licensee returned the plant to 100 percent power on        l
                          February 5,1989.                                                                    l
              3.        . Followup on Previously Identified NRC Items (92702)-
                            (0 pen) Violation (482/88200-01): Failure to Take Adequate Corrective              )
                                                                                                                l
                          Actions - Part three of this item concerned the chlorine monitors in the
,
l'
                          control room ventilation system. The monitors have been replaced and this
                          part of the violation is resolved. The overall violation remains open                ;
                          until all parts of the violation can be closed.
L
o
                                                                                                            -
                                                                                            - - - - _ _ _ _
 
l                                    -              .:
                                      '
            ,
                  v. ,
              ,
                  ,
                                                                      u
                                                        .
            ,                ,
                                                                          -4-
                                              '
                .
                  4.              .0perational Safety Verification (71707)
                                    The purpose of.this inspection area was to ensure that the facility was
                                    being operated safely and in conformance with license and regulatory
                                    requirements. It also was to ensure that the licensee's management
                                    control system was effectively discharging its responsibilities for
                                    continued safe operation.      The methods used to perform this inspection
                                    area included direct observation of activities and equipment, tours of the
                                    facility, interviews and discussions with licensee personnel, independent
                                    verification of safety system status and limiting conditions for
                                    operation, corrective actions, and review of facility records.
                                    Areas reviewed during this inspection included, but were not limited to,
                                    control room activities, routine surveillance, engineered safety feature
                                    operability, radiation protection controls, fire protection, security,
                                    plant cleanliness, instrumentation and alarms, deficiency reports, and
                                    corrective actions.
                                    Routine surveillance and operating activities witnessed and/or reviewed by
                                    the NRC inspectors are listed below:
                                    a.    On February 25, 1989, during a routine tour of the auxiliary            !
                                          feedwater (AFW) pump rooms, the NRC inspector observed an unsealed
                                          3/4-inch penetration.      The penetration was through the wall
                                          separating Motor Driven (MDAFW) Pump "A" from a hallway.      The MDAFW
                                          pump "A" is located in Room 1326 (Fire Area A-14) and the hallway is
  ,                                      Room 1329 (Fire Area A-33).      Upon notifying the control room, the
                                          shift supervisor initiated Work Request (WR) 01066-89 and dispatched
                                          an operator to temporarily plug both ends of the penetration.
                                          Discussions-with licensee personnel and review of fire hazards
                                          analysis showed that if a fire had migrated between Rooms 1326 and
                                          1329, through this inoperable barrier, the safe shutdown of the unit
                                          would not have been affected.
                                          At the conclusion of the inspection period, the licensee had not
                                          determined the origin of the inoperable fire barrier, but had assumed
                                          the penetration was abandoned during construction. Failure to seal
                                          the abandoned penetration or prepare a fire protector impairment
                                          control permit form in accordance with plant procedures is an
                                          apparent violation (482/8905-01) of TS 6.8.1.
                                    b.    On February 8, 1989, during a routine tour of the AFW pump and valve
                                          rooms, the NRC inspectors determined that Valve AL HV-12 (TDAFW pump
                                          discharge to steam generator (S/G) "C") was not lockwired in neutral
                                          as required by Procedure CKL AL-120, Revision 11, " Auxiliary
                                          Feedwater Normal Lineup," and Drawing M-13ALO5(Q), Revision 2,
                                          " Piping Isometric Auxiliary Feedwater Pumps Recirculation Piping."
                                          The NRC inspector also found Valve AL V037 (MDAFW Pump "B" discharge
                                          to S/G "D") locked as required by procedure, but the lock could be
    - _ _- -        _ _ _ _ _ _ .                                                                                l
 
- _ _ - _ - _ _              -    .          _                _
                .
                *    2
                                                            -5-
                              easily removed. When the shift supervisor was notified, operators
                              were dispatched and corrected the identified problems. Failure to
                              lock Valve AL HV-12 as required by licensee procedure is a violation
                              (482/8905-02).
                        c.    On February 15, 1989, during a routine tour of the auxiliary
                              building, the NRC inspectors identified Valve EG LV-2 (component
                              cooling water surge Tank "B" fill) as not being properly lockwired in
                              neutral. The valve did have a lockwire, but it did not prevent the
                              locking yoke frem being moved. This valve type is the same as
                              identified in Violation 482/8905-02.
                              NRC inspectors have identified a previous example of the discharge
                              valves to the TDAFW pumps not being properly lockwired. This was
                              documented in Violation 482/8618-02.
                        The three examples of the violation listed in NRC Inspection
                        Report 50-482/86-18 and the examples in paragraphs b and c above indicate
                        a licensee problem in properly locking and maintaining locked valves of
                        this type. All of the valves identified in the above examples were in
                        their required positions.
                  5.  _ Monthly Surveillance Observation (61726)
                        The purpose of this inspection area was to ascertain whether surveillance
                        of safety-significant systems and components were being conducted in
                        accordarice with TS requirements. Methods used to perform this inspection
                        included direct observation of licensee activities and review of records.
                        Items in this inspection area included, but were not limited to,
                        verification that:
                        o      Testing was accomplished by qualified personnel in accordance with an
                              approved test procedure.
                        o      The surveillance procedure was in conformance with TS requirements.
                        o      The operating system and test instrumentation calibration was within
                              its current calibration cycle.
                        o      Required administrative approvals and clearances were obtained prior
                              to initiating the test.
                        o      Limiting conditions for operation were met and the system was
                              properly returned to service.
                        o      The test data was accurate and complete and the test results met TS
                                requirements.
                                                                                                    I
 
                                                                                                    -__ ___ - _ -
                                        P
          kt
      3 ::"  .    .
                      _
                                                                                                <
                                                                                                                  , .
                                                                                                                      j
      g        ,
                        '
                                                              -6-
      ~
                                    x
                                                  4
                        i'
  4
        ,
                          Surveillance witnessed and/or reviewed by the NRC inspectors are listed                    l
      e                  below:                                                                                      ;
  i                      o    STS SE-001, Revision 8, " Power Range Adjustment to Calorimetric,"
    +                        performed February 15, 1989.
                          o    STS 10-255B, devision 5, " Analog Channel Operational Test Control
                              Room Air Intake Radiation Monitor GK RE-04," performed                                l
                              February 15, 1989.
                          o    STS IC-255A, Revis1on 5, " Analog Channel Operational Test Control
                              Room Air Intake Radiation Monitor GK RE-05," performed
                  (            February 15, 1989.
i
                          o-  . STS ' AL-103, Revision 8, " Turbine Driven Auxiliary Feedwater Pump
                                Inservice Pump Test," performed February 28, 1989.
                        .o    STS IC-203, Revision 5, " Analog Channel Operational Test 7300 Process                !
                                Instrumentation Protection Set III (Blue)," performed
                              , Februa ry' 21, 1989.
                          o    STS GG-0018, Revision 7, " Emergency Exhaust Filtration System Train
                                'B' 10 Hour Operability Test," performed February 28, 1989.
                        ' Selected NRC inspector observations are discussed below:
                          a.  The-NRC inspectors reviewed Design / Deficiency Report 89-008 concerning
                              errors in surveillance tests performed on two circuit breakers in
                              1986.    Procedure STS MT-024, " Functional Test of 480, 240, and 120
                              Volt Molded Case Circuit Breakers," contained acceptance criteria
                              errors for~ Breakers NG01BEF3 (ENHV-1, Containment Spray Pump "A"
                              suction isolation valve) and PG2108 (Pressurizer Heater Coils 5,
                              6,and27). TheSurveillanceTechnicalSpecification(STS) testing
                              sequence required a preliminary test to measure the instantaneous-
                              single-phase trip current. If the results of this test were outside
                              the acceptance criteria, the test was to be performed on two phases
                                in series. (The two-phases-in-series test is utilized to conclusively
                              determine operability of the circuit breakers.) The single-phase
                              tests for the two breakers .in question were considered acceptable
                              when, in fact, they were not. Because incorrect acceptance criteria
                              for the single-phase test was used in 1986, the final two-phases-in-
                                series tests for the two breakers were not performed.
L
                              As a collateral issue, the NRC inspectors observed that, in response
I
                                to Violation 482/8632-01, the licensee had implemented a directive                    ,
                                                                                                                      1
l                              for maintenance engineering to review completed maintenance
                                surveillance. This corrective action, which was implemented after
                                the 1986 surveillance discussed above, should catch errors such as
    .
                                using incorrect acceptance criteria. However, the directive
                                                                                                                      l
l
 
                                                                              ____ -___-_ _ _ -
      .
.  .
        '
  .
                                              -7-
            -
              discussing the maintenance engineering review was not circulated to
              personnel outside the maintenance organization. Thus, maintenance
              engineering depends upon the surveillance coordinator to ensure that
              maintenance engineering receives all of the surveillance that they
              are supposed to review.
              In January 1989, the licensee discovered the acceptance criteria
              error regarding the 1986 surveillance tests while entering the data
              into a computer data base to be used for trending breaker
              performance. The licensee decided to repeat the surveillance tests
              for the two breakers. Breaker NG01BEF3 passed the single phase test
              and was declared operable. ANSI /IEEE Standard 338-1977, "IEEE
              Standard Criteria for the Periodic Testing of Nuclear Power
              Generating Station Safety Systems," states that results of a failed
              test cannot be negated by a simple successful repetition. Because of
              the time interval since the failed test, this action may not have
              been inappropriate in this case.
              Breaker PG2108 did not pass the single phase portion of the repeated
              surveillance test. STS MT-024 required the performance of the
              two-phase-in-series test if the one-phase test failed; however,
              this requirement was missed by the test performer who initiated a WR
              to repair the breaker, rather than performing the required
              two-phase-in-series test. Subsequent to the repair, the breaker
              was retested and passed. This apparent violation of the requirements
              of STS MT-024 has not not been cited because it meets the NRC
              Enforcement Policy criteria for exercisising discretion in that it
              was a self-identified, nonwillful, less significant violation for
              which corrective action was taken. The licensee issued a change to
              STS MT-024 to clarify the requirement for its two-phase-in-series
              test.
          b.  During the performance of STS AL-103, the NRC inspector observed that
              one of the suction lines for the TDAFW pump was below room
              temperature. The other suction lines were at room temperature. Some
              flow was observed coming from the high point vent line on the cold
              suction line. The cold pipe was one of the essential service
              water (ESW)supplylinesandwasdownstreamofValveALHV-33. The
              licensee later determined that Valve AL HV-33 was leaking by and
              issued a corrective WR. The licensee explained that after a
              surveillance is performed to stroke test AL HV-33, the pipe from                l
              Valve AL HV-33 to the down stream check valve (AL V015) is drained              l
              and the high point vent valve (AL V139) is left open. Thus, that                l
              section of pipe is left filled with air. This also applies to the
              other ESW line for the TDAFW pump and to the ESW lines to the MDAFW
              pumps.    If the AFW pumps are running and a signal occurs to shift
                suction to ESW from the condensate storage tank (CST), there is a              1
                possibility that the entrained air may bind the pumps or cause a
              waterhamer. This portion of suction pipe was the subject of a                    l
                                      ,
                                                                                                l
                                                                                                l
                                                                                                !
                                                                                                I
 
                                                                                      -------- _ _
  .
        .
    .
                                                -8-
                                                                                                    1
                letter from the constructor to SNUPPS on March 16, 1981. The letter
                proposed to SNUPPS that the section of pipe between the isolation                  ,
                valve and the check valve be left filled with CST water, leakage from              j
                the continuous drain line be monitored, and this leakage be
                periodically sampled. The licensee stated that they were looking                  {
                into this issue and that corrective action was initiated. The
                licensee determined that the other three similar ESW supply lines
                were filled with CST grade water. They also stated that the
                surveillance procedure would be changed to leave the suction lines
                filled with water. Pending review of the licensee's evaluation of                  j
                pump operability, this is considered an unresolved                                j
                item (482/8905-04). The licensee later discovered that S/G water                    i
                chemistry and condenser hotwell chemistry were out of specifications                l
                and trending up. Investigation revealed that Valve AL HV-033 had
                leaked by enough to overcome the continuous vent and flow through the
                AFW recirculating line to the CST. The licensee adjusted
                Valve AL HV-033 to lower its leak rate and commenced cleanup of the
                CST.
      6.  Monthly Maintenance Observation (62703)
          The purpose of this inspection area was to ascertain that maintenance
          activities of safety-related systems and components were conducted in
          accordance with approved procedures and TS. Methods used in this
          inspection area included direct observation, personnel interview, and                    i
          record review.
          Items verified in this inspection area, where appropriate, included:
          o    Activities did not violate limiting conditions for operation and that
                redundant components were operable.
!          o    Required administrative approvals and clearances were obtained before
l                initiating work.
          o    Radiological controls were properly implemented.
l          o    Fire prevention controls were implemented.
          o    Required alignments and surveillance to verify postmaintenance
!                operability were performed.
          o    Replacement parts and materials used were properly certified.
          o    Craftsmen were qualified to accomplish the designated task and
                additional technical expertise was made available when needed.
          o    Quality control hold points and/or checklists were used and quality
                control personnel observed designated work activities.
 
                    _                        . _ .                                      _                . _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
,
                                                                                                (
        i
          l
    -
                            -
                      ,
                          ,
f                  "
L/ "                            ,
f;              ,
                                                                        -9-                                                                f
                                                                                                                                                  6
                                  o      Procedures used were adequate, approved, and up'to date.
4            ,x ,                Portions of the selected maintenance activities were observed on the WRs
                  -
                                  listed below and related documents were reviewed by the NRC inspectors:
                  /                    No.                              Activity
        a
                                  WR 50246-89                        Monthly maintenance on DC emergency
                                                                    lights
                                  WR 05444-88                        Repair Condenser Relief Valve SG K04A
                                  WR 01089-89                        GK V765 hydro motor forLSGK04A, repair
                                                                    oil leak under terminal block
                                  WR 50011-89                        Control Room Air Conditioning Unit
                                                                    SGK04A - 5 year replacement of contactor
                                  WR 50212-89                        Turbine building supply fans - semiannual
                                                                    fan inspection
                                  WR 50213-89                        Turbine building supply fans    perform
                                                                    2-month maintenance
                                  Selected NRC inspector observations are discussed below:                                                          l
                                  o      During the performance of WR 50011-89, the maintenance workers .
                                        observed that the replacement contactor was slightly different from
                                        the old contactor. The workers reverified that they had the correct
                                        part and then commenced the replacement.      When the workers attempted
                                        to reland the center lug, they discovered that it would not fit.                                    The
                                        workers. suspended the replacement and contacted maintenance
                                        engineering. Later, the lug was trimmed slightly in accordance with
                                        plant procedures and the contactor replacement was completed. The
                                        workers were observed to perform their associated tasks in a
                                        professional manner and to request assistance as needed.
                                  o      The workers performing WR 50212-89 and WR 50213-89 made use of
                                        temporary scaffolding.    The temporary scaffolding had been in place
                                        for several years and, to date, there are no plans to install
                                        permanent scaffolding. The licensee informed the NRC inspectors that
                                        an engineering evaluation request (EER) had been issued to install
                                        permanent scaffolding. The EER was issued on November 30, 1988.
                                  No violations or deviations were identified.
                          7.      Review of Licensee Event Reports (LERs)        (92700)
                                  During this inspection period, the NRC inspectors performed followup on
                                  WCGS LERs.      The LERs were reviewed to ensure:
  4
      *                                                                                                                                            I
            ''
_
                  '1    '
                            .__      __
 
m          ..          .                        .
                                                        ..
b              ,            $        .
                                          -
                                              ,l          ,
                                                            ,
  .
      .        .
                          -t                .
                                                    ,.
                                                                                                                j
c.
l
        .''
                        .
                                                                      -10-
                                                          *
          ,
                                          Ll
L                                o    . Corrective action stated in the report ha's been properly completed or
L                                    ' work is in progress.
                                o        Response to the event was. adequate.
                                o'    Response to the event met license conditions, commitments, or other
l                                      applicable regulatory requirements.
H                              -o      The'information contained in the report satisfied applicable
                                          reporting requirements.
                                ,o      Generic issues /were-identified.
                                    ~
                              'The following LER was. reviewed and closed:
                                o      89-004, " Loose' Terminal Connections:Cause Main Steam Isolation' Valve
                                        Closure Resulting in Reactor Trip." This LER is discussed in
                                        paragraph 8 and is closed.
                                During the refueling outage in 1988, the licensee replaced the chlorine
                                monitors in the' control room ventilation system. These chlorine monitors
                              .have been the. subject of several LERs reporting engineered safety feature-
                                actuations/ control room ventilation isolation signals. In most cases, the-
                                problems were because of chlorine' sensitive paper tape malfunctions. The-
                                paper tape would break or bunch giving false indications. Frequently, no
                                specific-root cause of the paper problem was found and the licensee
                              ' decided to replace the chlorine monitors. The new monitors have been.in
                                service for approximately 2 months and have not caused any false 6ctuation
                                signals. The chlorine monitor issue was also discussed in NRC Inspection
                                Report 50-482/88-17, although no specific inspection finding was issued.
                                The following'LERs all relate to the replaced chlorine monitors and are
                          "Iclosed:
                                o    87-032, " Engineered Safety Features Actuation - Control Room
                                      Ventilation Isolation Signal
                                                                '
                                                                    Caused By Paper Tape Breaking On
                                      Chlorine Monitor"
              *                o    87-035, " Engineered Safety Features Actuation - Control Room
                h                    Ventilation Isolation Signals - Two Events - Caused By Malfunctions
                    '-
                                      Of The Chlorine Monitors"
    .
                  ,              o    87-053, " Engineered Safety Features Actuation - Control Room            !
                                      Ventilation. Isolation Signal Caused By Paper Tape Bunching Up On
                                                      ~
                                      Chlorine Monitor"'
                                o      88-003', " Engineered Safety Features Actuation - Control Room
                                      Ventilation Isolation Signal Caused By Paper Tape Spurious Spike On
                      '
                                      Chlorine Monitor"
                                                                                                                l
                                                                                                                l
                            .
  .                                                                                                            ,
 
    -_ _ _ - _ _ _ _ .
l
                                  *
                      .
                            .      .
                                    ..
                              '
                                                                            -11-
                                        o    88-005, " Engineered Safety Features Actuation - Two Control Room
                                              Ventilation Isolation Signals Caused By Malfunctions Of The Chlorine
                                              Monitors"
                                        o    88-006, " Engineered Safety Features Actuation - Control Room
                                              Ventilation Isolation Signal Caused By Paper Tape Spurious Spike Of A
                                              Chlorine Monitor"
                                        o    88-008, " Engineered Safety Features Actuation - Control Room
                                              Ventilation Isolation Signal Caused By Paper Tape Breaking On
                                              Chlorine Monitor"
                                        o    88-010, " Engineered Safety Features Actuation - Control Room
                                              Ventilation Isolation Signal Caused By Loss Of Power To Chlorine
                                              Monitor"
                                        o    88-011', " Engineered Safety Features Actuation - Control Room
                                              Ventilation Isolation Signal Caused By Paper Tape Breaking On
  +
                                              Chlorine Monitor"
                                        o    88-012, " Engineered Safety Features Actuation - Control Room
                                              Ventilation Isolation Signal Caused By Failed Photocell On Chlorine
                                              Monitor"
                                f                                                                                  *
                                        o    88-013, " Engineered Safety Features Actuation. ' Control Room
                                            ' Ventilation Isolation Signal Caused By Paper Tape Bunched Up On
                                              Chlorine Monitor"
                                        o    88-022, " Engineered Safety Features Actuation - Control Room
                                              Ventilation Isolation Signal Caused By Loss Of Power To Chlorine
                                              Monitor"
                                        o    88-026, " Engineered Safety Features Actuation - Control Room
                                              Ventilation Isolation Signal Caused By Paper Tape Bunched Up On
                                              Chlorine Monitors"                                                      q
                                  8.    Onsite Followup of Events at Operating Power Reactors (93702)
                                        The purpose of this inspection activity was to provide onsite inspection
                                        of events at operating power reactors. Specific inspection activities
                                        included:
                                                                                                                      ]
                                        o    Observing plant status,
                                        o    Evaluating the significance of the events, performance of safety
                                              systems, and' actions taken by the licensee.
                                                                                                                      i
                                        o    Confirming that the licensee had made proper notification of the
                                              events and of any new developments or significant changes in plant
                                              conditions.
                                                                                                                      l
                                                                                                                      l
                                                                                                                      I
        _ _ _ _ _ _ _ - _ _
 
                                                                                                                                _ _. ___  __ _ _ _ _ _
    ,,
                  .
      1.
          .
l                .
            ,-
                - +.
                                                            -12-
  ,                    o    . Evaluating the need for further or continued NRC response to the-
                              events.
                        The following items were considered during the followup:
                        o    Details regarding the cause of the event.
                        o    Event chronology.
                        o    Functioning of safety systems'as required by plant conditions.
                        o    Radiological consequences and personnel exposure.
                        o    Proposed licensee actions to correct the cause of the event.
                        o,    Corrective actions taken or planned prior to resumption of facility
                              operations.
                      . The event that occurred.during this report period is listed in the table
                      below:
              '
                        Date        Event                Plant Status                                                Cause
        -              02/02/89. Reac. or Trip              Mode 1                                Loose screw
                                                          (100% power)
                        Selected NRC inspector observations are discussed below:
                      o      un February 2, 1989, at 1:22 p.m. /CDT) the plant tripped on low-low
                              water level in Steam Generator "f." The cause of the low level was a
                              pressure spike resulting from the tast closure of "C" main steam
                              isolation valve (MSIV).
                              An' instrumentation and control technician was performing work inside
                              one of,the solid state protection system (SSPS) cabinets and
                              apparently contacted a plastic cable raceway housing the wiring for
                              the actuation logic for "C" MSIV. Loose screws inside the housing
                              caused a single train fast closure signal to be generated. All
                              equipment functioned as designed with two exceptions:
                              a.    Immediately following .the reactor trip, crtrol room
                                    instrumentation indicated that MSIV "C" hE not fully closed.                                          A
                                    senior' reactor operator (SRO) dispatchM i the MSIV observed
                                    what he thought was MSIV "C" going slow closed. Licensee
                                    troubleshooting efforts identified indications that the valve
                                    did fast close and the SRO had seen normal valve pulsations.
                                    The MSIV passed surveillance testing and was returned to
                                    service.
                            b.    As would be expected for an MSIV closure at full power, the
                                    licensee observed two main steam code safety valves lift and
                                                                      _ _ - _ _ _ _ _ . _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - -
 
                                                      -
                                                                                                ,              -.
        77 7                                                                                                              ,
                                                                                                                              3
bl j.                            '
                                                                                                                    , _
o    s.        .
                                .
    in              ,
                              ,
                            >
                                    '
L                        .
                                                                          -13-                                          '
                                                  reseat.  One valve, however, was observed to have some slight
                                                  seat' leakage. After evaluation of appropriate accident
                                                  scenarios, the licensee concluded that this slight leakage did
                                                  not endanger public health and safety and was acceptable for a
                                                  plant restart.  The leakage has since_ stopped.
                                            The licensee was unable to determine the root cause of the Icase
                                            screw. However, the licensee did perform a check of all other vendor
                                            connections in the SSPS cabinets, balance of plant engineered safety
                                            features cabinets, process / control cabinets, reactor trip breaker
                                            cubicles, main control boards, and other control room cabinets.
                                            Overall, the number of loose screws was less than 1 percent.
                                            Generally, the licensee considered as loose any screw which could be
                                            tightened by a technician more than 25 percent of a turn. Most of.
                                            the SSPS cabinet connections were loose; however, there have been no
                                            surveillance failures attributable to these loose screws.    A check of
                                            these screws will be added to the preventative maintenance program.
                                            LER 482/89-004 is closed.
                                      No violations or deviations were identified.
                              9.    Installation and Testing of Modifications (37828)
                                      The purpose of this inspection was to evaluate onsite activities and
                                      hardware associated with the installation of plant modifications and to
                                      ascertain that related modification activities, which are not submitted
                                      for approval to the NRC, are in conformance with NRC requirements. The
                                      NRC inspectors examined installed hardware to selectively verify that the
                                      modifications conformed to licensee drawings. This included confirmation
                                      of equipment model or serial numbers, dimensions, materials, sizes, heat
                                      numbers, and lot numbers.
                                      Two plant modification requests (PMRs) were se M ted for review. One PMR
l
                                      required field installation work and one PMR required documentation
l                                    change.
                                      Selected NRC inspector observations are discussed below;
                                      a.    PMR 00264/KN84-088, " Aux Feedwater Pump Recirculation Flow
                                            Indication" - TSSR 4.7.1.2.1.a requires that, at least once per
                                            31 days, one of the AFW pumps be tested on recirculation flow for
                                          proper discharge pressure. The original system design had a flow
                                            orifice and connection points for a differential pressure gauge in
l*                                        each recirculating line; however, test instrumentation
                                            (flow-indicators) had to be installed and removed for each test.                ,1
                                            PMR 00264, Revision 1, installed permanent flow indicators'(actual
                                                                                                                            '
  *>
  <                                        field work was done under WR 05000-86 and WR Package 01359-86) during
ib                                        the Fall 1986 refueling outage.                                                    .j
L
I                                          The NRC inspector's field walkdown identified one discrepancy.
                                            Installation Drawing PMP CS-545-W-J-14AL26(Q), Revision 0, for Flow
        _ _ _ - _ - _ _ -
 
,
I
l
        :  J
l
            , .
          ,
                                                -14-
                Indicator AL FI-49, and Drawing PMP CS-545-W-J-14AL28(Q), Revision 0,
                for AL FI-51, both specify that the instrument tubing slope be at
                least 1 inch per foot (i.e., greater than or equal to 0.083). The
                NRC inspector's measurements showed that for AL FI-49 the high side
                tubing had a slope of 0.058 and the low side tubing had a slope of
                0.044 and that for AL FI-51 the high side tubing had a slope of 0.057
                and the low side tubing had a slope of 0.058. Procedure CNT-700,
                Revision 0, " Fabrication and Installation of Instrumentation,"
                Step 4.8.1.1.c states that a horizontal and/or vertical roll of up to
                (+ or -) 1 1/2 inches can be used and that the vertical roll shall
                not violate the minimum slope criteria specified on the instrument        j
                isometric drawings. In addition, Step 5.8.1.1.5 on QC has the QC
                insp6          verify that tubing is sloped per design drawings. However,
                the s_; +ry of the tubing run is such that changes to the slope
                            .
                could occur because of maintenance or other activity in the vicinity
                of the tubing after installation.
                Upon notification of the NRC inspector's findings, the licensee
                dispatched QC and construction personnel to evaluate the tubing.
                Using more accurate methods, the licensee determined that the tubing
                for AL FI-49 did meet acceptance criteria and that the tubing for
                AL FI-51 was sloped at 0.9 inches per foot and 0.6 inches per foot.
                Licensee personnel stated that the slight deviation did not invalidate
                the design pu gose of the pipe slope. The licensee issued
                WR 01217-89 to ask for an engineering disposition of this condition
                and issued Programmatic Deficiency Report PQ 89-02 to address
                programmatic concerns.
                During the procurement phase of PMR 00264, the licensee identified
                inconsistencies between the measured monthly surveillance flowrates
                and the design flowrates. This became readily apparent during
                attempts to use the new instrumentation. Using design flowrates, the
                flow instruments that were installed had a range of 0 to 100 gallons
                per minute (gpm). Flow rates during surveillance are up to 135 gpm.
                The new flow indicators for the MDAFW pumps were unusable and test
                instruments had to be used again. The flow indicator for the TDAFW
                pump was 0 to 200 gpm, and was not overranged. The licensee
                identified three issues:
                (1) The new instruments were unusable;
                (2) Too much flow might be going through the recirculating line
                      during normal injection lineups and preventing a proper amount
                      of flow from being injected into the steam generators; and
                (3) The high recirculating flows could be causing erosion / corrosion
                      of the recirculating lines.
                The licensee's engineering evaluation of Issue No. 2 showed that, at
                present, the AFW pumps were supplying the required flow to the S/Gs.
                However, over time, normal wear of the pumps would cause flows to be
                reduced. Issue No. 3 and the long-term portion of Issue No. 2 caused
                the licensee to consider adding throttle valves in the recirculating
_ _ _ _                                                                                  i
 
                    _ -__ __-
                              s          l
l                                      ,
                                        + .
                                                                          -15-
                                            line. This would reduce the flow rate through the recirculating line
                                            and, thus, the erosion / corrosion rate.  It would also reduce the flow
                                            rate to a point within the range of the newly installed
                                            instrumentation.
                                            This new design was identified as PMR 00264, Revision 2, and was
                                            originally planned to be implemented during Refueling Outage III.
                                            However, the modification was not perfomed and engineering failed to
                                            realize this until the NRC inspector asked for a copy of Request For
                                            Engineering / Design Assistance (REDA) N-P-8136-AL, Revision 1.  Block 3
                                            of the REDA states, in part, that the existing design is not an
                                            acceptable long-term solution. Block 5, required date of completion
                                            states, that the design is to be implemented by Refuel III. Upon
                                            realizing that a recommended modification was not completed on time,
                                            the licensee analyzed the existing flows to ensure that all AFW pumps
                                            still met their required flows and performed ultrasonic inspection of
                                            critical areas of the AFW recirculating piping. No erosion / corrosion
                                            areas were found. The licensee currently plans to perform PMR 00264,
                                            Revision 2, during Refueling Outage IV, scheduled to start in March
                                            of 1990.
                                            PMR 00264, Revision 1, had been completed in December 1986, however,
                                            as of the date of this inspection, plant drawings still had not been
                                            updated to show this modification. When questioned, the licensee
                                            stated that because the PMR was still being implemented, the drawings
                                            had not been "as-built." Because Revision 2 may not be completed
                                            until mid-1990, this could mean that control room drawings would be
                                            " redlined" for 31/2 years, before having Revision 1 incorporated.
                                            In addition, upon checking the control room drawings to ensure that
                                            they had been properly " redlined," the NRC inspector determined that
                                            the " redlining" was missing. When the licensee was notified, the
                                            drawings were promptly " redlined." Failure to color code (" redline")
                                            drawings to reflect changes in accordance with licensee procedures is
                                            an apparent violation (482/8905-03).
                                            In October of 1988, licensee M Surveillance TE: 53359 S-1659,
                                            " Redlining Control Room Drawings," identified licensee failures to
                                            keep control room drawings up to date. QA issued violations and the
                                            licensee initiated corrective actions. While the " redlining" of
                                            drawings has improved dramatically, the results of this inspection
                                            show that there are problems that still remain to be corrected. In
                                            addition, the NRC inspectors believe that the practice of not
                                            updating drawings in a timely fashion contributes to " redlining"
                                            problems such as cluttered or, in the worst case, illegible drawings.
                                            The NRC inspectors informed the licensee that having modifications
                                            made to the plant for 3 1/2 years before drawings were updated was
                                            not, in their view, a good practice. The licensee is encouraged to
                                              review its policy regarding this practice.
                                            Licensee documentation reviewed during the performance of this part
                                            of the inspection is listed in Attachment 1.
                                                                                                                      I
_ _ _ _ _ - - _ - - . - - _ - _ _ - - -                                                                              1
 
                                                                                                                        . . _ _ _ _ _
        .
.
  ,
      . .
                                                                                              -16-
                                                                                                                                        )
                                                                                                                                      i
                                    b.          The NRC inspectors reviewed PMR 2736 which revised the surveillance
                                                  testoftheultimateheatsink(UHS). This, in turn, led to a review                      )
                                                  of the surveillance test and . test results. The surveillance includes              I
                                                  checks for movement of the UHS dam, sedimentation of the UHS, and
                                                  sedimentation of the intake channel to the ESW intake structure. The
                                                  surveillance was compared with TS surveillance requirements and the
                                                  USAR.  In the " Periodic Inspection Report for Ultimate Heat Sink and
                                                  Associated Safety Related Structures" approved April 5,1988, the
                                                  licensee discussed an approximately 23 acre-feet (A-ft)
                                                  sedimentation. USAR Section 2.4.11.6 states that the maximum
                                                  estimated sedimentation is 33 A-ft/ square mile over 40 years. The
                                                  UHS was sized for two units and the licensee.had a study performed on
                                                  this issue. The new study, according to the April 5, 1988.
                                                  inspection report, showed that only 24 A-ft could be lost to
                                                  sedimentation and still support a two-unit shutdown; however,
                                                  129 A-ft could be lost and still support one unit shutdown. The
                                                  results of this study have not been reflected in the USAR. This is
                                                  considered an open item pending the licensee's revision.of the USAR
                                                  (482/8905-05).
    10.                              Exit Meeting
                                    The NRC inspectors met with licensee representatives (denoted in
                                    paragraph 1) on March 7,1989. The NRC inspectors summarized the scope
                                    and findings of the inspection. The licensee did not identify as
                                    proprietary any of the information provided to, or reviewed by, the NRC
                                      inspectors.
                                                                                                                                      I
      _ _ _ - - _ - _ _ _ - _ _ _ - _ . - _ - _ _                        . _ _ - _ _ . - _ _
 
                                                                                                                              l
                                        -
    .
  ;. e                            .
              ,
                                  .              .
                                                                                    ATTACHMENT 1
                                                  PMP-CS-545-W-J-14AL26(Q), Revision 0, " Turbine Driven Auxiliary Feedwater
                                                  Pump Flow-Instrument Isometric Drawing"'
                                                  PMP-CS-545-W-J-14AL27(Q), Revision 0, " Motor Driven Auxiliary Feedwater
                                                  Pump "A" Flow-Instrument Isometric Drawing"
                                                  PMP-CS-545-W-J-14AL28(Q), Revision 0, " Motor Driven Auxiliary Feedwater
                                                  Pump "B" Flow-Instrument Isometric Drawing"
                                                  J-07G37(Q), Revision 2, " Instrument Tubing Clamp Mounting Q Instrument
                                                  Installation"
                                                  J-07G17(Q), Revision 7, " Instrument Tubing Support"
                                                  J-07D12(Q), Revision 3, " Instrument Mounting Detail D.P. Indicator
                                                  (Barton) Packless Manifold"
  i
                                                  J-07G05(Q), Revision 6, "Five Valve Manifold Auxiliary Mounting Brackets"
                                                  J-07G01(Q), Revision 10, " Instrument Mounting Structure Floor Stand"
                                                  M-13ALO5(Q), Revision 2, " Piping Isometric Auxiliary Feedwater Pumps
                                                  Recirculation Piping"
                                                  J-07G22(Q), Sheets 1 and 2, Revision 11, " Bill of Materials "Q" Instr.
                                                  Installations"
                                                  C-1037(Q), Revision 0, " Civil Structural Standard Details Sheet No. 34"
                                                  M-12AL01, Revision 0, " Piping & Instr. Diag. Aux. Feedwater System"
                                                  KGP-1131, Revision 6, " Plant Modification Process"
                                                  CNT-700, Revision 0, " Fabrication and Installation of Instrumentation"
                                                  ADM 01-042, Revision 12, " Plant Modification Request Implementation"
                                                  Work Request 05000-86, Install Flow Indicators AL FI-49, -50, and -51
3
                                                  Work Request Package 03159-86, Install Flow Indicators AL FI-49, -50, and
                                                  -51
                                                  Field Change Request KN84-088-I-002                                          i
                                                  Plant Modification Request 00264/KN84-088, Revision 1, " Aux Feedwater Pump
                                                  Recirculation Flow Indication"
                                                  Nonconformance Report M-1318, Revision 0, " Items received built to a
                                                  different code edition"
                                                  Field Change Request KN84-088-C01
f
      - - - _ - _ _ _ _ _ . _ _ - _ - - - _ _ .                    __ _ _ - _ _ _ .
 
          - _ . - _ - _
              i
                        +
                            .
        >
                          * **'*
                                                                    3
t
                                QA Surveillance TE: 53359 S-1659,." Redlining Control Room Drawings"
                                QA Surveillance TE: 53359 S-1629, " Plant Modification Requests"
                                QA Audit TE:  50140-K?.11, " Modifications"
                                PMR 00264/KN84-088, Revision 2, draft
  .
. _ - - - - _ - - -
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